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Preface
A. General Provisions
Article 6

Existing Nuclear Installations ···················································································· 6-1

6.1

Existing Nuclear Installations in the Scope of this Convention ·········································· 6-1

6.2

The Safety of Existing Nuclear Installations ······································································· 6-2

6.3

Evaluation and Verification of Safety, and Position as to Continued Operation ··············· 6-12

B. Legislation and Regulation
Article 7

Legislative and Regulatory Framework ··································································· 7-1

7.1

System of Legislations Governing the Nuclear installations ··············································· 7-1

7.2

Legislative and Regulatory Framework at Each Stage ························································ 7-7

7.3

Record, Applicable Regulations and Enforcement of Terms of License ··························· 7-11

7.4

Change the Technical Standards for Nuclear Installations to Performance Requirements············································································································································· 7-12

Article 8

Regulatory Body ·········································································································· 8-1

8.1

Mandate and Duties of Regulatory Body ············································································ 8-2

8.2

Organizations for Enforcement of Safety Regulation of Nuclear Installations ··················· 8-2

8.3

Nuclear and Industrial Safety Agency (NISA) ···································································· 8-3

8.4

Organization related to NISA ······························································································ 8-6

8.5

The NSC ······························································································································ 8-7

8.6

The AEC ······························································································································ 8-9

8.7

Other Administrative Bodies ······························································································· 8-9

Article 9

Responsibility of Licensee ·························································································· 9-1

9.1

Regulatory Measures for the Licensee to take the Prime Responsibility ···························· 9-1

9.2

Supervision of Licensees by Regulatory Body ··································································· 9-2

9.3

Communication with Licensees ·························································································· 9-4

C. General Safety Considerations
Article 10

Priority to Safety ····································································································· 10-1

10.1 Basic Policy for Priority to Safety ····················································································· 10-1
10.2 Efforts for Improvement in Safety Culture ········································································ 10-2iArticle 11

Financial and Human Resources ··········································································· 11-1

11.1 Financial Resources of the Licensee ················································································· 11-1
11.2 Human Resources concerning the Nuclear Installation of the Licensee ··························· 11-2
11.3 Efforts for Ensuring Infrastructure of Human Resources in the Regulatory Bodies ········· 11-4
11.4 Maintaining Human Resources in Nuclear Fields in Japan ··············································· 11-5
Article 12

Human Factors ········································································································ 12-1

12.1 Efforts by Regulatory Body ······························································································· 12-1
12.2 Efforts by Licensee ············································································································ 12-4
Article 13

Quality Assurance ··································································································· 13-1

13.1 Legislative Framework for QA of Nuclear Installation ····················································· 13-1
13.2 Verification of Quality Assurance by Nuclear and Industrial Safety Agency ··················· 13-2
13.3 Implementation and Evaluation of QA Program by Licensee ··········································· 13-4
Article 14

Assessment and Verification of Safety ·································································· 14-1

14.1 Safety Assessment prior to Construction ··········································································· 14-2
14.2 Assessment and Verification of Safety Prior to the Commissioning ································· 14-3
14.3 Assessment and Verification of Safety during Operating Life Time ································· 14-5
14.4 Utilization of Probabilistic Safety Assessment in Regulation ··········································· 14-8
14.5 Assessment of Seismic Safety ························································································· 14-12
14.6 Promotion of Nuclear Safety Research ············································································ 14-14
Article 15

Radiation Protection ······························································································· 15-1

15.1 Summary of Laws and Requirements on Radiation Protection ········································· 15-1
15.2 Laws and Requirements and Response of Licensees ························································ 15-3
15.3 Regulatory Control Activities ···························································································· 15-8
Article 16

Emergency Preparedness ······················································································· 16-1

16.1 Development of Laws and Rules for Nuclear Emergency Preparedness ·························· 16-1
16.2 Nuclear Emergency Preparedness and the Measures ························································ 16-4
16.3 Implementation of Nuclear Emergency Drill ···································································· 16-7
16.4 International Framework and Relationship with Neighboring Countries ·························· 16-9

ii

D. Safety of Installations
Article 17

Siting······················································································································· 17-1

17.1 Basic Concept on the Siting of Nuclear Installations ························································ 17-1
17.2 Principal Assessment System Concerning the Siting of Commercial Power Reactors ····· 17-1
17.3 Evaluation to Events Caused by External Factor ······························································ 17-2
17.4 Evaluation for the Impacts to the Public of Accidents ······················································ 17-3
17.5 Environmental Impact Assessment ···················································································· 17-4
17.6 Re-evaluation of Site Related Factors ··············································································· 17-5
17.7 Arrangements with Neighboring Countries on Safety Impact of Nuclear Installations ···· 17-5
Article 18

Design and Construction ························································································ 18-1

18.1 Review at the Design and Construction Stage of Nuclear Installations ···························· 18-1
18.2 Realization of Defense in Depth and Confinement of Radioactive Materials at the Design andConstruction Stage ······················································································································ 18-2
18.3 Systems for preventing the occurrence of abnormalities and for mitigating the impact ofabnormalities (Regulatory Guide for Reviewing Classification of Importance of Safety Functions forLight Water Nuclear Power Reactor Facilities) ··········································································· 18-4
18.4 Safety Design Assessment ································································································· 18-6
18.5 New Regulatory Guide for Reviewing Seismic Design of Nuclear Power Reactor Facilities············································································································································· 18-7
18.6 Preparation of Accident Management Measures ····························································· 18-10
18.7 Measures to Ensure the Technical Reliability by Experience, Test and Analysis ············ 18-11
18.8 Consideration of Human Factors and Man-Machine Interface ······································· 18-13
Article 19

Operation ················································································································· 19-1

19.1 Initial License ···················································································································· 19-2
19.2 Limiting Conditions for Operation ···················································································· 19-2
19.3 Regulations for Operation, Maintenance, Inspection and Testing ····································· 19-2
19.4 Response to Accidents and Anticipated Operational Occurrences ·································· 19-18
19.5 Engineering and Technical Support ················································································· 19-19
19.6 Reporting of Incidents ····································································································· 19-19
19.7 Collection, Utilization and Sharing of Operating Experience Information ····················· 19-20
19.8 Management of Spent Fuel and Radioactive Waste ························································ 19-21
Activities for Improvement of Safety

Annexesiii

Acronym and Abbreviation Used in this Report
ABWR

Advanced Boiling Water Reactor

ACNRE

Advisory Committee for Natural Resources and Energy

AEC

Atomic Energy Commission

AESJ

Atomic Energy Society of JapanALARA

as low as reasonably achievable

ANRE

Agency of Natural Resources and Energy

APWR

Advanced Pressurized Water Reactor

BSS

Basic Safety Standards

BWR

Boiling Water Reactor

Comprehensive

Comprehensive Check of Electric Power Facilities

Check
DNB

Departure from Nucleate Boiling

Dose Limit

Notification for Dose Limits on the basis of the Rules for Commercial Power Reactors

Notification
Electric facility

Facility or equipment that is installed for power generation, transformation, supply,
distribution and utilization

ERSS

Emergency Response Support System

FY

Fiscal Year

IAEA

International Atomic Energy Agency

ICRP

International Commission on Radiological Protection

INES

International Nuclear Event Scale

INPO

Institute of Nuclear Power Operations

IRRS

Integrated Regulatory Review Service

JAEA

Japan Atomic Energy Agency

JANTI

Japan Nuclear Technology Institute

JCO Criticality

Criticality Accident at JCO Co. Uranium Fuel Fabrication Facility

Accident
JEA

Japan Electric Association

JEAC(G)

Japan Electric Association Code (Guideline)

JNES

Japan Nuclear Energy Safety Organization

JPDR

Japan Power Demonstration Reactor

JSME

Japan Society of Mechanical Engineers

LCO

Limiting Conditions for Operation

METI

Ministry of Economy, Trade and Industry

MEXT

Ministry of Education, Culture, Sports, Science and Technology

Minister of METI

Minister of Economy, Trade and Industry

Minister of MEXT

Minister of Education, Culture, Sports, Science and Technology

iv

MITI

Ministry of International Trade and Industry (METI at present)

MLIT

Ministry of Land , Infrastructure and Transportation

Monju

Prototype fast breeder reactor owned by JAEAMOX

Mix Oxide

NISA

Nuclear and Industrial Safety Agency

NISS

Nuclear and Industrial Safety Subcommittee

NSC

Nuclear Safety Commission

NS Network

Nuclear Safety Network

NTC

Nuclear Power Training Center

NUSS

Nuclear Safety Standards, IAEA

OECD/NEA

Organization of Economic Co-operation and Development/Nuclear Energy Agency

OSART

Operational Safety Assessment Review Team

PAZ

Precautionary Action Zone

PSA

Probabilistic Safety Assessment

PSR

Periodic Safety Review

PWR

Pressurized Water Reactor

QA

Quality Assurance

R & D reactor

Power reactors at the stage of research and development

Reactor Regulation

Law on the Regulation of Nuclear Source Material, Nuclear Fuel Material andLaw

Reactors

Regulatory Guide forRegulatory Guide for Reviewing Nuclear Reactor Siting Evaluation and Application

Reactor Siting

Criteria

Regulatory Guide forRegulatory Guide for Reviewing Safety Assessment of Light Water Nuclear Power

Reviewing Safety

Reactor Facilities

Assessment
Regulatory Guide forRegulatory Guide for Reviewing Seismic Design of Nuclear Power Reactor Facilities

Reviewing Seismic

etc

DesignRegulatory Guide forRegulatory Guide for Reviewing Safety Design of Light Water Nuclear Power Reactor

Safety Design

Facilities

SCC

Stress Corrosion Cracking

Special Law forSpecial Law of Emergency Preparedness for Nuclear Disaster

Nuclear
Emergency
RM

Relationship Management

SCAP

Stress Corosion Cracking and Cable Ageing Project

SSC

Structures, Systems and Components

UPZ

Urgent Protective Action Planning Zone

WANO

World Association of Nuclear Operatorsv

This page intentionally left blankPreface
1. Preparation of the Report
Organizations involved in the report preparation
This report was prepared by the Nuclear and Industrial Safety Agency (hereinafter referred toas “NISA”) of the Ministry of Economy, Trade and Industry (hereinafter referred to as“METI”). It was prepared in consultation with the relevant government agencies, as well as in
the support of the Incorporated Administrative Agency Japan Nuclear Energy Safety
Organization (hereinafter referred to as “JNES”) and with the cooperation of Japan Nuclear
Technology Institute. Moreover, the report was deliberated by the Nuclear and Industrial
Safety Subcommittee of the Advisory Committee for Natural Resources and Energy under
METI and it also reflects the opinion of the Nuclear Safety Commission.
Items to be considered in preparation of this report
In preparation of this report, special attention was paid to respond appropriately to the items
for which reporting was recommended in the “Summary Report” of the 3rd Review
Meeting(RM), questions on the previous report of Japan raised by other Contracting Parties,
and also items presented as questions or comments to the reporting of Japan in the 3rd RM.
Fundamental policies of the descriptions of this report
In composing this report, results of implementation by the national government and industry
were inserted corresponding to each article of the previous report, and description of it
without any alteration was concisely duplicated so that the whole aspect of the present
framework for ensuring safety could be understood. However, concerning the operating
experience, which is actual evidence for ensuring the safety and improvement of nuclear
installations, only operating experience of three years after the 3rd review meeting (in April2005) were provided as the subject of this report, so as to avoid duplication of the last report.
Description style of this report
In description of style, content revisions from the last report are shown in italics andwhere no changes were made the script remains as in the original (including editorial
amendments). This is so that the reader can easily identify the revisions corresponding to
the three years after the 3rd RM.
Under the legislative and regulatory framework of Japan, nuclear installations in the scope of
this Convention (land-based commercial nuclear power stations) correspond to commercial
power reactors and power reactors at the stage of research and development, of which thesafety regulations for these two types of nuclear installations are fundamentally the same. For
this reason, detailed description in this report is focused on the commercial power reactors as
examples, which have plenty of experience with siting, design, construction and operation, etc.

Preface-1

In addition, the response to “SYNOPSIS OF THE RELEVANT IAEA SAFETYREQUIREMENTSTATEMENTS REFLECTING THE ISSUES ADDRESSED BY ARTICLES
6 TO 19 OF THE CONVENTION ON NUCLEAR SAFETY” is provided in Appendix 5 of thisreport.
2. Current Status of Nuclear Energy Utilization in Japan(1) Situation of utilization of nuclear energy in general
Currently, there are a total of 56 nuclear installations in Japan in the scope of the Convention,
which include 55 units in operation and one unit under construction that has attained
criticality. They have become the main power source supplying about 30.6% of the electric
power production in the 2006 fiscal year.
In the Framework for Nuclear Energy Policy decided by the Cabinet in October 2005, it isstated that the basic concepts of nuclear power generation is 1) aiming to maintain or increase
the current level (around 30 to 40%) of electric power production in 2030 or later, 2) promote
nuclear fuel cycles, 3) utilize fast breeder reactors, etc. METI deliberated concrete measures
to realize these basic concepts, and adjusted the Nuclear Power National Plan in August 2006.
This plan was determined to be promoted intensively and to 1) realize construction of new andadditional nuclear power stations even under the circumstances of deregulation of the powerindustry, 2) utilize the established nuclear power stations by ensuring safety into its major
premises, 3) put the Fast Breeder Reactor Cycle into commercial use in as early a date as
possible, 4) secure quantity and quality of the technical human resources to sustain the next
generation, 5) support industries for the export of the nuclear power generation facility, 6)participate positively in making an international framework with the aim to coexist withexpansion of nuclear power generation and nuclear nonproliferation, etc.
(2) Current Status of Nuclear installations in Japan
As of August 2007, there are a total of 56 nuclear installations in Japan in the scope of theConvention, which include 55 units in operation and one unit under construction that has
attained criticality.
In FY 2006, the total capacity of 49.47GWe of nuclear power generation accounted for about
20.7 percent of the nation’s total capacity of electricity generation, and nuclear power
generated 304.5 billion kWh of electricity that was about 30.6 percent of 995.9 billion kWhelectricity generated in Japan. The average annual capacity factor of nuclear power plants is
70%. The average unscheduled shut-down frequency over FYs 2004 to 2006 was as small as
0.5 times per reactor-year.
3. International Activities for Ensuring Safety of Nuclear installations
Recognizing that international cooperation is essential for ensuring safety of nuclear
installations, Japan has been promoting multilateral and bilateral cooperation.

Preface-2

As for multilateral cooperation, Japan has actively participated in the information exchange
on safety regulation and study of issues concerning nuclear safety in the IAEA, theOrganization for Economic Cooperation and Development/ Nuclear Energy Agency(OECD/NEA), and furthermore, the International Nuclear Regulators Association (INRA) andthe International Nuclear Safety Group (INSAG). The obtained information and the study
results have been utilized for the substantial regulation of Japan.
A new international activity during this reporting period is that Japan participated in thesecond review meeting of the “Joint Convention on the Safety of Spent Fuel Management andon the Safety of Radioactive Waste Management.” This led Japan to have opportunities toreceive peer-reviews of the contracting parties on the safety of radioactive waste management
etc. as well as on the safety of nuclear installations. As for bilateral cooperation, Japan has
been exchanging regulatory information on nuclear safety with the regulatory authorities ofChina, France, Korea, Sweden, the U.K. and the U.S.A. under the bilateral agreement, hasshared its knowledge and experience with them and has been making efforts to enhance thesafety of each other’s nuclear power plants.
On the other hand, licensees are also cooperating actively in managing the World Associationof Nuclear Operators (WANO) Tokyo Center, in order to enhance the safety and reliability ofthe operation of nuclear power stations through the information exchange between utility
operators of Asian nations.
Units 3 and 6 of the Kashiwazaki-Kariwa Nuclear Power Station of TEPCO invited the OSARTof IAEA in November 2004.

4. Important Matters during Reporting Period
Revision of the Regulatory Guide for Reviewing Seismic Design
The Regulatory Guide for Reviewing Seismic Design of Nuclear Power Reactor Facilities to
new nuclear reactors was revised by the Nuclear Safety Commission on September 19, 2006. It
requires a higher level of seismic safety resulting from the alternation of the formulation andevaluation method of earthquake ground motion etc. NISA, deciding that the seismic safety
should be checked based on the new Guide for the existing nuclear installations, instructed
the operators (the licensees of all the nuclear power reactors) to conduct the seismic safety
evaluation and to report the results to on September 20, 2006.
Activities of the Task Force on Inspection System
While promoting firm establishment of the new inspection system which was introduced in
October 2003, the Task Force on Inspection System had resumed in November 2005, in orderto ensure the safety of the aged nuclear power station, and the report, “Improvement in the
Inspection System for Nuclear Power Generation Facilities,” was issued in September 2006.
The main points are (1) to encourage the licensees to conduct their maintenance activities

Preface-3

reflecting the individual characteristics of each nuclear installation with emphasis on aging
management, (2) to strengthen the operational safety activities by including the inspection
during reactor operation in addition to the activities performed during reactor shutdown, and(3) to include thorough remedial actions into the inspection system and to avoid
non-conformity by the licensees
Stipulating performance-code of the technical standard
Heretofore, the technical standard which defines technical requirements of the nuclear
installations included and not only items related to performance but also provisions which
define detailed specifications. In order to enable flexible response to technical innovations orlatest knowledge, the technical standard should be focused only to the performancerequirements necessary for ensuring safety, and the academic society and association
standards should be utilized as detailed specifications which meet the performancerequirements.
Promotion of safety research
NISA established the Nuclear Safety Infrastructure Subcommittee for ensuring nuclear safety
under the Nuclear and Industrial Safety Subcommittee in July 2006, and developed theframework for industries and regulatory organizations to perform planning, implementationand evaluation of the nuclear safety infrastructure study program. The subcommittee prepared
the roadmap of the nuclear safety infrastructure study program so that the program could be
implemented in a well-planned and efficient manner towards ensuring safety. In addition,
considering that many research facilities for nuclear safety studies are on a global scale in thecrisis of possible shutdown in recent years, it has been proposed that the Japan Materials
Testing Reactor (JMTR) of the Japan Atomic Energy Agency is to be positioned strategicallyas an essential facility for the safety infrastructure study program. It is decided that theinternational joint study should also be strongly promoted. For example, the special fundbusiness SCAP project, which reviews stress corrosion cracking (SCC) and the aging of cables,
was entrusted from Japan to OECD/NEA.
Establishment of the Japan Nuclear Technology Institute
In April 2005, the nuclear industry (licensees, nuclear fuel fabricators, plant manufacturers,etc.) established the Japan Nuclear Technology Institute (JANTI) who is actively encouraged
to further improve the voluntary safety activities by the industry and to share and upgrade thesafety culture. Objectives of JANTI are (1) to act for the safety-culture dissemination in
support of licensees, (2) to perform peer review to the activities of licensees, (3) to analyze
and evaluate information from the Nuclear Information Archives “NUCIA” etc., and (4) toevaluate the safety culture of the members objectively with the third-party’s view-points .
Measures for aging management
Implementation of Technical Assessment of Aging and development and implementation of the
Ten Years Maintenance Program were defined as requirements for aged nuclear installations

Preface-4

in laws and ordinances of October 2003. NISA has been confirming their implementing status.
In order to enhance the response to aged nuclear installations, NISA formed the Technical
Information Coordination Committee to share domestic and foreign technical information
among the industrial society, the academic society and the government organizations, and toutilize them effectively. In addition, the Special Committee was established under the Atomic
Energy Society of Japan, with the participation of NISA, JNES, universities, researchorganizations, power utilities, nuclear plant manufacturers and plant engineering companies.
And a road map to the measures for aging management and life-extension with safe-operationof light water reactors was developed by the Special Committee.
Securing human resources in the nuclear field
In Japan, sustaining human resources becomes one of the key issues, due to the retirement ofexpert engineers and decreasing construction opportunities of nuclear installations in recent
years. Consequently, the securing human resources in the future have been reviewed and thenational government and the industrial and academic societies are working together for
personnel training and succession of technical tradition.
Comprehensive check of the electric power facilities
The data falsification in the hydroelectric power station of The Chugoku Electric Power Co.,
Inc. in October 2006 led to the implementation of the Comprehensive Check of electric power
facilities including hydroelectric power, thermal power and nuclear power stations. As aresult, 316 cases on the whole, of which 98 cases were of nuclear power, of falsification andprocedural defect were discovered and reported by 12 electric power companies in March2007.
METI responded to the cases of nuclear installations by (1) invoking the order to the nuclear
power stations of Tokyo Electric Power Co., Inc., Hokuriku Electric Power Co., The Chugoku
Electric Power Co., Inc., and The Japan Atomic Power Co. to amend their Operational Safety
Programs, (2) moving forward or extending the Periodic Inspections of the nuclear power
stations of these electric power companies, (3) implementing special Operational Safety
Inspections and Periodic Inspections with the assignment of the Special Nuclear Installation
Management Supervisor, (4) promoting information sharing on international accidents andtroubles, and (5) directing planning of an action plan for the prevention of recurrence.
The action plan for the prevention of recurrence was submitted from each electric powercompany to METI. The main base examples include participation of the executive officers,
intensive education program on operational safety, recordkeeping of alarm typewriters,cooperation with free access of Nuclear Safety Inspectors, enhancement of independency ofthe chief engineer of reactors, promotion of information sharing through the NUCIA etc.,
notification of deviations from limiting conditions for operation, information sharing between
electric power companies, etc. NISA determined that these action plans are appropriate ingeneral, and they will verify the situation through the Operational Safety Inspection etc. in the
future. Moreover, NISA will also revise the required legislation corresponding to enforcement

Preface-5

of the action plan.
Receiving the report from NISA about the falsification and concealment concerning nuclear
facilities, the NSC has investigated and has tried to understand the whole story. Responding to
the actions taken by METI, the NSC determined “Actions to take, concerning the malicious
conduct of unreported alteration and concealment at nuclear facilities” (See Attachment to
the report on Article 6) on April 23, 2007.
Invitation of IRRSNISA and the NSC requested the Integrated Regulatory Review Service (IRRS) of the IAEA in
2007. The main objectives of the conventional review by the International Regulatory Review
Team (IRRT) was the verification of conformance with the safety standards of the IAEA, the
IRRS to Japan included, in addition to that, extensive policy dialogue with the participating
senior regulators about current issues on regulation. A preparatory meeting was held in
February 2007 and the IRRS plenary meeting was held in June of the same year.
Niigataken Chuetsu-oki Earthquake
On July 16, 2007, a magnitude 6.8 earthquake occurred offshore Chuetsu, in Niigata
Prefecture about 16 km away from the Kashiwazaki Kariwa nuclear power station of Tokyo
Electric Power Co., Inc. Units 2, 3, 4 and 7, which were in operation, automatically shut downwith the scram signal due to this earthquake. Units 1, 5 and 6, which were in shutdown for the
Periodic Inspection.
After this earthquake, a very little amount of radioactive material was released through thedischarge path to the sea in Unit 6 and from the vent stack to the air in Unit 7. The released
amount of activity was 9x10 4 Bq (2x10 -10 Bq/cm 3 , limit defined by law: 0.2 Bq/cm 3 ) to the sea
and 4x10 8 Bq (2x10 -7 mSv, limit defined by law: 1 mSv/year) into the air, respectively. It wasmuch less than the limit defined by law, and resulted in no significant radiation exposure to
the public. Tokyo Electric Power Co., Inc. is investigating the impact on the equipment, etc. in
the power station by this earthquake (as of the end of July 2007).
NISA accepted the IAEA investigation team, in order to achieve international information
sharing of the impact of this earthquake on the nuclear power station.
Moreover, the Investigation/Response on Offshore Chuetsu Earthquake Subcommittee wasestablished under the Nuclear and Industrial Safety Subcommittee of the Advisory Committee
for Natural Resources and Energy in order to investigate the specific impact on the
Kashiwazaki Kariwa nuclear power station by this earthquake and also examine issues to andresponses to be addressed to improve safety after this earthquake by the national government
and the licensee.
The NSC commissioners, immediately after the earthquake, had visited the site and carefully
observed the plants. The NSC received the report from NISA, held a meeting of the“Investigation Project Team on Seismic Safety of Nuclear Facilities” and worked for fact

Preface-6

finding, trying to deliberate future actions. On July 30, 2007, the NSC determined theviewpoint about the impact of this earthquake and the action plans and published the report“The NSC views on, and future actions to take for, the impacts due to the Niigata-ken
Chuetsu-oki Earthquake in 2007”( See Annex 4 of this National Report).

Preface-7

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Preface-8

A. General Provisions

This page intentionally left blankArticle 6 Existing Nuclear Installations
Each Contracting Party shall take the appropriate steps to ensure that the safety of
nuclear installations existing at the time the Convention enters into force for thatContracting Party is reviewed as soon as possible. When necessary in the context of thisConvention, the Contracting Party shall ensure that all reasonably practicable
improvements are made as a matter of urgency to upgrade the safety of the nuclear
installation. If such upgrading cannot be achieved, plans should be implemented to shut
down the nuclear installation as soon as practically possible. The timing of the
shut-down may take into account the whole energy context and possible alternatives aswell as the social, environmental and economic impact.

There are a total of 56 nuclear installations in Japan in the scope of the Convention as of theend of June 2007, which include 55 units in operation and one unit under construction thathas attained criticality.
Since the previous report, Unit 5 of the Hamaoka Nuclear Power Station, Chubu ElectricPower Co. Unit 1 of the Higashidoori Nuclear Power Station, Tohoku Electric Power Co., Inc.
and Unit 2 of the Shika Nuclear Power Station, Hokuriku Electric Power Co., which were
under construction, have attained criticality and they enter the scope of the Convention. The
decommissioning plan of Fugen that had been under preparation for decommissioning has
been authorized and the plant has come out of the scope of the Convention.
Matters on ensuring safety for existing nuclear installations are described in the following.
In addition, for reference, a list of accidents and failures reported since the previous report isshown in Table 6-1 together with the INES evaluation results.
6.1 Existing Nuclear Installations in the Scope of this Convention
There are a total of 56 existing nuclear installations in Japan in the scope of the Convention asof the end of June 2007. The breakdown is shown in the following:
Breakdown of Nuclear Installations in the Scope of the Convention
Type
StatusBoiling water
Commercial
in operation
reactor (BWR)
power
Pressurized water
reactorin operation
reactor (PWR)
Power reactor in research and underdevelopment stage
construction*

Number of units Remarks
32
23
1

Monju

* Unit under construction, which has attained criticality
The existing nuclear installations are listed in Annex 1, and their locations are shown in Fig.
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6-1.
6.2 The Safety of Existing Nuclear Installations
Major studies and measures implemented for ensuring safety of existing nuclear installations
since the previous report are; 1) measures taken for the secondary-system pipe rupture
accident at Unit 3 of the Mihama Power Station (the event was partly reported in the reportfor the 3rd review meeting and at the review meeting), 2) measures taken for the crack etc. of
the hafnium-plate-type control rods of BWR nuclear power plants, 3) measures taken for thestrainer plugging issue in BWR and the containment sump clogging issue in PWR, 4) measurestaken for falsification of the test data on the feed water flow instrumentation, 5) safety review
and assessment of power plants loading mixed-oxide fuel partially in the core, 6) measurestaken for faulty design of the ABWR turbine blades, 7) check on the status of implementing themeasures against sodium leakage of the Monju power station, 8) check on seismic safety of the
Onagawa Nuclear Power Station, 9) measures taken for the partial omission of Periodic
Inspections conducted by the Incorporated Administrative Agency, Japan Nuclear Energy
Safety Organization, 10) implementation of the Comprehensive Check on power generation
facilities, etc.
These cases are summed up as below.
(1) Measures Taken for the Secondary-System Pipe Rupture Accident at Unit 3 of the Mihama
Power Station
1) Summary of the Accident and Investigation Results
On August 9, 2004, at Unit 3 of the Mihama Power Station, the Kansai Electric Power
Co., Inc (hereinafter referred to as "KEPCO")., a main condensate pipe ruptured during
the steady-state of operation at rated thermal power, discharged secondary-system steam
in the turbine building. Workers near the ruptured opening were exposed to thedischarged steam, and five persons died and six persons got injured.
The Nuclear and Industrial Safety Agency (hereinafter referred to as "NISA"), aiming at
an investigation of cause and prevention of recurrence of a similar accident, establishedan accident investigating committee immediately after the accident and compiled thefinal report on March 30, 2005.
The direct cause of the accident was wall thinning of the pipe concerned due to erosion
and corrosion, overlooked for years since KEPCO and Mitsubishi Heavy Industries(hereinafter referred to as "MHI") had not included the ruptured pipe section in the
checklist, and the root cause was the inadequacy of the maintenance management andquality assurance system of these business operators.
The efforts made by NISA and the Nuclear Safety Commission (hereinafter referred to as"the NSC") for this case is described in the following.

6-22) Action by the Nuclear and Industrial Safety Agency(a) Establishment of the Accident Investigating Committee (the Investigating Committee
for the Secondary-System Pipe Rupture Accident at Unit 3 of the Mihama Power
Station)On August 10, 2004, the Accident Investigating Committee was established under theReactor Safety Subcommittee (it has held the meeting 11 times henceforth).
On September 27, 2004, the interim report was compiled and the special Operational
Safety Inspection by NISA and the strict Audit of Licensee's Periodic Check System by
the incorporated administrative agency, Japan Nuclear Energy Safety Organization(hereinafter referred to as “JNES”) of the licensee were implemented based on thereport.
On March 30, 2005, the final report was compiled.
On March 28, 2006, the implementation status of the measures to prevent recurrence
was evaluated (the Investigating Committee meeting was held).
(b) Administrative Actions against Licensees concerning the Secondary-System Pipe
Rupture Accident at Unit 3 of the Mihama Power Station
・ In September 2004, under the name of the Minister of Economy, Trade and Industry(hereinafter referred to as the "Minister of METI"), a severe warning in writing, theorder for conformity to technical standards concerning Unit 3 of Mihama Power
Station and a notification to downgrade the evaluation of the Audit of Licensee's
Periodic Check System were issued to the Kansai Electric Power Co., Inc.
・ The Minister of METI directed JNES to perform a strict review at the Audit ofLicensee's Periodic Check System of the nuclear power stations of the Kansai
Electric Power Co., Inc.,・ The Director General of NISA issued a notification to licensees other than the
KEPCO, to inform his strong expectation to reflect the accident in their preventive
actions.
(c) Implementation of the special Operational Safety Inspection (from the 3rd calendar
quarter, 2004 to the 4th calendar quarter, 2005)
In order to comprehensively confirm the implementation of the measures to prevent
recurrence, the inspection was carried out with the inspection system fully enhanced byinter-changing the inspectors of the Nuclear Safety Inspector's Offices of NISA in the
Wakasa area and dispatching the Nuclear Safety Inspectors of NISA. Moreover, on July
1, 2005, the Regional Nuclear Safety General Manager (in charge of the Wakasa area)
was newly assigned, and the subsequent inspections were conducted with the General

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Manager involved in it.
(d) Confirmation by the Usual Operational Safety Inspection (from the first quarter,
2006)
During the usual Operational Safety Inspection conducted by the Mihama Nuclear
Safety Inspector's Office of NISA, the implementing status of the measures to prevent
recurrence of the accident has been confirmed. In addition, for the matters beyond the
responsibility of the Mihama Power Station, and which is under the responsibility of
the Nuclear Power Division, KEPCO Head Office, confirmation was carried out by theinspectors from three offices; namely, the Takahama Nuclear Safety Inspectors Office,Mihama Nuclear Safety Inspectors Office and Ohi Nuclear Safety Inspectors Office,and the Regional Nuclear Safety General Manager.
(e) Confirmation by the Nuclear Maintenance Confirmation Committee (establishedinside KEPCO on June 17, 2005)
NISA participates as an observer in the Nuclear Maintenance Confirmation Committee,
which is an internal audit system of KEPCO, and consists of third-party members, andchecks implementation of the measures to prevent recurrence.
(f) Confirmation of the measures to prevent recurrence of MHINISA has checked the status of the measures to prevent recurrence of MHI by periodic
hearing (about once / quarter) since August 29, 2005.
The measures to prevent recurrence of MHI has been already implemented, since it isrequired to steadily promote the evaluation and improvement of activities, NISA keeps
a close watch on MHI's measures to prevent recurrence whether or not they haveautonomously implemented checking of the improvement status of the procurement
control by KEPCO.
3) The action status by the NSC
The NSC, immediately after the occurrence of the accident, the NSC held an extra
conference, to investigate the outline of the accident, and, in order to help in the response
to be taken, compiled a report, "Accident at Unit 3 of the Mihama Nuclear Power Stationof KEPCO" In order to investigate and review the technical matters etc. for ensuringsafety of the secondary-system piping of nuclear installations, the Subcommittee on theSecondary System Piping Rupture at Unit 3 of the Mihama Nuclear Power Station wasestablished under the Special Committee on Analysis and Evaluation of Nuclear
Accidents and Failures, and the final report was compiled on April 28, 2005. The
Accident Investigation Subcommittee received the accident investigation progress reports
from NISA, compiled the items to be followed, and the views of the NSC, and informed
NISA of them.

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In addition, the NSC reviewed the responses taken by the government agencies for theaccident of Unit 3 of the Mihama Power Station, through the "Subsequent regulation
reviews” of regulatory body’s programs concerning the regulations of periodic licensee’s
inspections of commercial nuclear power generating facilities and the "Subsequent
regulation reviews” of the regulatory body’s programs concerning the quality assurancein safety preservation activities at the operating commercial nuclear power generating
facilities and the NSC presented their own opinion in a written report issued in February
2005.
Also, the NSC conducted the "Subsequent regulation reviews” of the regulatory body’s
checking actions for recurrence prevention concerning the secondary piping rupture
accident of Unit 3, Mihama Nuclear Power Station, The Kansai Electric Power Company,
Inc. They checked the system of the government agencies to confirm implementation
status of the measures to prevent recurrence of the KEPCO, and compiled the report inMarch, 2006.
4) Events Associated with the Accident and the Measures Taken
Through the investigation of the rupture accident due to wall thinning of thesecondary-system piping of the Mihama Unit 3, the importance of the maintenancemanagement of aged plants and the necessity of measures taken for degradation of theorganizational climate are recognized, and the measures for those issues have been
improved and enhanced (refer to Section 14.3). The steam, which discharged at theaccident, spread into the central control room, which revealed that the air-tightness ofthe central control room was inadequate. Therefore, the habitability of the central control
rooms is currently under study (refer to Section 18.7).
(2) Measures Taken for the Crack etc. of the Hafnium-Plate-Type Control Rods of BWR
Nuclear Power Plants
In January 2006, cracks of the hafnium-plate-type control rods in-service were found at Unit 6of the Fukushima Daiichi Nuclear Power Station of Tokyo Electric Power Co., Inc(hereinafterreferred to as “TEPCO”)., and the licensee reported the event to NISA. NISA informed thelicensees who own BWRs (hereinafter referred to as the "BWR licensees") of the investigationand the report of the results on the usage of the same-type control rods, on the operability ofthe same-type control rods in operating reactors, and on the identity of the cracks etc. of thesame-type control rods in reactors during shutdown.
In February 2006, NISA, as a result of the evaluations of the above-mentioned reports,
informed the BWR licensees using the same-type control rods to use them in the fully-inserted
positions to avoid any potential of not being able to insert them due to cracks, when they are
to be used until their neutron irradiation reaches more than a certain level. (The subject units,
as of the end of March 2006, are the following 23 nuclear reactors using the same-type control
rod; Tsuruga Unit 1, Higashidori Unit 1, Onagawa Units 1, 2 and 3, Fukushima Daiichi Units
2, 3, 4, 5 and 6, Fukushima Daini Units 1, 2, 3 and 4, Kashiwazaki Kariwa Units 1, 2, 3 and 6,

6-5

Hamaoka Units 1, 2, 3, 4 and 5.)
NISA received the reports from TEPCO and Chubu Electric Power Co., Inc(hereinafterreferred to as “Chubu”) concerning the causes analysis and countermeasures forhafnium-plate-type control rods on which the cracks were found in the in-service or spentcontrol rods in May 2006. In response to this, NISA publicly released the NISA's investigation
report, and directed to perform a visual inspection of all same-type control rods at the annual
periodic licensee's check as short-term measures. In addition, NISA and licensees identified
the issues to be studied for medium- to long-term. These measures were reported to theWorking Group for Accident and Failure Countermeasures, Subcommittee for Nuclear
Emergency Preparedness of the Nuclear and Industrial Safety Subcommittee and the NSC.
(3) Measures Taken for the Strainer Plugging Issue in BWR and the Containment Sump
Clogging Issue in PWR
In June 2004, concerning the issue that the recirculation function of the emergency corecooling system could be affected by the plugging of the strainers or by containment sumpscreen clogging during a loss of reactor coolant accident of BWRs and PWRs, NISA, as aresult of this study based on an overseas review status, directed licensees to make a report ofinvestigation on as-build status of thermal insulation etc about BWRs and PWRs andevaluation results on the effectiveness of strainers or screen about BWRs.
In April 2005, the investigation results on the thermal insulation etc. of BWRs and PWRs andinterim measures such as a revision of operation procedures, were reported by several
licensees. NISA considered that the interim measures reported by the licensees were adequate
as they were equivalent in content to the measures of which effectiveness was verified in the
U.S.A. NISA considers that full implementation of the interim measures are essential in order
not to cause an important-to-safety problem until the measures on the equipment, such aslarger strainers or screens etc., will be completed, and NISA instructed the licensees toimplement the interim measures and has been checking the implementation status by the
Operational Safety Inspection etc. For the effectiveness evaluation of the strainers of BWRs, it
was reported that the effectiveness was not confirmed at some nuclear installations as a resultof a conservative evaluation in accordance with the U.S. regulatory guides. NISA studied theadequacy of those evaluations and measures at the Working Group for Safety Assessment, of
the Nuclear Reactor Safety Subcommittee, under Nuclear and Industrial Safety Subcommittee.
As NISA, receiving the results that the larger strainers and replacement of the thermal
insulation are required, instructed in October 2005, to implement the measures for BWRs. All
BWR licensees have been implementing the measures in accordance with the instruction. This
status of implementation was also reported to the NSC.
In October 2005, NISA instructed PWR licensees to submit reports on the result of theeffectiveness evaluation of the PWR screens by August 2006. In August 2006, proposed
measures for the equipment of each plant and responses to the effectiveness evaluation
methods were all submitted and reported to the Working Group, and in May 2007, it was
concluded these measures were necessary to be implemented.

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(4) Measures Taken for the Falsification of Full-Flow Test Data on Feed Water Flow
Instrumentations
NISA, on receipt of an in-house whistle-blowing on November 15, 2005: "almost entire
full-flow certification examination data on the reactor feedwater flow instrumentation of
Toshiba Corporation were falsified", investigated ten nuclear reactors of TEPCO and Tohoku
Electric Power Co., Inc., using the flow instrumentations made by the company. As a result, it
was confirmed that the fraudulent correction had been made on the test data of feed waterflow instrumentations supplied to three units of both licensees. But, NISA determined after its
evaluation that there was no safety or legal issue involved due to the inaccuracy of the feedwater flow instrumentations.
NISA ordered Toshiba Corporation to conduct a thorough root cause analysis and to report
the recurrence-preventing measures established based on the analysis results in order to nevermake such a falsification again. NISA requested of TEPCO and Tohoku Electric Power Co.,
Inc., to establish measures for quality assurance, of procurement for prevention of recurrence
and to then report them. NISA conducted the on-site inspection at Toshiba Corporation, in
accordance with the Reactor Regulation Law in June, October and December 2006, in order
to confirm the implementation status of the recurrence-preventing measures.
(5) Safety Review and Assessment of Power Plants Loading Mixed-Oxide Fuel Partially in theCore
On this matter, the Genkai Units 3 (Kyushu Electric Power Co., Inc.: PWR) was authorized on
September 7, 2005, and the Ikata Unit 3 (Shikoku Electric Power Co., Inc.: PWR) on March
28, 2006, and the Hamaoka Unit 4 (Chubu Electric Power Co., Inc.: BWR) on July 4, 2007.
(6) Measures Taken for Faulty Design of the ABWR Turbine Blades
In June 2006, at Unit 5 of the Hamaoka Nuclear Power Station, Chubu Electric Power Co.,
Inc. (hereinafter referred to as "Hamaoka Unit 5"), the steam turbine and nuclear reactor
were automatically shutdown due to excessive turbine-shaft vibration caused by the breaking
off of a blade of the 12th stage of the low pressure turbine(B) NISA performed a visualinspection of the blades etc. removed from the shaft, which belonged to the same stage as theone of the low-pressure turbine (B) where the blades came off, confirmed breakages andcracks at the fork-shape joint section of the blades, and determined that the turbine did notconform to the Technical Requirements for Nuclear Power Generation Equipment. NISA,
judging that it was necessary to confirm conformity of the steam turbine of Unit 2 of the Shika
Nuclear Power Station, Hokuriku Electric Power Co.,(hereinafter referred to as the "Shika
Unit 2", which is of the same design as the one at Hamaoka Unit 5, according to the same
Technical Requirements, directed to have the blades checked. As a result, breakages andcracks were confirmed on the blades of the low-pressure turbine (B) also at the Shika Unit 2.
In October of the same year, NISA received a report concerning the causes andcountermeasures from Chubu Electric Power Co., Inc. and Hokuriku Electric Power Co., Inc.
The report describes that the short-term measures are to remove all of the 12th-stage blades of
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the low pressure turbines and to apply pressure plates (distributors) to the stationary bladesof the same stage prior to the restart, and the long-term measure is to design and manufacture
new blades of the 12th stage, taking into account the fluid-induced vibration force due torandom vibration and flashback. In November of the same year, NISA released aninvestigation report showing that the licensees' measures for prevention of recurrence are
adequate.
NISA reported these measures to the Working Group for Accident and Failure
Countermeasures of the Subcommittee for Nuclear Emergency Preparedness, under Nuclear
and Industrial Safety Subcommittee and the NSC.
Chubu Electric Power Co., Inc. submitted the construction plan of installation of pressure
plates to the Hamaoka Unit 5 turbine on November 8, 2006. NISA reviewed the plan. Chubu
Electric Power Co., Inc. passed its pre-service inspection and resumed the commercial
operation of the Hamaoka Unit 5 on March 13.
NISA, upon receiving the construction plan of installation of distributors to the Shika Unit 2
turbine on November 13, 2006 from Hokuriku Electric Power Co., examined the plan.
(7) Check on the Status of Implementing the Measures against Sodium Leakage of the Monju
Power Station
At the fast breeder prototype-reactor "Monju", a sodium leak accident of the secondary
cooling system occurred in December 1995, and the reactor has been kept in a
low-temperature shutdown state since that time.
The licensee, Japan Nuclear Cycle Development Institute (in October 2005, the Institute was
integrated with the Japan Atomic Energy Research Institute, and has become an incorporated
administrative agency, Japan Atomic Energy Agency), in July 2007, completed construction
work for the measures to cope with sodium leakage that started in September 2005. Monju is
now conducting verification tests of the completion of the work. NISA, as the prerequisite to
the restart of the Monju, is conducting the following:1) Integrity check on equipment, system and fuel that have not been used for a long period oftime, with pre-service inspection and on-site inspection2) Check on the quality assurance relating to technical capabilities, the time required to
extract the secondary-system sodium, the plant maintenance plan, with pre-service
inspection and on-site inspection.
3) Safety examination for the application of change in fuel assemblies for the first coreloading.
Concerning the safety examination for the application of change in fuel assemblies for the
first core loading, in October 2006, Japan Atomic Energy Agency applied the usage of fuelassemblies prepared for reload of the core as the first core loading to be used at the restart of

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Monju. NISA asked for consultation to the NSC in July 2007.
Various activities for ensuring the safety of Monju have been confirmed openly in the "Study
Group for Confirmation of the Monju Safety", which was established under the Nuclear andIndustrial Safety Subcommittee in November 2005. As of June 2007, the Study Group was held
9 times.
The NSC conducted the audit of NISA’s activities based on the reports about 1) The Results ofConfirmation about the Monju Integrated Safety Check and 2) The Results of Deliberation
about the Important Items after the Approval of Amendment of Monju Establishing License,
and the NSC appraised, in its decision in June 2007, the NISA’s activities concerning
confirmation of Quality Management System as appropriate.
(8) Check on Seismic Safety of the Onagawa Nuclear Power Station
When the earthquake occurred in Miyagiken-oki in August 2005, the Onagawa Nuclear Power
Station of Tohoku Electric Power Co., Inc. experienced earthquake ground motion exceeding
the design basis earthquake ground motion. NISA directed Tohoku Electric Power Co., Inc. toanalyze the factors that caused earthquake ground motion which exceeded the design basis
earthquake ground motion, to check the seismic safety in reference to the important-to-safety
equipments of the power station. In accordance with the direction, Tohoku Electric Power Co.,
Inc. first reported the results of the check on seismic safety for Units 2 and 3 of the power
station to NISA. NISA, based on the evaluation of the report, notified the licensee that theevaluation methods of seismic safety and study results of the seismic safety of the licensee are
adequate. Since Unit 1 of the Onagawa Nuclear Power Station has been in operation for 22
years after commissioning, NISA directed the licensee to assess also the effect of aging on theseismic safety. In September 2006, NISA notified the licensee that the evaluation methods andresults of the check on the seismic safety submitted by the licensee are adequate. For
additional information, refer to the report of Article 14 (Section 14.5).
Units 1, 2 and 3 of the Onagawa Nuclear Power Station were automatically shutdown whenthe earthquake occurred and had undergone the Periodic Inspection after that, they resumed
their operation after receiving notifications of confirmation of the equipment seismic safety
and completing the Periodic Inspections one by one.
(9) Measures Taken for the Partial Omission of Periodic Inspections Conducted by the
Incorporated Administrative Agency, Japan Nuclear Energy Safety Organization (JNES)On February 22, 2007, NISA received a report from JNES, which indicated that an incomplete
inspection (failure to perform a part of functional test) was discovered at the 21st Periodic
Inspection of the Tokai No.2 Power Station, the Japan Atomic Power Company, conducted by
JNES in 2005. On February 23, 2007, NISA ordered to report the result of the check if there
are any other incomplete inspections to JNES in accordance with the Electricity Utilities
Industry Law. On March 9, 2007, JNES reported to NISA that three defects in the record check
were discovered at the Periodic Inspections of other power plants in addition to theabove-mentioned one. The report describes that the root cause of the defects in record check
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(failure to check records, failure to prepare inspection records, etc.) is an inadequate
mechanism to not file or check records due to human error, and the measures to prevent
recurrence are process management by making an appropriate management table andimprovements in the mechanism to check by the administration department. The Director
General of NISA expressed that “it is regrettable that defects were found in the Periodic
Inspections conducted by JNES”, and he issued a severe warning in writing and directed thePresident of JNES to take measures of thorough recurrence-prevention.
(10) Comprehensive Check of Electric Power Facilities
1) Background and Circumstances of the Comprehensive Check of Electric Power Facilities
On November 21, 2006, NISA received a report from the Chugoku Electric Power Co., Inc.,
which describes that the company falsified data regarding a dam for a hydroelectric power
plant in the past. Receiving information that other licensees have also conducted
construction work on hydraulic power production plants without obtaining authorization inaccordance with the River Law, NISA directed all licensees to conduct investigations into
hydraulic power production plants. At the same time, at nuclear power plants, cases such as
inappropriate corrections for the measured temperatures of seawater for cooling were also
revealed.
Receiving a series of such reports, NISA, under direction of the Minister of METI, directedlicensees on November 30, 2006 to check whether there exists any similar data falsification(hereinafter referred to as the "Comprehensive Check") concerning the facilities forhydraulic-power, thermal-power and nuclear-power generation and to make a report on the
results of the check by March 31, 2007.
On March 30, 2007, the licensees reported 316 cases in total to NISA as the results of theComprehensive Check regarding data falsification etc., and on April 6 of the same year,
measures to prevent recurrence were reported.
On April 20, 2007, the Minister of METI issued a document reconfirming the purposes andobjectives of the Comprehensive Check to the presidents of all electric power companies.
The document explains that the main purport of the Comprehensive Check is "to disclose
facts without hiding them", specifically, (1) to cut off the vicious circle of continuing tofalsify records on the premise of past falsifications, (2) to establish a mechanism not toallow a falsification, (3) to share information on accidents and troubles, and utilize themfor prevention of recurrence, and (4) to improve the culture and climate of electric powercompanies by steadily promoting such activities.
The 316 cases were classified into the following four groups according to the contents;
Group 1: Cases that conflict with legislations and regulations, and have effect on thesafetyGroup 2: Cases that are confirmed not to compromise safety, but conflict with

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legislations and regulations and are issues in terms of complianceGroup 3: Cases that are not necessarily related to safety requirements, but conflict withlegislations and regulations and are issues in terms of complianceGroup 4: Errors in writing etc.

The Minister of METI decided to make administrative dispositions according to the contents
and group of the reported cases, and in accordance with the provisions of the
Administrative Procedure Law, issued a document listing notices of the dispositions, sayingthat an appeal against the notice of the dispositions is acceptable to the licensees. Also, theMinister issued a document instructing to submit a specific program and schedule formeasures to prevent recurrence. In addition, the Minister issued a document to the
manufacturers concerned, requesting them to make an action plan to improve a level of
nuclear safety, including a mechanism of sharing information and to promote sharing
information concerning safety technologies when responding to the requirement fromelectric power companies for maintenance management and procurement.
The above-mentioned administrative dispositions were implemented, since there was nooverture of an appeal from the licensees. METI, based on the Comprehensive Check resultson the power generation facilities, also established an action plan that specifies responses
to be taken by METI itself.
2) Instructions to Improve Safety and Safe Operation of Nuclear Power Generation Facilities
The Comprehensive Check results of nuclear power generation facilities showed that there
had been no falsifications since October 2003 when the inspection system was revised.
Although they were the cases within the period before the revision of the inspection system,
there were a total of 98 cases, including 11 cases conflicting with the Reactor Regulation
Law and the Electricity Utilities Industry Law and impairing the safety that the laws intend
to ensure (Group 1) (refer to Table 6-2). Major actions taken by METI after receiving theresults include; (1) issuing an order to amend the Operational Safety Program of thenuclear power stations of TEPCO, Hokuriku Electric Power Co., the Chugoku ElectricPower Co., Inc., and the Japan Atomic Power Co., (2) earlier-than-scheduled
implementation of, or extension of period for the Periodic Inspections of these electric
power companies' nuclear power stations, and implementation of special inspections of the
Operational Safety Inspections and Periodic Inspections including the appointment of NISA
Special Nuclear installation Management Supervisors, etc., (3) amendment of a Ministerial
Order for making it mandatory to report the accidental withdrawal of control rods, etc., andthe promotion of international sharing of information concerning accidents and failures.
On May 21 ,2007, all licensees concerned submitted the action plans for prevention ofrecurrence to METI. The major specific contents include; involvement of executives to theprogram, thorough implementation of education on operational safety, retention of records
by alarm typewriters, supporting free access of NISA Nuclear Safety Inspectors to nuclear
installations, independency of chief reactor engineer, promotion of information sharing with
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NUCIA etc., report of a deviation from the limiting conditions for operation,interdepartmental information sharing, information sharing among electric powercompanies. NISA judged that these action plans are generally adequate as they are basedon the instructions issued by METI and the time schedule and methods to implement theseactions are concretely provided, and decided to check the implementation status through the
Operational Safety Inspection etc. NISA also decided to make necessary amendments etc. oflegislations and regulations corresponding to implementation of the action plan. These
responses were reported to the NSC.
Receiving the report from NISA about the falsification and concealment concerning nuclear
facilities, the NSC has investigated and has tried to understand the whole story. Responding
to the actions taken by METI, the NSC determined “Actions to take concerning themalicious conducts of unreported alteration and concealment at nuclear facilities” on April
23, 2007. As a part of these actions, the NSC ; on May 15, 2007, invited the Chief Reactor
Engineers from each nuclear power plants, interviewed and exchanged the opinions, ; on
May 17 and 31, 2007, heard the opinions of Executive of Hokuriku Electric Power Co. Ltd,and ;on May 28 and June 4. 2007, heard from Japan BWR Owners Group and Japan PWR
Owners Group respectively, about their activities for information sharing and for analyzing
accident and failure.
6.3 Evaluation and Verification of Safety, and Position as to Continued Operation
NISA had implemented the necessary safety assessment and verification for existing nuclear
installations at planning, licensing, construction and operation stages. They are explained in
the reports of Article 7 to Article 19.
Through those assessments and verification, principles of this convention have been applied toensure the safety of existing nuclear installations for every stage from licensing to operation.
As shown in Section 6.2, for events that occurred during the period after the previous
reporting, NISA had judged appropriately whether any issue on safety existed or not,instructed licensees to take measures for ensuring safety, as necessary, and confirmed that themeasures were appropriately taken. Therefore, it is appropriate to continue operation ofoperating nuclear installations.
For the Monju, which is a nuclear installation under construction that has attained criticality,
implementation of safety measures against sodium leakage and the passing thepre-operational inspection are requirements for its operation.

6-12

Table 6-1 Accidents and Failures at Nuclear Power Stations Reported by Licensees during
the Reporting Period

Name ofpower plant
Takahama,
Unit 4
Sendai, Unit 1
Tomari, Unit 1
Fukushima
Daiichi, Unit 2
Mihama, Unit
1
Sendai, Unit 2
Hamaoka,
Units 1 & 2
Ikata, Unit 1
Tsuruga, Unit
2
Kashiwazaki
Kariwa, Unit 1
Onagawa, Unit
1
Mihama, Unit
1
Mihama, Unit
1
Ikata, Unit 3
Kashiwazaki
Kariwa, Unit 5
Shimane, Unit
1
Tokai, No.2
Fukushima
Daiichi, Unit 5
Mihama, Unit
1
Fukushima
Daiichi, Unit 2
Fukushima
Daini, Unit 2
Tomari, Unit 1
Sendai, Unit 1

Title

Event dateSignificant indication by the eddy current
examination of steam generator tubes
Significant indication by the eddy current
examination of steam generator tubes
Significant indication by the eddy current
examination of steam generator tubes
Automatic trip of one reactor-coolant
recirculation pumpInsufficient wall thickness of theturbine-driven auxiliary feedwater piping
Significant indication by the eddy current
examination of steam generator tubes
Cracks of Units 1 and 2 common stack duct
joint
Cracks of ventilation stack of the reactor
auxiliary building
Significant indication by the eddy current
examination of steam generator tubes
Steam leak from the small-bore drain pipe in
the turbine building

September 6,
2004
September 10,
2004
September 21,
2004
September
29, 2004
October 25,
2004
December 15,
2004
December 21,
2004
December 23,
2004
January 18,
2005
February 4,
2005
February 25,
2005
March 19,
2005

Nitrogen leak from the reactor containment

INES level
0−
0−
0−
0−
0−
0−
0−
0−
0−
0−
0−

Failure of B make-up-pump manifold cover0−
bolts
Cracks of the lower part of auxiliary building
April 28,
Out of scaleventilation stack, and the faulty connection
2005
event
of a drain pipe
Failure of the chiller for the central control
May 12, 2005
0−room air conditioner
Automatic reactor shutdown following theJuly 3, 2005
0+
turbine trip due to low condenser vacuum
Indication trouble for closed position of a
July 8, 2005
0−
drywell vacuum breaker valve
Valve stem failure of motor-driven reactor
August 10,
0−
feedwater pump discharge valve
2005
Valve stem failure of the test bypass valve ofAugust 22,0−
the core spray system
2005
Make-up water leakage from the seal of
September 29,
0−
primary coolant pump No. 3
2005
Automatic trip of one reactor-coolant
October 9,0−recirculation pump2005
Damage of a seawater strainer of the cooling November 2,0−system for residual heat removal equipment
2005
Cracks near the weld of stiffener for
January 6,
0−emergency ventilation stack
2006
Significant indication by the eddy current
January 13,0−
examination of steam generator tubes
2006

6-13

Name ofpower plant

Title

Manual shutdown of the nuclear reactor due
to malfunction of a steam supply isolation
Shika, Unit 2
valve of the reactor core isolation coolingsystemFukushima
Cracks etc. of hafnium-plate-type control
Daiichi, Unit 6 rods
Fukushima
Cracks etc. of hafnium-plate-type control
Daiichi, Unit 3 rods
Fukushima
Automatic trip of one reactor-coolant
Daiichi, Unit 2 recirculation pump
Ikata, Unit
Cracks of the weld of steam distributor in the1moisture separator and reheater
Fukushima
Cracks of a valve stem of a residual heat
Daini, Unit 1
removal system flow control valve
Hamaoka,
Automatic shutdown of the nuclear reactor
Unit 5following steam turbine trip
Observation of water puddle containing
Onagawa, Unit
radioactive materials in the reactor-building
2
torus room
Hamaoka, Unit Cracks etc. of hafnium-plate-type control
3
rods
Fukushima
Tritium release to the outside of control zone
DaiichiAutomatic shutdown of the nuclear reactor
Takahama,
due to low steam-generator level during
Unit 3
power down
Cooling water leak from heat transfer tube of
Tsuruga, Unit
the reactor building closed cooling water2
systemShimane, Unit
Corrosion of condensate storage tank1
Shimane, Unit Wall thinning of piping at condensate filter
1outlet header
Manual reactor shutdown following earth
Fukushima
fault of circuit for the automatic
Daiichi, Unit 2
depressurization system
Cracks of extraction piping of the surplus
Genkai, Unit 2
extraction water system
Fukushima
Reactor power change due to a misoperation
Daiichi, Unit 4
Automatic shutdown of the nuclear reactor
Fukushima
due to an alarm of "main steam pipe
Daini, Unit 4
radioactivity high-high trip"
Manual reactor shut-down due toFukushimamalfunction of a valve of the core spray
Daiichi, Unit 5
systemSignificant indication by the eddy current
Sendai, Unit 1
examination of steam generator tubes

6-14

Event date

INES level

January 27,
2006

0−

February 1,
2006
March 3,
2006
March 14,2006
June 5,2006
June 7,2006
June 15,
2006
August 3,
2006
August 7,
2006
August 11,
2006

1
1
0
Out of scale
event
0
0+
0−
1
0−

August 18,
2006

0+

October 4,
2006

0−

October 13,
2006
November 9,
2006
January 17,
2007

0−
0−
0−

January 24,
2007
February 11,
2007

0+(tentative)

February 18,
2007

0−(tentative)

February 20,
2007

0+(tentative)

May 10, 2007

0−(tentative)0−Table 6-2 List of 11 Cases Classified as Group 1 in Nuclear Power
Plant name

Hokuriku
ElectricPower
Co.

Unit 1 of the
Shika
Nuclear Power
StationJune 1999

Unit 3 of the
Fukushima
DaiichiNuclear Power
StationNovember
1978

Unit 4 of the
Fukushima Daini
Nuclear Power
StationFrom
October
1988 toJanuary 1990

Unit 1 of theKashiwazaki
KariwaNuclear Power
StationMay 1992

Unit 2 of the
ShimaneNuclear Power
StationMay 1998

6-15

Electric
power
company

Tokyo
Electric
PowerCo., Inc.

The
ChugokuElectric
PowerCo., Inc.

Time period

Summary• Criticality accident occurred during reactor shutdown (during Periodic Inspection)
During the Periodic Inspection, a criticality accident occurred due to the withdrawal of threecontrol rods caused by inadvertent isolation work of hydraulic control units. The logbook etc.
was falsified in this case and also a report to the national government required by legislations
and regulations was not made. Furthermore, no investigation into the cause was conducted andno measures to prevent recurrence taken.
• Criticality of nuclear reactor due to control-rods withdrawal and falsification of logbook,etc.
During Periodic Inspection, five control rods were withdrawn due to inadvertent isolation
works of hydraulic control units, which resulted in the criticality of the reactor. Since theoperation shift at that time did not recognize the occurrence of criticality and did not take any
special measures, the criticality lasted over 7 and a half hours. In addition, the logbook was
falsified to hide the fact.
• Unlawful actions for construction plan and pre-service inspection of a control rod drive
mechanism,
At scram tests of the control rod drive mechanisms (CRD), trouble occurred at one CRD.
Replacement of the CRD concerned was performed without submitting the construction planfor it. After that, unlawful action such as undergoing the pre-service inspection of the CRD
while using a fake unit was also committed.
• Falsified indication for operation of a residual-heat-removal cooling intermediate pump(A)
Although the motor of the residual-heat-removal cooling intermediate (RHIW) pump (A) was
out of order, emergency diesel generators underwent inspection with the indication lamp
tampered in the central control room so as to make it look as if the pump started. Then, thenuclear reactor was started without checking the integrity of other systems, which is requiredby the Operational Safety Program.
• Negligence of checking the operability of other trains in repairing diesel engine coolingwater leakage
Although one train of the emergency diesel generators was inoperable with the reactor at rated
power, the record of having conducted tests for other trains, which are required by the
Operational Safety Program to continue operation, could not be confirmed.

Electric
power
company

Plant name
Unit 1 of the
ShimaneNuclear Power
Station

6-16
The Japan
AtomicPower
Company

Time period

June 2001

Unit 2 of the
Tsuruga Power
StationJanuary 1994

Unit 1 of the
Tsuruga Power
StationFrom
September
1995 toMarch 2000

Unit 2 of the
Tsuruga Power
StationFrom April
to December
1996

Unit 2 of the
Tsuruga Power
StationJuly 1997

Tokai No.2
Power Station

2001 orbeforeSummary• Negligence of checking the operability of other trains in repairing high-pressure- coreinjection-system main stop valve (HPCI MSV) as it failed to open
Since the main stop valve failed to open with the reactor at rated power, the repair was
performed, but the record of having checked the operability of the alternative emergency corecooling systems, which is required by the Operational Safety Program to continue operation,
could not be confirmed.
• Negligence of checking the operability of other systems in replacing a part (gasket) for
leak-tightness of emergency diesel generator
Although a water leakage occurred at the cooling water system of one emergency diesel
generator, reactor operation continued without checking the integrity of other systems.
• Hiding the corrosion event on the outside surface of condensate storage tank
Although the plate thickness of the tank lower section became less than the required minimum
thickness specified in the application of construction permit due to corrosion, the tank wasused with its water level lowered without checking the required strength.
• Hiding the occurrence of a very small leak of the reactor coolant
When a leak was found in the piping in the containment, the operation should have been
discontinued for repair, but the fact was hidden and the operation was continued without repair
for about 8 months.
• Deceptive action to an equalizing valve during an inspection of containment leakage rate
An official inspection of containment leakage rate was conducted with a block plate installed at
the outlet of an inboard equalizing valve with a leak identified, of the regular airlock, without
following the appropriate in-house procedures. Then, the nuclear reactor was started up withthe equalizing valve replaced, but the local leak rate test was not carried out prior to thestart-up.
• Falsification of flow-rate data during a functional test of the reactor building gas
processing system
Since the airflow did not satisfy a specified flow rate during a functional test of the standby gas
treatment system, the data was falsified with instrument adjustments so as to satisfy the
specified flow rate.

Fig. 6-1 Locations of Nuclear Installations(as of the end of June 2007)
Hokkaido Electric Power Co., Inc.
Tomari Power Station
Units 1 & 2 <PWR> (in operation)
Unit 3 <PWR> (under construction (sub-critical))Tokyo Electric Power Co., Inc.
Kashiwazaki Kariwa Nuclear Power Station
Units 1, 2, 3, 4 & 5 <BWR> (in operation)
Units 6 & 7 <ABWR> (in operation)

6-17

Hokuriku Electric Power Co.
The Japan Atomic Power Company
Shika Nuclear Power Station
Tsuruga Power Station
Unit 1 <BWR>, Unit 2 <ABWR> (in operation)
Unit 1 <BWR> (in operation)
Unit 2 <PWR> (in operation)
Units 3 & 4 <APWR> (in preparation for construction)

Japan Atomic Energy Agency
Monju <FBR>(under construction (attained criticality))

Japan Atomic Energy Agency
Fugen <ATR>(in preparation for decommissioning)
The Kansai Electric Power Co., Inc.
Mihama Power Station
Units 1, 2 & 3 <PWR> (in operation)

The Kansai Electric Power Co., Inc.
Ohi Power Station
Units 1, 2, 3 & 4 <PWR> (in operation)
The Kansai Electric Power Co., Inc.
Takahama Power Station
The Chugoku Electric Power Co., Inc. Units 1, 2, 3 & 4 <PWR> (in operation)
Shimane Nuclear Power Station
Units 1 & 2 <BWR> (in operation)
Unit 3 <ABWR> (under construction (sub-critical))
The Chugoku Electric Power Co.,Inc.
Kaminoseki Nuclear Power Station
Units 1 & 2 <ABWR>(iti ft ti )
Kyushu Electric Power Co., Inc.
Genkai Nuclear Power Station
Units 1, 2, 3 & 4 <PWR> (in operation)
Kyushu Electric Power Co., Inc.
Sendai Nuclear Power Station
Units 1 & 2 <PWR> (in operation)

Electric Power Development Co., Ltd.
Oma Power Station,
Unit 1 <ABWR> (in preparation for construction)
Tohoku Electric Power Co., Inc.
Higashidoori Nuclear Power Station,
Unit 1 <BWR> (in operation))
Tokyo Electric Power Co., Inc.
Higashidoori Nuclear Power Station,
Units 1 & 2 <BWR> (in preparation for construction)
Tohoku Electric Power Co., Inc.Onagawa Nuclear Power Station,
Units 1, 2 & 3 <BWR> (in operation)
Tokyo Electric Power Co., Inc.
Fukushima Daiichi Nuclear Power Station,
Units 1, 2, 3, 4, 5 & 6 <BWR> (in operation)
Tokyo Electric Power Co., Inc.
Fukushima Daini Nuclear Power Station,
Units 1, 2, 3 & 4 <BWR> (in operation)
The Japan Atomic Power Co.
Tokai Power Station <GCR> (under decommissioning)
Tokai No.2 Power Station <BWR> (in operation)
Chubu Electric Power Co., Inc.
Hamaoka Nuclear Power Station
Units 1, 2, 3 & 4 <BWR>, Unit 5 <ABWR> (in operation)

Shikoku Electric Power Co., Inc.
Ikata Power Station
Units 1, 2 & 3 <PWR> (in operation)

A nuclear installation in preparation for construction means one
authorized by the Electric Power Development Working Group (theElectric Power Development Coordination Council up to January 2001) of
the Advisory Committee for Natural Resources and Energy, but has notreceived an approval for its first construction plan.

Attachment

The NSC 2007-D8 (Decided on 23 April 2007)
Actions to take, concerning the malicious conducts of unreported alteration andconcealment at nuclear facilities (Summary)
The NSC received on 20 April 2007 a report from NISA on the “Assessment and follow-up
actions concerning comprehensive checks of power generation facilities (Comprehensive
Inspection Report)” and another report on the “Investigation of the criticality accident in 1999
Unit 1, Shika Nuclear Power Station, Hokuriku Electric Power Company, Inc. and other
unforeseen cases of control rod dislodgement during reactor outages (Criticality Accident
Report).”Repeated malicious conducts of unreported alteration and concealment jeopardize the baseof ensuring nuclear safety. It is really shameful that some serious cases were among them.
Currently nuclear licensees have been in the process of implementing recurrence prevention
measures under the regulatory frames strengthened in 2002 and 2003. The actions were being
taken based upon the lessons uncovered of similar data alterations in the past. Most of
malicious conducts recently uncovered were before these recurrence prevention measures
came into effect. Nevertheless, the recently uncovered cases again hampered the public trustin nuclear energy.
For recovering public trust, the NSC considers it really important to put into effect soonest
the follow-up actions proposed by NISA. Following are the immediate actions by the NSC.
1. Criticality accidents and control rod dislodgement
The criticality accident on Unit 1, Shika Nuclear Power Station, Hokuriku Electric Power
Company, Inc. is really a matter of serious concern in the context that an unforeseen criticality,
which is an emergency (at a nuclear power plant), is secretly covered. The NSC shall examine
closely the unforeseen cases of control rod dislodgement, in view of their risks to fundamental
nuclear safety.
It is to the NSC awareness, meanwhile that, among the cases of control rod dislodgement,
which came into light starting with the criticality accident of Hokuriku Electric Power
Company, there are some cases irrelevant to data alterations.
(1) The ensuring of safety during reactor outages
The Criticality Accident Report (NISA) states, “Important is to examine how to ensure
reactor safety, as part of operation management during outages, including hardware
modifications if necessary, because safety functions of the facility could be temporarily lifted
during periodic inspection outages for the purpose of inspections or tests.”6-18

It is the NSC position that the principle of defense-in-depth should be maintained for safetymeasures and their certain implementation even during the outages.
Therefore, the NSC will conduct necessary examinations, being informed from NISA of its
investigation and examination results of safety measures during outages such as operation
management including hardware modifications as needed, based on the lessons from thecriticality accident at Shika or from abroad.
(2) Criticality accident on Unit 1, Shika Nuclear Power Station, Hokuriku Electric Power
Company, Inc.
The Criticality Accident Report (NISA) also states that “the analysis (by the Hokuriku
Electric Power Company) meets with margin the safety criteria of the abnormal transientsassumed in the (earlier) safety examination process, and, hence, the fuel integrity wasapparently not jeopardized.”Upon NISA instructions, JNES is to perform an “independent (cross) check” of this analysis.
The NSC requests NISA to report to the NSC on its evaluation of the case based on the JNES
analyses.
Besides the regulatory requirements, the NSC plans to interview, at the earliest convenience,
the top management of the operator concerning its determined measures to let such criticaloccurrence never concealed again.
2. Analyses of, and feedback from, the information on accidents and troubles(1) Global sharing of the information on accidents and troubles, and its utilization
The Criticality Accident Report (NISA) also clarifies that: Six unforeseen criticalities havebeen reported at commercial BWR plants, three out of which occurred during the outages; allthese three cases occurred when control rods were maneuvered for insertion or withdrawal, thesituation being different from the recent case of Hokuriku Electric Power Company; and, forthis reason, no significant information could have been withdrawn from these cases forpreventing the recent case of Hokuriku Electric Power Company.
Besides the cases mentioned in the Criticality Accident Report (NISA), the criticality
accident in 1987 during the outage at the Oskarshamn Nuclear Power Station, Sweden,
prompted NRC to draw attention of the US BWR owners.
The NSC must admit that Japan is still insufficient to share the global information on
nuclear accidents and troubles, considering not only the criticality cases but also other cases,
such as the secondary circuit pipe rupture at Unit 2, Surry Nuclear Power Station, U.S. (1986),
or the strainer blockage at Unit 2, Barseback Nuclear Power Station, Sweden (1992).
Needless to say, it is quite important to extract lessons for enhancing nuclear safety of ournational nuclear facilities from the information on nuclear accidents and troubles in other6-19

countries. It is equally important to carefully examine our cases, in disseminating our national
experience to other countries, in terms of their impacts to our national nuclear facilities aswell as their user-friendliness to international recipients.
Therefore, NISA should strengthen its activities on the collection/dissemination and theanalysis/utilization of international information relevant to nuclear accidents and troubles, by
the use of multi-national (IAEA, for instance) or bilateral cooperation frameworks.
The NSC also examines the measures to make use of internationally shared information,such as the IAEA databases, for preventing accident recurrences in our country.
(2) Information sharing among utilities and reactor manufacturers, and its utilization
Important in the utilization of the information relevant to accidents/troubles is for electricutilities and reactor manufacturers to analyze the causes closely and share the recurrence
prevention measures based thereupon. As pointed out in the Criticality Accident Report(NISA) 1 , the electric utilities and reactor manufacturers should, through the association suchas the Japan BWR Owners Group and the Japan PWR Owners Group, acknowledge theimportance of sharing the technological information and its use for accident recurrence
prevention.
With the intention of promoting such activities, the NSC plans to interview, at theappropriate timing, the Japan BWR Owners Group and the Japan PWR Owners Group
concerning their activities for information sharing.
(3) Knowledge base building of reactor operating information by nuclear operators
The information on operation management should be appropriately recorded andaccumulated for analysis and use for enhancing nuclear safety. However, the Comprehensive
Inspections Report (NISA) has revealed a number of cases with defects in this regard.
The NSC takes a position that a station-level knowledge base on the operation management
information should be the generic base condition for recurrence prevention. To this end, anelectronic system at a station-level is an effective means for sharing the operation
management information and for its automatic recording and preservation. The NSC requests
nuclear operators to proactively tackle this issue, although it is not a regulatory requirement.
The system would also contribute to the prevention of malicious conducts such as dataalteration.
3. Advancement of revising the legally bound inspection regime
1

The Japan BWR Owners Group is tackling the issue of information sharing, withparticipation of the electric utilities and the reactor manufacturers. Such activities should bevigorously and effectively exercised.
6-20

NISA has been examining, at its “Committee on the Inspection System,” a more scientificand rationalized inspection regime with due consideration to plant specific maintenanceprograms.
The NSC takes a position that the lessons from the Comprehensive Inspections Report(NISA) should be aggressively reflected in this revision process and, therefore, theCommittee’s examination be advanced so that the operators’ safety measures be further
promoted. The NSC expects to be informed, at the appropriate timing, of the concrete
approach of NISA for advancing the revision process.
4. The NSC plans for motivating field-oriented approaches by means of site interviews ofReactor Chief Engineers
The reactor chief engineers are, as prescribed in the Reactors Regulation Law, responsible
for supervising the plant safety measures. They are due to play crucial roles in ensuring
reactor safety, with full use of their professionally technological knowledge and experience.
The NSC plans to promote placing higher importance to the plant site, or the very front lineof ensuring safety. To this end, the NSC plans to share the problem awareness with reactor
chief engineers through direct interviews at the earliest convenience and also encourage
mutual information exchange between them.
5. NISA Plan of reviewing the revision of operator’s safety rules and of special safety
inspections
NISA plans to execute special programs for nine units at seven nuclear power stations,which have been ranked as “I” in its Comprehensive Inspections Report. The NISA special
programs: require the revision of safety rules to prescribe the reporting system, upon major
accidents, to the top management for recurrence prevention; plan special inspections at theearliest periodic inspection; execute special supervision by special nuclear facility
superintendents; and implement special safety inspections.
The NSC is to receive reports from NISA as appropriate and watch the development.
6. Subsequent Regulation Reviews of NISA’s follow-up actions
The NSC plans Subsequent Regulation Reviews for the regulation-related items from among
NISA’s follow-up actions, in order to enhance nuclear safety.

6-21

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6-22

B. Legislation and Regulation

This page intentionally left blankArticle 7 Legislative and Regulatory Framework
1. Each Contracting Party shall establish and maintain a legislative and regulatory
framework to govern the safety of nuclear installations.
2. The legislative and regulatory framework shall provide for:(i) the establishment of applicable national safety requirements and regulations;(ii) a system of licensing with regard to nuclear installations and the prohibitionof the operation of a nuclear installation without a licence;(iii) a system of regulatory inspection and assessment of nuclear installations toascertain compliance with applicable regulations and the terms of licenses;(iv) the enforcement of applicable regulations and of the terms of licences,including suspension, modification or revocation.

The Atomic Energy Basic Law has been established as the basic law governing the utilizationof nuclear energy, and the Law on the Regulation of Nuclear Source Material, Nuclear Fuel
Material and Reactors (hereinafter referred to as the “Reactor Regulation Law”), theElectricity Utilities Industry Law etc. have been established as the laws to govern the safety of
nuclear installations, which are subjects to the convention.
As a progress after the previous report, the technical standard in accordance with theElectricity Utilities Industry Law has been revised to define performance requirements, andthe detailed specifications conforming to the performance requirements were defined bystandards of the private sectors. And the Regulatory Guide for Reviewing Seismic Design ofNuclear Power Reactor Facilities etc. (hereinafter referred to as the “Regulatory Guide for
Reviewing Seismic Design”) has been revised, incorporating the latest knowledge into thetechnical requirements. The issues which are now under review and will be reflected to futurerevision or establishment of legislations and regulations are the new inspection systemcurrently being reviewed by “The Taskforce on Inspection System” (see Article 19) and theuse of risk information in legislations (see Article 14).
7.1 System of Legislations Governing the Nuclear installations(1) Legislations concerning Utilization of Nuclear Energy in General
The overall system of legislations and regulations for utilization of nuclear energy is based on
the Atomic Energy Basic Law. The objectives of the Atomic Energy Basic Law are quoted as"to secure future energy resources, achieve progress in science and technology, and promoteindustry, by encouraging research, development, and utilization of nuclear energy, and thereby
contribute to improvement of the welfare of human society and the people's living standard."
The basic policy is prescribed as follows: "the research, development and utilization of
nuclear energy shall be limited to peaceful purposes, on the basis of the highest priority ofensuring safety, and performed on a self-controlled basis under democratic administration, andthe results obtained shall be made public and actively contribute to international cooperation."

7-1

In order to achieve these objectives and the basic policy, the law provides establishment of aset of laws to govern following areas:
・ Establishment of the Atomic Energy Commission and the Nuclear Safety Commission, andtheir duties, organization, administration, and authorities.
・ Regulations governing nuclear fuel materials and nuclear source materials.
・ Regulations for construction, etc. of nuclear reactors.
・ Prevention of radiation hazards.
・ Compensation for a nuclear damage
The law also provides that those who will utilize nuclear energy shall manage their facilities
with the first priority on safety under the supervision of the regulatory body in accordancewith these laws.
The basis to establish organizations related to regulation and the missions of theorganizations are provided in the laws, such as the "Law for Establishment of the Atomic
Energy Commission and the Nuclear Safety Commission", "Law for Establishment of the
Ministry of Economy, Trade and Industry", "Law for the Japan Nuclear Energy Safety
Organization, Incorporated Administrative Agency."(2) System of Legislations and Regulations Governing the Safety of Nuclear Installations,
Subjects of the Convention on Nuclear Safety
Major legislations and regulations for safety regulation of nuclear installations are shown in
Figure 7-1. As shown in the figure, 1) the Law on the Regulation of Nuclear Source Material,
Nuclear Fuel Material and Reactors, 2) the Electricity Utilities Industry Law, 3) the Basic
Law for Emergency Preparedness and the Special Law of Emergency Preparedness for
Nuclear Disaster, 4) The Laws for Radiation Protection., 5) the Environment Impact
Assessment Law, and 6) the Law on Compensation for Nuclear Damages have been enacted as
major legislations and regulations.
These laws provide requirements for approval, inspection and notification of licensing,construction, operation and decommissioning of establishment of nuclear installations. They
are enacted after deliberations at the Congress, which means the revision also requires theresolution at the Congress.
The nuclear installations in the scope of this Convention are any land-based civil nuclear
power plants. They are using nuclear fuels and are regulated by both laws of safety
regulations for reactors using nuclear fuel (Reactor Regulation Law) and safety regulations
for commercial electric power generation (Electricity Utilities Industry Law). However, theregulations in accordance with the both laws are ruled not to overlap for the same matter.
The authorization is entrusted to regulatory body by in a law ,and the ordinances, in the
7-2

hierarchy under the law, describing the procedures for approval, inspection and notification,
are enacted or revise, by the competent authorities, after obtaining the decision by the
Cabinet. In the hierarchy under the ordinance, there are the rules which the competentauthorities can establish, in accordance with the authorization by the law and the ordinance,
the to define the details of application, the basis for approval, the technical requirements,
the control items for radiation protection, the licensee’s measures for safe operation, etc forapproval, inspection and notification of various matters. And the competent authorities of thelegislations and regulations can establish the technical standards (notice) and the guidelinesof detailed technical requirements in accordance with these legislations and regulations.
(3) Outlines of Major Laws and Regulations related to the Safety Regulation for Nuclear
Installations
Outlines of each major law and regulation are described as follows:
1) Law on the Regulations of Nuclear Source Material, Nuclear Fuel Material and Reactors(Reactor Regulation Law)
The Reactor Regulation Law, in order “to ensure that the utilization of nuclear source
material, nuclear fuel material, and reactors are limited to peaceful purposes, and carriedout in a planned manner, and to ensure safety of the public by preventing the hazards dueto these utilization and providing physical protection of nuclear fuel material, in
accordance with the spirit of the Atomic Energy Basic Law”, provides prescriptions on
establishment and operation etc. of reactors.
And the following matters are established for reactor facilities
・ Regulations on the basic design and policies of reactor facilities for establishment ofreactor facilities,
・ Regulations on the detailed design for construction of reactor facilities (approval ofdesign and construction methods),
・ Inspections at the time of construction of reactor facilities (approval of the welding
method, welding inspection and pre-service inspection),・ Regulations for operation of reactor facilities (approval of the Operational Safety
Program and the Operational Safety Inspection) (inspection of the observance statusof the Operational Safety Program), notification of operation plan),・ Inspection for operation of reactor facilities (Periodic Inspections of reactor
facilities),
・ Safe operation of reactor facilities and measures to be taken for protection of specificnuclear fuel materials,
・ Regulations on transfer, succession or merger of reactor facilities, and
7-3

・ Dismantling of reactor facilities.
The allegation system was established so that personnel of licensees can allege violationof Reactor Regulation Law to the Minister of Economy, Trade and Industry or the
Nuclear Safety Commission (the NSC), without unfavorable treatment. The systemprovides the rule to handle the allegation, such as for protection of personal data ofallegers and for appropriate implementation of the procedure for investigation anddisclosure of an alleged case. By the end of March, 2007, Ministry of Economy, Trade andIndustry( METI) investigated 33 alleged cases, and the NSC investigated 6 alleged cases.
Licensee’s violation of legislations and regulations can be discovered early by theallegation system and nuclear disaster is expected to be prevented.
The Minister of METI is required to report quarterly the status of regulatory activities forapproval of the Operational Safety Program and their change to the NSC.
For commercial power reactors, the provisions of Electricity Utilities Industry Law areapplied for regulations on the detailed design for construction of reactor facilities (approvalof design and construction methods), inspections at the time of construction of reactor
facilities (approval of the welding method, welding inspection and pre-service inspection),and inspection after commissioning (facility Periodic Inspection), and it is prescribed in theReactor Regulation Law that the corresponding provisions of the Reactor Regulation Law
are exempted from application.

2) Electricity Utilities Industry Law
The Electricity Utilities Industry Law was established to ensure safety of public and topreserve environment by regulating construction, maintenance and operation of electricfacilities, providing safety regulations for that purpose.
In view of ensuring safety of electric facilities used for electric utilities industry, theprovisions on the Approval of Construction Plan, Fuel Assembly Inspection, Audit ofLicensee’s Welding Check System, Pre-service Inspection, Periodic Licensee's Check,
Audit of Licensee's Periodic Check System, Periodic Inspection, Operational Safety
Program for Nuclear Facilities are defined in the Electric Utilities Industry Law.
The Minister of METI is required to make quarterly reports to the NSC on the status of
regulatory activities, such as Approval of Construction Plan, Pre-service Inspection, Fuel
Assembly Inspection, Periodic Inspection, and Audit of Licensee's Periodic Check
System.
3) Basic Law for Emergency Preparedness and Special Law for Nuclear Emergency
The nuclear emergency had been addressed within the legal framework of the Basic Law
for Emergency Preparedness. Taking account of the special characteristics of a nuclear
emergency, the Special Law for Nuclear Emergency was established in December 1999.

7-4

The Law stipulates special measures for nuclear emergency, including licensee’s
obligation for preventing nuclear emergency, the Declaration of Nuclear Emergency, andestablishment of the Nuclear Emergency Headquarters, as well as activation of emergency
measures in nuclear emergency. It also stipulates that Senior Specialists for Nuclear
Emergency be stationed in the vicinities of nuclear installations, which guides andadvises licensees in preparing preventive measures for nuclear emergency, and conducts
other activities necessary to prevent the occurrence and progression of a nuclear
emergency.
The volume for nuclear emergency preparedness in the Basic Plan for Emergency
Preparedness in accordance with the Basic Law for Emergency Preparedness, clarifies themeasures to be activated at each step of the occurrence of an abnormal event, progression
into a nuclear emergency, and recovery from the emergency.
4) Laws for Radiation Protection
The radiation protection at nuclear installations is regulated by the Reactor Regulation
Law, the Electricity Utilities Industry Law and the Industrial Safety and Health Law.
The Reactor Regulation Law stipulates zone control for radiation protection, dose controlof personnel engaged in radiation work, measurement and monitoring of radiation levels,
etc. in order to protect personnel and the public. The Electricity Utilities Industry Law
prescribes the radiation instrument devices to be installed in nuclear installations. The
Industrial Safety and Health Law defines the dose limits of personnel engaged inradiation work, which are equivalent to the Reactor Regulation Law. In accordance with
the Law for Technical Standards of Radiation Hazards Prevention, the Radiation
Council was established to take a consistency among technical standards for radiation
hazards.
In order to prevent hazards due to the use of radioisotopes at nuclear installations, the
Law Concerning Prevention from Radiation Hazards to Radioisotopes, etc.(hereinafterreferred to as “Radiation Hazard Prevention Law”) stipulates zone control of radiation
protection, dose control of personnel engaged in radiation work and radiation
measurement in controlled area etc.
Relevant legislations were revised incorporating the ICRP Recommendation 1990 and
enforced in April 2001.
5) Environment Impact Assessment Law
The Environmental Impact Assessment Law was enacted in June 1999, replacing theDecision of the MITI Departmental Council, July 1977, which stipulated theenvironmental impact assessment of nuclear installation other than safety assessment. The
environmental impact assessment is implemented in accordance with the law.
The objective of the Environmental Impact Assessment Law is for licensees to perform

7-5

proper assessment of a large business plan which may pose large impact on theenvironment, and to prepare proper plan to preserve the environment. The law provides aset of procedures for it. Environmental assessment on commercial power plants, including
nuclear installations, is performed in accordance with the provisions of the Environment
Impact Assessment Law and the corresponding provisions of the Electricity Utilities
Industry Law. The environmental impact assessment is obligatory for nuclear installation
regardless of its scale.
6) Law on Compensation for Nuclear Damage
The Law on Compensation for Nuclear Damage establishes the basic system oncompensation for nuclear damage caused by a nuclear accident.
The Law adopts the “liability without fault” principle and imposes sole liability ofcompensation for nuclear damage on licensees, exempting claimants from proving
licensee’s fault in accordance with the general principle of the Civil Law. Also, infinite
liability of compensation is imposed on the licensee.
To secure the fund of and to facilitate the compensation, the licensee is required to makethe Financial Arrangement for Nuclear Damage Liability. The amount of the Arrangement
is sixty billion yen for a nuclear installation in general.
The Arrangement consists of the Nuclear Damage Liability Insurance Contract with acivil insurer and the Indemnity Agreement for Compensation with the national
government. The latter supplements the former in the case of large-scale accident such as
caused by earthquake or volcanic eruption.
And in case the total amount arranged by the licensee is not sufficient for fullcompensation, the national government, on the basis of decision by the Diet, would aid tocover the licensee. In the case of enormous natural disaster or social disturbance, thenational government bears the compensation, exempting licensee from liability forcompensation.
(4) Provisions on Technical Requirements in the Safety Regulations
Technical requirements that will be used for review of basic design of nuclear installations areprovided in the Guides for safety design and safety assessment established by the Nuclear
Safety Commission. The system of the Guides is shown in Table 7-1. These Guides are used toassess the adequacy of licensee’s application for the license for establishment at the safety
review and assessment of the application.
For technical requirements necessary for regulations at the subsequent stage such as approvalof construction plan and the pre-service inspection after approval of establishment of a
nuclear installation and the Periodic Inspection after commissioning, the Nuclear andIndustrial Safety Agency (NISA) has established the technical standards as performancerequirements. This system of technical standards is shown in Figure 7-2.

7-6

Furthermore, NISA uses the standards of the academic societies and associations that are
standards of the private sectors, as the specification requirements to realize the performancerequirements, after doing the technical evaluation. Major standards of academic societies andassociations are shown in Table 7-2.
NISA also defines the internal regulatory guides, such as guidelines of various evaluations
and review procedures.
As revisions of guides carried out after the previous report, the revised Regulatory Guide for
Reviewing Seismic Design was issued in September, 2006 aiming at further improvement ofseismic safety and reliability of nuclear installations (see Article 18). In order to verify theseismic safety of existing nuclear installations based on the revision, licensees were required
to implement the seismic safety evaluation referring to the revised Regulatory Guide for
Reviewing Seismic Design of Nuclear Power Reactor Facilities.
As revisions of technical standards carried out after the previous report, the technicalstandards that included specification requirements was revised to the ones that provide
performance requirements only, and standards of academic societies and associations areused for providing specifications after evaluating the standards technically.
7.2 Legislative and Regulatory Framework at Each StageThe overview of safety regulations on the basis of the Reactor Regulation Law, the Electricity
Utilities Industry Law, etc. from planning stage through operation stage is shown in Figure 7-3.
A summary of safety regulations for a commercial power reactor is stated in this section.
(1) Planning Stage
When selecting a site for a nuclear installation, the electric utility, on the basis of theEnvironmental Impact Assessment Law and the Electricity Utilities Industry Law, performs
environmental impact assessment, and submits to METI the draft Environmental Impact
Statement (draft EIS) explaining current status of the environment and measures to protect it.
The draft EIS is sent to the related local governments to be disclosed for public comments.
The utility prepares their views addressing residents’ comments. Assessments on air, water,and soil pollution due to radioactive substances are performed under the Reactor Regulation
Law and exempted from application of the Environmental Impact Assessment Law.
METI conducts the evaluation, soliciting experts’ opinion.
In order to have opportunities to invite the opinions from the general public widely on various
issues concerning establishment of the installation, the first public hearings with explanationof the electric power company are held by METI to obtain deeper understanding andcooperation of residents in the vicinity. The results of public hearings are taken into
consideration in the safety examination..

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(2) Establishment Stage
The license applicant, having completed the procedure of planning stage, submit application
format for a license for establishment to the Minister of METI in accordance with the Reactor
Regulation Law. Applicants attach documents to the application format including a description
on safety design of the nuclear installation, radiation control, and accidents and failures.
NISA conducts an examination to determine the adequacy of the site, and the basic design ofstructure and equipment from the points of prevention of radiological hazards, focusing on theevaluation of the safety of the reactor core and the radiation exposure due to establishment of
the nuclear installation. In addition, the regulatory body confirms that the nuclear installation
should be used for peaceful purpose and in line with the planned development and utilizationof nuclear energy, and the applicant has sufficient technical capability to ensure safety andsufficient financial basis to execute the plan.
In this examination, the regulatory guides in Table 7-1 and other documents established by
the NSC are used. In the examination, site surveys, and analysis are conducted, whennecessary.
The Minister of METI consults with the Atomic Energy Commission and the NSC on the
results of its examination. During the review process of METI’s results, the NSC reviews
independently focusing on safety problems specific to the installation, and gives its views to
the Minister of METI. The Minister of METI considers these views, asks for the consent of the
Minister of MEXT, and then issues the license.
At the establishment stage, the second public hearing is held by the NSC to hear the opinionsof residents in the vicinity on the safety specific to the facilities and take the opinions into
consideration at the time when the NSC investigate and review the result of safety review andassessment by METI for the application of reactor establishment or alteration license appliedby the electric power company. At the second public hearing, METI will explain the overview
of safety review and assessment and present the view on the stated opinions. And NISA is
conducting public relations activities positively to residents as provided in 8.3 (4) of thisreport.
In addition to these activities, Nuclear and Industrial Safety Subcommittee of the Advisory
Committee for Natural Resources and Energy of METI invites opinions (Public Comments)
when they formulate or alter fundamental policies, or newly introduce or alter the system
which might affect the rights and duties of the people. When the NSC makes important policy
decision or when Committee on Examination of Reactor Safety investigates and reviews thesafety review and assessment, its executive office opens the contents of issue items for a fixed
period of time to the public, and invites the opinion from the general public.
(3) Construction StageIn accordance with the Electricity Utilities Industry Law, the licensee submits theConstruction Plan for establishment of electric facilities, and obtains an approval of the

7-8Minister of METI before starting construction works. NISA examines the Construction Plan toconfirm that the detailed design of electric facilities is consistent with the basic design anddesign policies approved at the stage of licensing for establishment, and is in conformity withthe technical standards based on the Electricity Utilities Industry Law. And the licensee mustdesignate Chief Electrical Engineers and Chief Engineers of Boiler and Turbine and notify
NISA of it.
After obtaining an approval or notification of the Construction Plan, the licensee shallundergo the Pre-Service Inspection by NISA at each process of construction and at thecompletion of all construction works, which confirms that construction is conducted inaccordance with the construction plan and is in conformity with the technical standards. The
licensee shall obtain design approval by NISA for fuel assemblies to be loaded in the reactor
and undergo the Fuel Assembly Inspection conducted by NISA. The licensee shall perform
Licensee’s Welding Check for welding of pressurized parts and containment vessels and shallundergo review by JNES on the organization conducting the inspection, the inspection method,schedule control, and other items provided by the ordinance of METI (Audit of Licensee's
Welding Check System).
(4) Operation StageAt the start of operation, the licensee must notify NISA of the operation plan, obtain an
approval of the Operational Safety Program that prescribes procedures of operational
management, operational limits and safety education of personnel, designate Chief Reactor
Engineers, Chief Electrical Engineers and Chief Engineers of Boiler and Turbine, who
supervise the safety of the operation, and the Persons Responsible for Operation, and notify
NISA of them. The licensee is required to notify NISA of the operation plan annually.
The 17 items prescribed in the Operational Safety Program are provided in Reactor Regulation
Law, which includes the Periodic Assessment, quality assurance, maintenance management,
etc. of nuclear installations
The licensee must control the radiation exposure of personnel engaged in radiation work sothat their doses do not exceed the dose limit, and shall report the exposure dose of personnel
to NISA periodically.
Licensees must discharge gaseous and liquid radioactive waste generated during operation
into the environment, in compliance with the concentration values which are lower than theconcentration limits stipulated in the Reactor Regulation Law. Licensees must make efforts to
reduce discharge amount as small as possible so that annual public exposure in the vicinity
will be kept below 50 μ Sievert in accordance with the Regulatory Guide for the Annual Dose
Target for the Public in the Vicinity of Light Water Nuclear Power Reactor Facilities(hereinafter referred to, as the “Regulatory Guide for Annual Dose Target”).
After starting operation, the licensees, in accordance with the Electricity Utilities Industry
Law, shall perform the Periodic Licensee’s Check to confirm that the installations conform to
the technical standards. The records of the Periodic Licensee’s Check have to be stored by the
7-9licensee for five years after decommissioning of the electric facilities. And fitness-for-service
assessment is required to important part such as reactor pressure boundary. And also, thelicensee must undergo the Periodic Inspection by NISA on the specified part of structures
important to safety. The Periodic Inspection and the Periodic Licensee’s Check are conductedduring shutdown of operation within the interval not exceeding 13 months from the date ofstart of operation or the date of completion of the previous inspection. Since October 2003,
JNES has conducted part of the Periodic Inspection and notifies NISA of the results on thebasis of the revision of the relevant laws. Also, licensees must undergo the Periodic Safety
Management Review, in which JNES reviews the licensee’s organization conducting the
Periodic Licensee’s Check, inspection method, schedule control, and other items provided in
the ordinance of METI, and report the result to NISA for the evaluation. NISA shall evaluate
the result comprehensively.
Licensees must undergo the Operational Safety Inspection by the Nuclear Safety Inspectors on
the observance of the Operational Safety Program, including the licensee’s organization,
quality assurance, maintenance management, operation, maintenance and repair of component,
surveillance, radiation control, and management of radioactive wastes, discharge control ofgaseous and liquid radioactive wastes, monitoring, and safety education. NISA conducts theOn-site Inspection of nuclear installation to confirm compliance with safety regulation, if
necessary.
If any failure occurs in a nuclear installation, the licensee must report the failure etc.
immediately to NISA in accordance with the provisions of the Reactor Regulation Law and theElectricity Utilities Industry Law, and shall report to NISA, without delay, on the situation of
the failure and the measures taken. In order to improve transparency of information to the
public, the reporting criteria for failures etc. were more clearly defined by amending theReactor Regulation Law in October 2003. Licensees have established the system to collectinformation on events, including minor events that are outside of the reporting criteria, anddisclose them to the public.
Criteria for necessity of approval or notification of the Construction Plan for any modification
or repair work of electric facilities after startup of operation was clarified by the amendmentof the Rules for the Electricity Utilities Industry Law in October 2003. NISA established the“Regulatory Guide on the Construction Plan" to identify the details of the amendment andnotified licensees of it.
MITI (present METI) issued, in 1992, a Decision of the METI Departmental Council torequest licensees to voluntarily perform the periodic safety review at a regular operatinginterval (approximately every ten years), including incorporation of operating experiencesfrom commissioning to date and the latest technological knowledge, and probabilistic safetyassessment. On the basis of the amendment of the Reactor Regulation Law, in October 2003,
the periodic safety review at a regular interval (approximately every ten years) was
incorporated into the Operational Safety Program, the observance of which the Nuclear Safety
Inspector inspects at the Operational Safety Inspection. In December 2005, NISA decided toinclude review with respect to enhancement of Measures for Aging Management and

7-10

degradation of licensee’s organizational climate. The implementation of probabilistic safetyassessment, however, remains to be a voluntary activity of licensees as yet.
On the basis of the amendment of the Reactor Regulation Law, in October 2003, licensees
were obliged to perform technical evaluation on ageing of nuclear installation before
continuous operation more than thirty years and must prepare a ten-year maintenance plan
based on the technical evaluation. The subsequent evaluation should follow within ten years.
In December 2005, NISA issued the Performance Guidelines for Measures for Aging
Management and the Standard Review Procedures for Measures for Aging Management tocomplete measures for aging management.
7.3 Record, Applicable Regulations and Enforcement of Terms of License
Licensees are obliged to record and save required items for operation and use of reactor for
every nuclear reactor. The concrete items and the period are provided in the Rules on
Establishment and Operation of Commercial Power Reactor.
In accordance with the Reactor Regulation Law, the Minister of METI may revoke the license
for establishment or issue a Shutdown Order of nuclear installation for up to one year, undercircumstances such as operating a nuclear installation without a license for establishment,
violating an order legally issued by NISA, failing to implement measures necessary for safety
prescribed by NISA, or failing to obtain approval for the Operational Safety Program.
The Reactor Regulation Law also prescribes imprisonment and/or fines under circumstancessuch as establishing a nuclear installation without a license for establishment, violating a
Shut-Down Order, or failing to take relevant emergency measures, which are prohibited by theLaw. NISA may order changes in the Operational Safety Program whenever it is deemednecessary for preventing potential radiological hazards. Licensees failing to abide by suchorders would be punished with a fine.
In accordance with the Electricity Utilities Industry Law, if it is judged for an electric facility
not to conform to the technical standards, the Minister of METI may order repair, alteration,
relocation, temporary suspension of usage, or limitation of usage.
The Electricity Utilities Industry Law prescribes fines if a licensee violates a technicalstandard order for conformity, or establishes or alters an electric facility without obtaining
necessary approval for a construction plan, or uses an electric facility without undergoing or
passing the Pre-Service Inspection or the Fuel Assembly Inspection, or fail to receive the
Audit of Licensee ’ s Welding Check System without performing Licensee's Welding Check. It
also prescribes to revoke the business license, if an electric utility violates the law or orders
based on the law causing serious damage to the public benefits.
On the basis of the amendment of the Reactor Regulation Law in October 2003, when an
employee violates a law and is punished by a fine, the legal person who legally employs him
or her is also punished by a fine as heavy as 100 times of the employee’s fine, to prevent
organizational illegal acts.

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7.4 Change the Technical Standards for Nuclear Installations to Performance Requirements
The technical standard based on the Electricity Utilities Industry Law defines the technicalrequirements for nuclear installations. The previous technical standard included not only theperformance requirements but also detailed specifications. The technical standard has beenrevised so that it only provides performance requirements which are necessary for safety andthe detailed specifications for conforming to these performance requirements are to be definedby using standards of academic societies and associations..
The background, present status, basic policies to use, conditions as regulatory standards andmethods of verification to change the technical standards including only performancerequirements and the use of standards of academic societies and associations are as follows:・ Background
The Japanese previous technical standards included detailed and concrete "specification
requirements" on structures, materials and dimensions. Though those standards had
advantages that the requirements are clear and judgment of success or failure can be madeclearly and fairly, but, on the other hand, had disadvantages that the flexible responses totechnological innovation or latest knowledge was not easy. This disadvantages will besolved, if the regulatory body defines “performance requirements”, describing only
objectives or functions of safety equipments or facilities to achieve safety levels to beconsidered as necessary by the regulatory body, and detail specifications conforming to theperformance requirements are defined separately by using “detailed specifications”
established by specialists of each field by collecting their knowledge. In recent years, it has
been socially demanded that the technical standards shall be defined as “performancerequirements” as much as possible in order to reflect promptly domestic and overseas
operating experiences, latest knowledge etc. into “detailed specifications”
・ Present status
In Japan, there are standards of private sectors which have been established independently
by industries such as guidelines and rules of the Japan Electric Association (JEAC, JEAGetc.), and most of them have been used by licensees. However, they were not included in theregulatory standards officially and remained as reference information. These standards
have been regarded as "independent standards of private sectors" established by industries.
Recently, academic societies and industrial associations, such as the Japan Society of
Mechanical Engineers, the Atomic Energy Society of Japan as well as the Japan Electric
Association establish the standards of private sectors with emphasis on fairness, justice andopenness taking a process to be reviewed by neutral and fair members in the public. And
these standards are decided to be used officially, calling “standards of academic societies
and associations.”
・ Basic policies to use the standards of academic societies and associations by NISA

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(1) Regulatory standards specify performance requirements, and standards of academic
societies and associations are positively used to establish specifications to realize theperformance requirements.

(2) When the standards of academic societies and associations used by licensees areproved to be the standards which conform to the performances required by NISA, NISA
shall open the fact after technical evaluation of the Standards.

(3) When applicable standards of academic societies and associations do not exist, NISA
encourages its establishment, and until establishment of the standard, the performance,that will be realized by the fulfilling the conventional specifications, is regarded as theperformance standards required by regulatory standards.
・ Conditions of standards of academic societies and associations as regulatory standards
and its verification
When the standards of academic societies and associations are confirmed to fulfill thefollowing conditions, it is regarded that the regulatory standards are satisfied;(1) It corresponds to the items representing performances required by the regulatory
standards,(2) Concrete methods or specifications on technical matters necessary to achieve
performances required by the regulatory standards are provided, and(3) Technical adequacy has been proven for the concrete methods or specifications provided
in the standards of academic societies and associations.
No matter what organization of academic societies and associations have established thestandards ,they are judged to satisfy the performance required by the regulatory standards,
if it can be confirmed that they have been established by the process with emphasis on
fairness, justice and openness and that they fulfill the above three conditions. The
regulatory body make judgment promptly, in order to make the regulation effective andefficient with respect for the technical knowledge of specialists participated in thedevelopment process.
・ Present status regarding use of standards of academic societies and associations andfuture policy
An amendment of the Ordinance of Establishing Technical Standards for Nuclear Power
Generation Equipment (Ordinance of METI, No. 62, 1965), which is for the purpose to
execute the above mentioned policies, was promulgated on July 1, 2005 and enforced on
January 1, 2006 after announcement to the World Trade Organization (WTO) and invitationof public comments.
In practicing the above ministerial order, when corresponding standards of academic

7-13

societies and associations exist, NISA clearly specifies the application of them in the“Interpretation” of the Technical Standards, Ordinance No. 62, after technical evaluation,
and when corresponding standards of academic societies and associations do not exist,
the items which NISA requires are shown in the “Interpretation”.
As of March 31, 2007, 21 standards of academic societies and associations were evaluatedtechnical adequacy by NISA.
NISA will perform further technical evaluation of standards of academic societies andassociations whenever they are prepared.

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Table 7-1 Major Regulatory Guides Specified by the NSC for
Power Generating Light Water Reactors
SitingDesignHazards Prevention
SafetyEvaluation

Dose Target

Technical Competence・ Regulatory Guide for Reviewing Nuclear Reactor Siting
Evaluation and Application Criteria
・ Regulatory Guide for Reviewing Safety Design of Light Water
Nuclear Power Reactor Facilities
・ Regulatory Guide for Reviewing Classification of Importanceof Safety Functions for Light Water Nuclear Power Reactor
Facilities
・ Regulatory Guide for Reviewing Seismic Design of Nuclear
Power Reactor Facilities
・ Regulatory Guide for Reviewing Fire Protection of Light Water
Nuclear Power Reactor Facilities
・ Regulatory Guide for Reviewing Radiation Monitoring in
Accidents of Light Water Nuclear Power Reactor Facilities
・ Fundamental Policy to be Considered in Reviewing of Liquid
Radioactive Waste Treatment Facilities
・ Regulatory Guide for Reviewing Safety Assessment of Light
Water Nuclear Power Reactor Facilities
・ Regulatory Guide for Evaluating Core Thermal Design ofPressurized Water Cooled Nuclear Power Reactors
・ Regulatory Guide for Evaluating Emergency Core Cooling
System Performance of Light Water Power Reactors
・ Regulatory Guide for Evaluating Reactivity Insertion Events ofLight Water Nuclear Power Reactor Facilities
・ Regulatory Guide for Evaluating Dynamic Loads on BWR
MARK-I Containment Pressure Suppression Systems
・ Regulatory Guide for Dynamic Loads on BWR MARK-II
Containment Pressure Suppression Systems
・ Regulatory Guide for Meteorological Observation for Safety
Analysis of Nuclear Power Reactor Facilities
・ Regulatory Guide for the Annual Dose Target for the Public in
the Vicinity of Light Water Nuclear Power Reactor Facilities
・ Regulatory Guide for Evaluating the Annual Dose Target for
the Public in the Vicinity of Light Water Nuclear Power
Reactor Facilities
・ Guide for Radiation Monitoring of Effluent Released from
Light Water Nuclear Power Reactor Facilities
・ Regulatory Guide for Reviewing Technical Competence of
Nuclear Operators

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Table 7-2-1 Standards etc. of Academic Societies and Associations(Guidelines and Rules of the Japan Electric Association)
NumberJEAC 4111-2003 (*)
JEAC 4121-2005

Title
Rules of Quality Assurance for Safety of Nuclear Power Plants
Guidelines for Application of Rules of Quality Assurance for Safety of
Nuclear Power Plants (JEAC 4111-2003)- Operation Stage of Nuclear
Power PlantsJEAC 4201-2004
Method of Surveillance Tests for Structural Material of Nuclear
Reactors
JEAC 4202-2004
Drop-Test Method for Ferritic SteelJEAC 4203-2004 (*) Primary Reactor Containment Vessel Leakage Testing
JEAC 4205-2000
In-service Inspection of Light Water Cooled Nuclear Power
Plant Components
JEAC 4206-2004
Methods of Verification Tests of the Fracture Toughness for Nuclear
Power Plant Components
JEAC 4209-2003 (*) Rules of Maintenance Management of Nuclear Power Plants
JEAC 4602-2004 (*) Definitions of Nuclear Reactor Coolant Pressure Boundary and Reactor
Containment Boundary
JEAC 4605-2004 (*) Definitions of Engineered Safety Features and Related Systems of Nuclear
Power Plants
JEAG 4101-2000
Guideline of Quality Assurance for Nuclear Power Plants
JEAG 4102-1996
Guideline of Emergency Measures for Nuclear Power Plants
JEAG 4204-2003
Guideline for Quality Control of Nuclear Fuel for Nuclear Power Plants
JEAG 4207-2004 (*) Ultrasonic Examination Guideline for In-service Inspection of Light
Water Cooled Nuclear Power Plant Components
JEAG 4208-2005
Eddy Current Test Guideline for In-service Inspections of SteamGenerator
Heat Transfer Tubes for Light Water Type Nuclear Power Plants
JEAG 4601-1987 (*) Technical Guidelines for Aseismic Design of Nuclear Power Plants
JEAG
Technical Guidelines for Aseismic Design of Nuclear Power Plants:
4601-S-1984 (*)
Classification and Allowable Stress
JEAG 4601-1991 (*) Technical Guidelines for Aseismic Design of Nuclear Power Plants:
Supplement
JEAG 4603-1992
Guideline for Design of Emergency Electric Power Supply Systems
for Nuclear Power Plants
JEAG 4604-1993
Guideline for Design of Plant Protection Systems for Nuclear
Power Plants
JEAG 4606-2003
Guideline for Radiation Monitoring for Nuclear Power Plants
JEAG 4607-1999 (*) Guideline for Fire Protection of Nuclear Power Plants
JEAG 4608-1998
Lightning Protection Guidelines for Nuclear Power Plants
JEAG 4609-1999
Application Criteria for Programmable Digital Computer System
in Safety–Related System of Nuclear Power Plants
JEAG 4610-2003
Personal Dose Monitoring for Nuclear Power Plants

7-16

JEAG 4611-1991

Guideline for Design of Instrumentation & Control Equipment with Safety
Functions
JEAG 4612-1998
Guideline for Safety Grade Classification of Electrical and MechanicalEquipment with Safety Functions
JEAG 4613-1998
Technical Guidelines for Protection Design against Postulated Piping
Failures in Nuclear Power Plants
JEAG 4614-2000
Technical Guidelines on Seismic Base Isolation System for Structural
Safety and Design of Nuclear Power Plants
JEAG 4615-2003 (*) Guideline for Design of Radiation Shielding for Nuclear Power Plants
JEAG 4616-2003
Technical Guideline for Design of Base Structures for Dry Cask Storage
Buildings
JEAG 4617-2005
Guideline for Development and Design of Computerized Human-MachineInterface in the Central Control RoomJEAG 4618-2005
Technical Guidelines for Aseismic Design of Steel Plate Concrete
Structures – for Buildings and Structures
JEAG 4801-1995
Guideline for Operating Manual of Nuclear Power Plants
JEAG 4802-2002
Guideline for Education and Training for Nuclear power Plant Operator
JEAG 4803-1999
Guideline for Operational Safety Preservation of Light Water Cooled
Reactors

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Table 7-2-2 Standards etc. of Academic Societies and Associations(Guidelines and Rules of the Japan Society of Mechanical Engineers)
Number
JSME S CA1-2005

Title
Standards for Nuclear Power Generation Equipment: Standards
for Piping Wall Thinning Management (2005 enlarged issue)--Standards for Nuclear Power Generation Equipment: Standards
for Piping Wall Thinning Management for Pressurized Water
Reactor (2006)--Standards for Nuclear Power Generation Equipment: Standards
for Piping Wall Thinning Management for Boiling Water Reactors(2006)
JSME S NA1-2002 (*) Standards for Nuclear Power Generation Equipment: MaintenanceStandards (revised in 2002)
JSME S NA1-2004
Standards for Nuclear Power Generation Equipment: MaintenanceStandards (2004)
JSME S NB1-2001 (*) Standards for Nuclear Power Generation Equipment: WeldingStandards
JSME S NC1-2001 (*) Standards for Nuclear Power Generation Equipment: Design andConstruction Standards
JSME S NC1-2005 (*) Standards for Nuclear Power Generation Equipment: Design andConstruction Standards (2005)
JSME S ND1-2002
Standards for Nuclear Power Generation Equipment: DesignStandards
for Prevention of Piping Break
JSME S NE1-2003 (*) Standards for Nuclear Power Generation Equipment: Concrete
Reactor Containment Vessel
JSME S NF1-2006
Standards for Nuclear Power Generation Equipment: Evaluation
Method of Environmental Fatigue (2006)
JSME S FA1-2001
Standards for Spent Fuel Storage Facility: Structural Standard
for Metallic Cask
JSME S FB1-2003
Standards for Spent Fuel Storage Facility: Structure Standard forConcrete Cask, Canister Repack Device and Canister Transport
Cask
JSME S01 2 (*)
Evaluation Guideline on Hydro-Dynamic Vibration of Piping
Internal Cylindrical Structure
JSME S014
Standard for Validation of Active Component for Nuclear Power
PlantsJSME S016
Guideline for Prevention o f Fluid Induced Vibration of Tube and
U-Tube of Steam Generator
JSME S017 (*)
Evaluation Guideline on High Cycle Thermal Fatigue of Piping
JSME S NA-CC-001
Standards for Equipment, Maintenance and Case of Nuclear Power
Generation Facilities: SCC Crack Propagation Velocity of Nickel

7-18

Rich Alloy in PWR Primary Water Environments
JSME S NA-CC-002
Standards for Equipment, Maintenance and Case of Nuclear Power
Generation Facilities: Alternative Provision for Allowable DefectAngle Restriction for the Defect of Peripheral Direction
JSME S NA-CC-003
Standards for Equipment, Maintenance and Case of Nuclear Power
Generation Facilities: Criteria for Volumetric Examination (Eddy
Current Test) of a Steam Generator Tube
JSME S NC-CC-001(*) Standards for Equipment, Design and Construction and Case of
Nuclear Power Generation Facilities: Standard for Protection ofOver-pressurization
JSME S NC-CC-002(*) Standards for Equipment, Design and Construction and Case of
Nuclear Power Generation Facilities: Standard: Considerations
for Inhibition of Occurrence of Stress Corrosion Cracking

7-19

Table 7-2-3 Standards etc. of Academic Societies and Associations(Guidelines and Rules of the Atomic Energy Society of Japan)
NumberAESJ-SC-P001:2002
AESJ-SC-P002:2003
AESJ-SC-P003:2003
AESJ-SC-P004:2006
AESJ-SC-P005:2007
AESJ-SC-R003:2006
AESJ-SC-F001:2000
AESJ-SC-F003:2002

AESJ-SC-F005:2005
AESJ-SC-F006:2006
AESJ-SC-F007:2006
AESJ-SC-F008:2006

Title
Procedure of Probabilistic Safety Evaluation on Shutdown Condition of
Nuclear Power Station
Evaluation Criteria of Fuel Integrity after Transient Boiling Transition
for BWR
Performance Criteria of Wind Tunnel Test to obtain the effective heightof Discharge Source
Performance Criteria of the Periodic Safety Review for Nuclear Power
Plants (2006)
Performance Criteria of the Measures for Aging Management for Nuclear
Power Plants(2007)
Planning and Execution of Decommissioning of reactor facilities (2006)
Periodic Inspection Criteria of Cask for Spent Fuel, MOX fueland High Level Radioactive Waste
Measurement method of Sorption Distribution Coefficient―
Basic Procedure of Batch Method for Barrier Material of Near Face
Disposal
Method of Judgment on the Clearance Level (2005)
Safety Design and Inspection Criteria for Transfer Container for Spent
Fuels, New Mixed Oxide Fuels, and High Level Radioactive Wastes
Method of Safety Evaluation for Disposal of Very Low Level Radioactive
Wastes (2006)
Method of Measurement of Sorption Distribution Coefficient – Basic
Procedures of Measurement for Barrier Material for Deep Geological
Disposal (2006)
Performance Criteria of Probabilistic Safety Assessment of Events Causedby Earthquake on Nuclear Power Plants (2007)

7-20

Table 7-2-4 Standards etc. of Academic Societies and Associations(Guidelines and Rules of the Thermal and Nuclear Power Engineering Society)
NumberTNS-S3121-2003

Title
Qualification Standards for Industry Product on Weld of Electric
Facilities
JBWR-NCG-01-2005 (*) Guidelines for Accumulation Prevention of Mixed Gas (Hydrogen andOxygen) inside BWR Piping
Note: * Standards etc. of academic societies and associations that NISA has evaluated their technicaladequacy in order to utilize as exemplification standards of specification codes.

7-21

Atomic Energy Basic Law

Major Examination Guides
for Safety Design ofLight Water Nuclear Power
Reactor Facilities(Table 7-1)

Rules for the Installation,
Operation, etc. ofCommercial Power

7-22

Notification for Dose
Equivalent Limits on theBasis of the Rules forCommercial Power Reactors

Law on the Regulation of Nuclear
Source Material, Nuclear Fuel
Material and Reactors
Ordinance for the Regulation of
Nuclear Source Material, Nuclear Fuel
Material and Reactors

Rules for the Installation,
Operation, etc. of Nuclear
Power Reactors at the Stage of
Research and Development

Notification for Dose Limits on thebasis of the Rules for Manufacturing
Business of Nuclear Fuel Material etc.

Electricity Utilities Industry
Law
Electricity Utilities
Industry Law,
Ordinance

Electricity Utilities
Industry Law,
Rules
Ordinance Establishing
Technical Standards forElectrical Equipment(Fig. 7-2)

Technical Standards of
the Thermal and Nuclear
Power Engineering
Society of Japan

Technical Standards of
the Japan Society of
Mechanical Engineers
Society

Technical Standards of
the Atomic Energy
Society of Japan

Technical Standards of
the Thermal and Nuclear
Power Engineering
Society of Japan

Technical Standards of
The Japan Society of
Mechanical Engineers
Society

Law onCompensation for
Nuclear Damage

Industrial Safety and
Health Law

Law for TechnicalStandards ofRadiation Hazards
Prevention
Environment Impact
Assessment Law< Standards of Academic Societies and Associations >Technical Guideline
and Rules of theJapan Electric
Association

Special Law ofEmergency
Preparedness for
Nuclear Disaster

Law for Preventionof Radiation Hazards
due toRadioisotopes, etc.

< Technically Evaluated Standards of Academic Societies and Associations >Technical Guideline
and Rules of theJapan Electric
Association

Basic Law for
Emergency
Preparedness

Technical Standards of
the Atomic Energy
Society of Japan

Fig. 7-1 Major Legislations Governing the Safety Regulation of Nuclear InstallationsElectricity Utilities Industry Law

Ordinance for the
Enforcement theElectricity Utilities
Industry Law

Ordinance ofEstablishing TechnicalStandards on the
Welding of Electric
Structures – repealed
and included into theOrdinance No. 62

Rules for the
Enforcement of theElectricity Utilities
Industry Law

Ordinance ofEstablishing
Technical
Standards on
Nuclear Fuel
Material for Power
Generation

Ordinance ofEstablishing TechnicalStandards for Nuclear
Power Generation
Equipments (Ordinance
No. 62) – Changed to
Performance Req.

Ordinance ofEstablishing
Technical
Standards forElectric
Equipment

Ordinance ofEstablishing
Technical Standards
on Thermal Power
Generation(apply as a substitute with modifications)7-23

Technical Standards on
Structures etc., of
Nuclear Power
Generation Facilities(Notice) – (repealed)Technical Standards on
Dose, etc. due to
Radiation from Nuclear
Power Generation
Facilities

Technical Standards on
Structures etc., forConcrete Reactor
Containment(Notice) – (repealed)< Technically Evaluated Standards of Academic Societies and Associations >Society of
MechanicalEngineers of Japan
MaintenanceCriteria

Society of MechanicalEngineers of Japan
Standards for Design
and Construction

Society of
MechanicalEngineers of Japan
Standard forConcrete Reactor
Containment Vessel

Thermal and Nuclear
Power Engineering
Society Guidelines forProtection of
Accumulation of Mixed
Gas

The Japan Electric
Association
Rules for Primary
Reactor Containment
Vessel Leakage Testing

Other technicallyevaluated standardsof private sectors< Standards of Academic Societies and Associations >Technical Rules andGuidelines of the Japan
Electric Association(Table 7-2-1)

Technical Standards of the Japan
Society of Mechanical Engineers ofJapan(Table 7-2-2)

Technical Standards of theAtomic Energy Society ofJapan(Table 7-2-3)

Fig. 7-2 Systems of Technical Standards

Technical Standards of the
Thermal and Nuclear PowerEngineering Society( Table 7-2-4 )Site Selection
EnvironmentalImpact Assessment
ImpactEnvironmental
The
PublicHearing
PublicHearing(1st)(1st)Application for Reactor Establishment

Application for Reactor Establishment

Asking consultation

Asking consultation(peaceful use, planned
implementation andfinancial foundation)

Atomic Energy
Commission

Mandatory
Advice(radiological hazards prevention
and technical competency)

Examination by NISA

Establishment License

Mandatory
AdviceNuclear Safety
Commission

Public Hearing (2nd)Application ofConstruction Plan
Pre-Service Inspection
Licensee’s Welding
Inspection,
Welding Safety
Management Review
Fuel assembly inspection
Approval of Operational
Safety Program
Periodic Inspection
Licensee’s Periodic
Inspection,
Periodic Safety
Management Review
Operational Safety
Inspection

Construction Work

Commissioning

Operation
Notification of Dismantling
Implementation Order
Dismantling
Notification of Decommissioning

On-site Inspection

Implementation Order
Collection of Reports

Decommissioning

Fig. 7-3 Flow of Safety Regulations based on Legislations, etc.
for Nuclear Installations

7-24

Article 8 Regulatory Body
1. Each contracting Party shall establish or designate the regulatory body entrustedwith the implementation of the legislative and regulatory framework referred to in
Article 7, and provided with adequate authority, competence and financial and humanresources to fulfill its assigned responsibilities.
2. Each Contracting Party shall take the appropriate steps to ensure an effective
separation between the functions of the regulatory body and those of any other body ororganization concerned with the promotion or utilization of nuclear energy.

The regulatory body for ensuring the safety of facilities and activities for utilization of nuclear
energy in Japan is the Nuclear and Industrial Safety Agency (hereinafter referred as "NISA").
NISA, based on The Law on the Regulation of Nuclear Source Material, Nuclear Fuel Material
and Reactors etc. (hereinafter referred to as the "Reactor Regulation Law etc."), has definite
authority and power on safety regulations. Moreover, the Nuclear Safety Commission(hereinafter referred to as the NSC") established in the Cabinet Office supervises and audits
the regulatory activities of NISA. NISA is ensured to be effectively independent from otheragencies or organizations that manage matters for the promotion of nuclear energy utilization.
In addition, the Japan Nuclear Energy Safety Organization (JNES) was established under theauthority of METI, and is performing part of the Operational Safety Inspection for nuclear
installations in accordance with the Reactor Regulation Law and the Electricity Utilities
Industry Law, and is supporting NISA.
Since the previous report, NISA has conducted the Audit of Licensee's Periodic Check System
etc. in order to promote firm establishment of the inspection system revised in 2003, and hasreported the results of the said Audit to the NSC. Also, corresponding to the NSC’s result ofsupervision and audit on regulatory activities of NISA, NISA is making an effort to reform theregulations further. NISA also introduced the NISA Work Management System for improving
the transparency and efficiency of NISA’s work.
In addition, NISA, together with the NSC, invited the Integrated Regulatory Review Services(IRRS) by the International Atomic Energy Agency (hereinafter referred to as the “IAEA”) in
2007, in order to receive an international review concerning the nuclear-safety-regulatory
activities. Concerning the independency of the regulatory body from promotion organizations,the IRRS showed that "NISA is effectively independent from the Agency of Natural Resources
and Energy (hereinafter referred to as the “ANRE”) in correspondence with the GS-R-1(international standard of the IAEA). Moreover, this situation could be reflected in thelegislation more clearly in future.” 1

1 The IRRS Report is under the preparation by the IAEA at the time of this publication.

8-1

8.1 Mandate and Duties of the Regulatory Body
The mandate of the regulatory body is to ensure the safety of nuclear installations, and its
duties are to enforce the legislative and regulatory framework described in the report ofArticle 7.
One of the important requirements for the regulatory body satisfying his responsibility is, as
indicated in Article 8, Paragraph 2 of this Convention, to ensure effective separation between
functions of the regulatory body and those of any other body or organization concerned withthe promotion or utilization of nuclear energy. Another important function of the regulatory
body is to keep communicating independently with the public of its regulatory decisions, its
opinions and its basis.
On the basis of the Atomic Energy Basic Law, the regulatory body is responsible to conduct
regulatory activities prescribed in the Reactor Regulation Law, the Electricity Utilities
Industry Law, etc. As for legislations and regulations etc. applied to the examination of thebasic design or basic design policies of nuclear installations and to the inspection of nuclear
installations in the construction and operational stages, NISA and the NSC work toward
improvement and enhancement of legislations and regulations based on operating experiences,
trends of the latest knowledge of the technology advancement, etc., and the international
consensus.
In the case of a nuclear emergency, the Basic Law on Emergency Preparedness, the Special
Law of Emergency Preparedness for Nuclear Disaster and other related laws are applied.
Relevant administrative bodies in such a case are described in the report of article 16.
8.2 Organizations for Enforcement of the Safety Regulation of Nuclear Installations
In Japan, the Minister of Economy, Trade and Industry (hereinafter referred to as “METI”)
serves as the minister in charge of safety regulation for all facilities and activities concerning
the utilization of nuclear energy. NISA has been established in METI as an independent"special organization" dedicated to the administration of safety regulations.
NISA has been executing policies independently from the Agency of Natural Resources andEnergy dedicated to promote nuclear energy. The incorporated administrative agency, JNES
was established in October 2003. JNES, together with NISA, provides infrastructure to assure
safety in the use of the nuclear energy.
The NSC and the Atomic Energy Commission (AEC) were established in the Cabinet Office.
The commissioners of both of these commissions are appointed by the Prime Minister with theconsent of the Diet.
Each of these two commissions’ plans, deliberates, and decides policies concerning either
nuclear power application or ensuring the safety, from the standpoint to regulate respectively
in the country as a whole.

8-2

As described in the report of Article 7, NISA conducts a safety examination of nuclear
installations, and the Minister of METI consults the NSC and the AEC on the results of theexamination.
The NSC submits to the Minister of METI a report, after an independent examination andpublic hearings on the specific safety of the nuclear installation. The NSC establishes guides
to be used for the examination.
Fig. 8-1 presents an overview of administrative organizations that are responsible for thesafety regulation of nuclear installations.
8.3 Nuclear and Industrial Safety Agency (NISA)(1) The Role of NISANISA administrates the safety regulations for nuclear installations. Specifically, NISA,
entrusted by the Minister of METI, conducts clerical work concerning the competence of the
Minister of METI as follows:
The Minister of METI, who is the competent minister stipulated in the Reactor Regulation
Law, has the authority to issue licenses for establishment of nuclear installations, after
conducting the examination of siting, structure, and equipment, so that a radiological hazard
due to establishment of a nuclear installation is prevented. The Minister of METI has theauthority to revoke a license under circumstances such as violation of the Reactor Regulation
Law by the licensee.
The Minister of METI has the authority to establish ministerial orders concerning measures
for the safe operation and physical protection of specific nuclear fuel material, the Operational
Safety Program, Chief Engineer of Reactors, measures in emergency, etc. for the operation ofreactors. The Minister of METI has the competence of authorizing the Operational Safety
Program, approval of the operation plan and decommissioning plan of nuclear installations,
accepting the notification concerning appointment/dismissal of Chief Engineers of Reactors,
collecting reports from licensees and conducting On-site Inspection of the licensees,
revocation or discontinuance of utilization of a license for establishment of a nuclear
installation, ordering of measures for safe operation etc., approval of a Chief Engineer ofReactors, an order on measures for safe operation etc., dismissal order of a Chief Engineer ofReactors, implementation order concerning a decommissioning, implementation order for an
emergency preparedness, etc.
The Minister of METI, and the Minister of MEXT, conducts examinations for Chief Engineersof Reactors and issues the licenses. The Minister of METI has the authority also to order toreturn such licenses in a case of violation of the law by the Chief Engineers.
The Minister of METI, who is the competent minister stipulated in the Electricity Utilities
Industry Law, has powers to establish ministerial ordinances relating to the technical standard,
pre-service inspection, the fuel assembly inspection, the Audit of Licensee’s Welding Check

8-3

System, the Periodic Inspection, and the Audit of Licensee’s Periodic Check System. The
Minister of METI has powers to conduct approval of the construction plan, the pre-service
inspection including verification of the safety performance of a whole power station, the fuelassembly inspection, the Periodic Inspection, and to issue an Order for Conformity to thetechnical standards when a case of nonconformity to the technical standards is found. The
Minister of METI has powers also to hold the affairs of examinations for Chief ElectricalEngineers, to issue licenses for the Chief Electrical Engineer and the Chief Engineer of Boiler
and Turbine, and to order the return of such licenses in case of violation of the law by a Chief
Engineer.
NISA evaluates results of the Audit of Licensee’s Periodic Check System performed by JNES.
In the Audit of Licensee's Periodic Check System, JNES evaluates the organization, inspection
methods, process control, and other matters of a licensee concerning implementation of theperiodic licensee's check, which are determined by a METI ordinance.
NISA evaluates also the results of the Audit of Licensee’s Welding Check System performed
by JNES.
(2) Organization of NISANISA was established as a "Special Organization" in METI, and has 11 divisions dedicated tothe administration of the safety regulation of nuclear installations (including nuclear fuelcycle facilities).
They are Policy Planning and Coordination Division, Nuclear Safety Public Relations andTraining Division, Nuclear Safety Regulatory Standard Division, Nuclear Safety Special
Investigation Division, Nuclear Power Licensing Division, Nuclear Power Inspection Division,
Nuclear Fuel Transport and Storage Regulation Division, Nuclear Fuel Cycle Regulation
Division, Radioactive Waste Regulation Division, Nuclear Emergency Preparedness Division
and Electric Power Safety Division. The assigned duties of those divisions are provided inTable 8-1.
Nuclear Safety Inspectors are assigned to each site of the nuclear installations. Fig. 8-2 shows
the locations of the Nuclear Safety Inspectors Offices.
NISA has a total of approximately 350 staff engaged in the nuclear safety regulation, out of
which 100 staff members are Nuclear Safety Inspectors and Senior Specialists for Nuclear
Emergency stationed at nuclear installations.
(3) Quality Improvement of NISA’s Regulatory Activities
NISA provides a strong commitment to its mission, scientific and reasonable judgments,
transparency, neutrality and fairness as the code of conduct for their activities. In this context,
the Policy Planning and Coordination Division watches and assesses the performance of otherdivisions of NISA in discharging their duties, and take timely remedial actions after
consulting with the senior managements. In order to improve the quality of regulatory

8-4

activities, the development of the NISA Work Management System started in fiscal year 2006
and implemented from fiscal year 2007. According to the NISA Work Management System,NISA’s goals in the medium term and tasks in fiscal year 2007 were released in June, 2007.
The NSC, an independent organization from NISA, supervises and audits the appropriatenessof NISA’s regulatory administration in the construction and operation stages after issuance ofthe license, from the view points of rationality, effectiveness and transparency. Thus, theframework that confirms the quality of the safety administration is maintained.
In addition, NISA makes a continuous effort to maintain the high quality of regulation througheducation and training of the personnel as stated in the report of Article 11, international
activities and the hearing of advice from experts e.g. members of the Nuclear and Industrial
Safety Subcommittee (hereinafter referred to as the “NISS”).
The "Law for Evaluation of the Policies Executed by Administrative Organizations" wasenforced in April 2002, and in accordance with this law, a framework, with which eachadministrative organization of the government evaluates and improves his own policies
systematically, was built. METI has developed plans to evaluate the regulatory systems within
its jurisdiction in fiscal year 2004, and NISA, according to these plans, evaluates the nuclear
safety regulation system on the basis of the Reactor Regulation Law and the Electricity
Utilities Industry Law.
(4) Further Approach to Information Disclosure
NISA started information service activities systematically in the form of integrating with theregulatory work process in September 2001, introducing relationship management (RM) as a
new effort, which makes the feedback from the outside into a qualitative operation ofregulatory activities, and is promoting positive information disclosure activities. The
objectives of the RM are to make effort for improvements in recognition of NISA’s
responsibility, in the people’s understandings about NISA’s daily activities, in ensuring a
steady response to the people’s concerns, in establishing consent to a better regulatory
system, in the preparation to an emergency, and in the activation of internal communication.
In April 2004, NISA allocated a new budget in order to enhance further activities to hear fromthe public, and at same time, established the Nuclear Safety Public Relations and Training
Division newly formed in NISA, and appointed the Resident Public Relations Officers. The
main activities of NISA from fiscal year 2004 to fiscal year 2006 were as follows, (1) NISA
executive’s visit and give explanation of NISA’s policy and activities to the local government(133, 113 and 64 visits in fiscal year of 2004 to 2006, respectively ), (2) publication of
newsletters and mail magazines, (3) explanation of policies and activities of the nuclear safety
regulation to the general public, (public meeting on the clearance system and amendment of
the Reactor Regulation Law, Pu-thermal symposium, public meeting on seismic safety,"one-day seminars to introduce NISA" were held in major cities and site municipalities), (4)making direct dialogue group communication with site area residents (in fiscal year 2006,"dialog meetings" were held at ten places across the country, such as Tomari-mura, Hokkaido

8-5

and Genkai-cho, Saga-Prefecture), (5) activities to hear from the public at the Nuclear Safety
Inspector's Offices, (6) implementation of risk-communication technical training for
local-government personnel etc., (7) introduction of NISA and Nuclear Safety Inspectors
Offices; editing of a video explaining about essential policies, such as the new inspection
system by managements themselves and televising by site area CATVs and placing suchinformation on a homepage.
Also, NISA opens the Nuclear Energy Library in JNES, where the public can accessdocuments for the reactor establishment license, reports of incidents and accidents of nuclear
installations and, books and booklets on energy and nuclear power generation.
8.4 Organizations related to NISA
(1) Council etc.
On the basis of the Law for Establishment of the Ministry of Economy, Trade and Industry, theAdvisory Committee for Natural Resources and Energy (hereinafter referred to as the“ACNRE”) was established, a subcommittee of which is the NISS that proposes policies on
nuclear safety and safety of electric power as terms of reference. The organization of the NISS
is given in Table 8-2.
The Minister of METI appoints members of the ACNRE from persons of knowledge andexperience, and these members select a chairperson of the ACNRE mutually. Subcommittees
are established by a resolution of the ACNRE, and the chairperson designates members of thesubcommittees including the NISS. The members of the subcommittees are assigned based on
their expertise and experience from the fields of nuclear and thermal-hydraulic design, nuclear
fuel design, system design, equipment design, seismic design, material strength, radiation
control, meteorology, geology, soil etc.
"What challenges exist in the future in order to assure safety in nuclear power generation andsafety in the electric power system operation, while under rapid social and economicalchange" were entrusted to NISS to be discussed. The NISS and other subcommittees havedeliberated on what nuclear safety regulation systems should be, and the results were reported
to NISA.
NISA solicits views of experts and members of NISS.
(2) JNESJNES, consisting of about 420 officers and staff, was established in October 2003 as anorganization that establishes the infrastructures in cooperation with NISA to ensure the safetyof utilization of nuclear energy.
The mission of JNES is to implement its duties with full application of its technical andengineering competence based on scientific judgments to contribute to the improvement of
nuclear safety regulation and, to deliver and transmit actively the safety information to the
public.

8-6

JNES is expected to ensure the nuclear safety and build the confidence of the people innuclear safety by implementing such duties.
JNES implements the following activities:•

Inspection of nuclear installations and reactor facilities, and related work,•

Safety analysis and evaluation of designs of nuclear installations and reactor facilities;•

Work for the establishment of nuclear emergency preparedness, prevention of theescalation of a nuclear emergency (including minimization of the probability ofoccurrence of a nuclear emergency), and restoration from a nuclear emergency;•

Investigation, testing, research, and training to ensure safety in utilization of nuclear
energy; and•Collection, analysis and delivery of information to assure nuclear safety.

The procedures for JNES to implement activities, keeping in relation with NISA of METI, are
as shown in the following:•

NISA develops a goal on each activity based on the regulatory needs, and defines themedium-term objectives in accordance with the Act on General Rules for Incorporated
Administration Agencies, and the Minister of METI assign them to JNES.

•JNES prepares a scheme (medium-term scheme) to accomplish the medium-term
objectives, applies for and obtains the approval of the scheme to the Minister of METI,,
then JNES prepares a program in accordance with the medium-term scheme for every FYterm, notifies the program to the said minister and implements it.

8.5 The NSC
The Atomic Energy Basic Law was partially revised on October 4, 1978 to establish the NSC
under the Prime Minister’s Office. The NSC administers the function of safety regulation, had
belonged to the AEC up until then, in order to strengthen the system of ensuring nuclear safety.
(The NSC was transferred from the Prime Minister’s Office to the Cabinet Office due tocentral government reform in January 6, 2001.)
The NSC is responsible for planning, deliberation and decisions on matters that are related toensuring safety of the research, development, and utilization of nuclear energy.
The NSC conducts its own review of the results of NISA’s examination on the applicationfrom the view points of the licensee’s technical capability and non hindrance to the preventionof radiological hazards. The NSC supervises and audits the appropriateness of NISA’s
regulatory administration in construction and operation stages after issuance of the license,from the viewpoint of reasonableness, effectiveness and transparency. Thus, the framework

8-7

that confirms the quality of the safety administration is maintained.
When the NSC deems it necessary as a part of its assigned duties, the NSC may recommend
and may request reports and cooperation concerning the submission of materials, statementsof viewpoint, and explanation to the heads of relevant administrative organizations, by way of
the Prime Minister.
Since April 2003 (partially, from October 2003), the above functions have legally beenenacted. The NSC receives from NISA the following; reports on the quarterly bases after theapproval of a license to establish nuclear installations: reports concerning the conduct of theregulatory activities such as approval of the construction plan, pre-service inspection, Periodic
Inspection, Audit of Licensee's Periodic Check System, Audit of Licensee's Welding Check
System, Approval of Operational Safety Program, implementation states of regulations, suchas the Operational Safety Inspection, report of accidents and failures of nuclear installations.
The NSC also has the authority to inquire directly of the licensees, maintenance andinspection contractors in order to supervise and to audit the safety regulation implemented byregulatory body.
In the case of a violation of safety regulations in any of nuclear facilities, the employee candirectly allege the fact to the NSC, and the NSC has the authority to investigate the allegation.
The Minister of METI, before issuing a license to establish nuclear installations, must receive
the viewpoint of the NSC on the following matters: (1) that the applicant for the license of a
nuclear installation has adequate technical capability to establish and reliably operate anuclear reactor, and (2) that the site, structures and equipment of the nuclear installation
would not cause any hindrance to the prevention of radiological hazards.
The NSC is composed of five commissioners appointed by the Prime Minister with theconsent of the Diet, and these commissioners elect a chairman among them. General affairs of
the NSC are performed by the NSC Secretariat of the Cabinet Office. The NSC Secretariat iscomposed of the Secretary-General, the General Affairs Division, the Regulatory Guides andReview Division, the Radiation Protection and Accident Management Division and theSubsequent Regulation Review Division and has about 100 personnel.
Under the NSC, two safety examination committees, eight special committees and sevenothers are organized as shown in Table 8-3. The Special Committees may organize working
groups under them, if necessary.
The members of the Committee on Examination of Reactor Safety and the Committee on
Examination of Nuclear Fuel Safety are appointed from persons of knowledge and experienceby the Prime Minister in accordance with the Law for Establishment of the Atomic Energy
Commission and the Nuclear Safety Commission. The Emergency Technical Advisory Body iscomposed of the commissioners of the NSC and the commissioners on the Emergency
Technical Advisory Body who are also appointed by the Prime Minister from persons ofknowledge and experience.

8-8

Results of the investigation and evaluation by each review board and special committee arereported to the NSC and are deliberated by the NSC. Reflecting the results of the discussion in
the Emergency Technical Advisory Body, the NSC determine the recommendation items for an
emergency.
Deliberations of all committees, including the special committees and working groups under
the NSC are open to the public. The contents of the deliberations are provided for the public
on a homepage (http://www.nsc.go.jp/) and at the Nuclear Energy Library.
8.6 The AEC
The AEC was established under the Prime Minister's Office on January 1, 1956, on the basisof the Atomic Energy Basic Law and the Law for Establishment of the Atomic Energy
Commission and the Nuclear Safety Commission, to conduct national policy concerning
research, development and utilization of nuclear energy in a planned manner and to ensure thedemocratic administration of the nuclear energy policy. (The AEC was transferred to the
Cabinet Office in January 2001.)
The AEC has duties of planning, deliberation, and decisions concerning the research,development and utilization of nuclear energy (excluding matters relating to regulations on
ensuring safety).
If the AEC deems it necessary as part of its assigned duties, it may advise by way of the Prime
Minister, and request reports and cooperation including the submission of materials,
statements of viewpoint, and explanation from the heads of relevant administrative
organizations.
The Minister of METI, before issuing a license to establish nuclear installations, shall receive
views of the AEC with regard to the following items: (1) the nuclear installations will not be
used for any purposes other than peaceful purposes, (2) the license will cause no hindrance tothe planned development or utilization of nuclear energy, and (3) the applicant has anadequate financial basis to construct and maintain the nuclear installations.
The AEC is composed of a chairman and four other commissioners appointed by the Prime
Minister with the consent of the Diet.
8.7 Other Administrative Bodies
Establishment of nuclear installations necessitates the compliance with other laws such as theFire Protection Law and the Port Regulation Law. Therefore, the relevant safety regulations
are conducted by the relevant government offices e.g. the Fire Protection Agency and theMinistry of Land, Infrastructure and Transport.

8-9

Table 8-1 Assigned Duties of the Divisions Related to Safety Regulation of Nuclear
Installations (including nuclear fuel cycle facilities), NISA, METI
Policy Planning and
Coordination Division
Nuclear Safety Public
Relations and Training
Division

Nuclear Safety Regulatory
Standard Division

Nuclear Safety Special
Investigation Division
Nuclear Power Licensing
Division
Nuclear Power Inspection
Division
Nuclear Fuel Transport andStorage Regulation Division
Nuclear Fuel Cycle Regulation
DivisionRadioactive Waste Regulation
Division
Nuclear Emergency
Preparedness Division

Electric Power Safety Division

・ Planning and coordination the general policy of the
NISA
・ Activities for public hearing and public relations
concerning the nuclear safety
・ Administration of the Nuclear Safety Inspectors andSeniorSpecialists
for
Nuclear
Emergency
Preparedness
・ Training and education of personnel to gain and to
improve their competency
・ Planning and coordination concerning technology andthe system to ensure the nuclear safety
・ Regulation of nuclear power reactors in the stage of
research and development
・ Research and development, etc.
・ Management of allegation and litigation concerning
nuclear safety
・ Regulation of commercial powerreactors in thedesign and construction stage
・ Regulation of commercial power reactors in theoperation stage
・ Regulation of spent nuclear fuel storage business・ Regulation concerning transportation of nuclear fuel・ materials from sites
・ Regulation concerning businesses of refining,processing,・ fabrication, spent-fuel storage, and reprocessing
・ Regulation of radioactive waste business, dismantlingand decommissioning of nuclear installations including
nuclear fuel cycle facilities
・ Planning of nuclear emergency preparedness
・ Prevention and investigation of incidents andaccidents in nuclear businesses
・ Administration of activities in a nuclear emergency
・ Matters concerning physical protection
・ Regulation of welding for electric structures
・ Environmental impact assessment

8-10

Table 8-2 Organization of the Nuclear and Industrial Safety Subcommittee, ACNRE
Basic Safety Policy
Subcommittee
Nuclear Reactor Safety
Subcommittee
Nuclear Fuel Cycle Safety
Subcommittee
Decommissioning Safety
Subcommittee
Radioactive Wastes Safety
Subcommittee
Seismic and Structural Design
Subcommittee
Nuclear Emergency
Preparedness Subcommittee

・ General matters securing safety
・ Technical matters on commercial power reactors andpower reactors at the stage of research and development
・ Fabrication and reprocessing of nuclear fuel, storage ofspent fuel, transportation of nuclear fuel material, andthe technical standards
・
Decommissioning of nuclear installations *・ Securing safety of disposal and storage of radioactive
wastes・ Technical matters on seismic safety and structural
integrity of nuclear installations
・ - Measures for incidents and failure, and general crisis
management for emergencies of nuclear installations *and physical protection of nuclear material
INES Evaluation Subcommittee ・ INES Evaluation on incidents and accidents of nuclear
installations *Subcommittee for the・ Matters related to the Convention on Nuclear Safety andConvention on Nuclear Safety ・ international standards on nuclear safety
Electrical Power Safety
Subcommittee
Study Group on the Way ofInspection
Subcommittee for the Joint
Convention on Radioactive
Waste and Spent Fuel Safety
Subcommittee for theInstitution of Nuclear Safety
Regulation

・ Securing safety of electrical power・ Matters concerning the inspection system of nuclear
power generation facilities and nuclear fuel cycle
facilities
・ Matters related to the Convention on Joint Convention
Radioactive Waste and Spent Fuel Safety

・ Study of the legal system for the prevention offalsification of the self-controlled inspection record
based on the investigation of the background of thefalsification
Subcommittee for・ Study of the following, in the cases where a plant has
Fitness-for-Service
cracks in a core shroud or reactor coolant
Assessment etc. of nuclear
re-circulation system piping:(1) Verification of validity in the check methods for
power systemcore shroud etc.
(2) Technical fitness-for-service assessment judgment
method(3) Fitness-for-service verification etc. of individual
plants based on check results specifically
Nuclear Safety Infrastructure ・ Study of the current state and issues etc. of safety
Subcommittee
infrastructure (safety infrastructure study, codes andstandards, human-resources infrastructure, study offacility infrastructure, knowledge base)
Aging Countermeasure
・ Clarification of the requirements used as the rationale
Examination Committeeof measures for aging management, guidelines, etc.
and study about the way of the reasonable safety
inspection by the government*: Including nuclear fuel cycle facilities

8-11

Table 8-3 Examination and advisory bodies under the NSC
Committee on Examination of・ Matters concerning the safety of nuclear reactor facilities
Reactor Safety
Committee on Examination of・ Matters concerning the safety of nuclear fuel material
Nuclear Fuel Safety
Emergency Technical Advisory
・ Technical advice in emergency measures in case ofBody
occurrence of an accident or a failure that meet the given
standard level in nuclear installation etc.
Emergency Technical Advisory
・ Technical advices in emergencies/ disasters of nuclear
Body for Disaster Prevention ofcarriers
Nuclear Carriers・ Technical advices in nuclear emergencies/ disasters due to
Emergency Technical Advisory
armed attacks
Body for Nuclear Carrier Nuclear
Disaster Prevention due to armed
attackSpecial Committee for Nuclear
・ Matters concerning safety standards and guides of nuclear
Safety Standards and Guides
reactors, nuclear fuel facilities, and other nuclear
installations
Special Committee on
・ Matters concerning safety assurance in radioactive waste
Radioactive Waste anddisposal・ Matters concerning the safety assurance in
Decommissioning
decommissioning nuclear installation
Special Committee on Safety
・ Establishment of safety goals
Goals
Special Committee on Radiation ・ Matters concerning the radiation protection considering
Protection
domestic and international trends
Special Committee on Safety
・ Matters concerning the safety assurance in transportation
Transport of Radioactiveof radioactive materials considering domestic andMaterials
international trends
Special Committee on Analysis
・ Analysis and evaluation of domestic and international
and Evaluation of Nuclear
nuclear accidents and failures
Accidents and Failures
Special Committee on Nuclear
・ Planning of nuclear safety research programs・ Monitoring of the nuclear safety research programs
Safety Research・ Evaluation of the nuclear safety research programs
Special Committee on Nuclear
・ Emergency preparedness in the vicinity of nuclear
Disaster
installations, etc.
Task Force for introduction of・ Review and analyses of the issues in the introduction of
Safety Regulations Using Risk
safety regulation using risk information
Information
Project Team on Safety Survey ・ Survey and analysis of matters relevant to the safetyof Reprocessing Facilities
regulation activities during the test operation of the
Rokkasho reprocessing facility
Safety Investigation on・ Technical matters concerning the safety assurance in the
Disposal of Specialized
final disposal of high-level radioactive wastes
Radioactive Wastes
Investigation Project Team on
・ The review of Results of Seismic Safety Re-evaluation ofSeismic Safety of Nuclear
Existing Plants
・ Matters concerning the newest knowledge on seismic
Facilities
safety8-12Administrative Organization for Supervision and Auditing of Regulatory Activities
Prime Minister
Nuclear Safety
Commission *Cabinet OfficeSecretary - General

Management andCoordination Division
Regulatory Guides andReview Division
Radiation Protection and
Accident Management
Division
Subsequent Regulation
Review Division* Established in accordance with the Law
for Establishment of the Atomic Energy
Commission and the Nuclear Safety
Commission

Administration Organization for Regulatory Activities

Policy Planning and
Coordination Division
Nuclear Safety Public
Relations and Training
Division

Minister
Ministry of Economy, Trade and Industry

Nuclear Safety Regulatory
Standard Division
Nuclear Safety Special
Investigation Division

Director-General
Nuclear and Industry Safety Agency

Nuclear Power Licensing
Division

Director-General for NISA,
Deputy Directors-General,
DeputyDirector-General
Safety Examination

Nuclear Power Inspection
Divisionfor(Established in accordance with
the Ministry of Economy, Trade
Industry Establishment Law)

Nuclear Fuel Cycle
Regulation Division
Nuclear Fuel transport andStorage Regulation Division
Radioactive Waste
Regulation

Incorporated Administration Agency,
Japan Nuclear Energy Safety Organization(Established in accordance with the
General Rules for the Incorporated
Administrative Agency and the Law for
the Incorporated Administrative
Agency, Japan Nuclear Energy Safety
Organization)

Nuclear Emergency
Preparedness Division
Electric Power Safety
Division
Nuclear Safety Inspectors(Located in 17 sites of the
Nuclear Power Plants)

Fig. 8-1 Outline of the Safety Administrative Organization for Nuclear Installations(including the Nuclear Fuel Cycle)

8-13

Tomari Nuclear Safety Inspectors Office

Units 1 & 2 <PWR> (In operation), & Unit 3 <PWR> (Under construction)
Tomari PS, Hokkaido Electric Power Co.
Kashiwazaki-kariwa Nuclear Safety Inspectors Office

Units 1, 2, 3, 4 & 5 <BWR>), & Units 6 & 7 <ABWR> (In operation)
Kashiwazaki-kariwa NPS, Tokyo Electric Power Co.
Sapporo -city

Shika Nuclear Safety Inspectors Office

Unit 1 <BWR> & Unit 2 <ABWR> (In operation),
Shika NPS, Hokuriku Electric Power Co.
Higashidori Nuclear Safety Inspectors Office

Tsuruga Nuclear Safety Inspectors Office

8-14

Unit 1 <BWR> & Unit 2 <PWR> (In operation)
Tsuruga NPS, the Japan Atomic Power Company
Fugen <ATR> (Under decommissioning setup) & M onju <FBR>(Under construction, First Criticality Attained)
Japan Atomic Energy Agency

Unit 1, <BWR> (In operation),
Higashidori NPS, Tohoku Electric Power Co., Inc.
Aomori-city

Mihama Nuclear Safety Inspectors Office

Onagawa Nuclear Safety Inspectors Office

Units 1, 2 & 3 <PWR> (In operation),
Mihama PS, the Kansai Electric Power Co., Inc
Sendai-city
Ohi Nuclear Safety Inspectors OfficeNiigata-city
Fukushim a-city

Units 1, 2, 3, & 4 <PWR> (In operation),
Ohi PS, the Kansai Electric Power Co., Inc
Takahama Nuclear Safety Inspectors Office

Units 1, 2, 3, & 4 <PWR> (In operation),
Takahama PS, the Kansai Electric Power Co., Inc

Kanazawa- city
Fukui- city

Units 1 & 2 <BWR> (In operation),
Shimane NPS, the Chugoku Electric Power Co., Inc

Units 1, 2, 3, 4, 5, & 6 <BWR> (In operation),
Fukushima Daiichi NPS, Tokyo Electric Power Co., Inc.
Units 1, 2, 3, & 4 <BWR> (In operation),
Fukushima Daini NPS, Tokyo Electric Power Co., Inc.
Tokai & Oharai Nuclear Safety Inspectors Office

Matsue- city
Shizuoka-city

Genkai Nuclear Safety Inspectors Office

Units 1, 2, 3, & 4 <PWR> (In operation),
Genkai NPS, Kyushu Electric Power Co., Inc

Matsuyama-city
Saga-city<GCR> (Under decommissioning), Tokai PS &<BWR> (In operation), Tokai No.2 PS
The Japan Atomic Power Company

Hamaoka Nuclear Safety Inspectors Office

Units 1, 2, 3 & 4 <BWR> & Unit 5 (ABWR) (In operation),
Hamaoka NPS, Chubu Electric Power Co., Inc

Sendai Nuclear Safety Inspectors Office

Unit 1 & 2 <PWR> (In operation),
Sendai NPS, Kyushu Electric Power Co., Inc

Fukushima Daiichi Nuclear Safety Inspectors Office

Fukushima Daini Nuclear Safety Inspectors Office

Mito-city

Shimane Nuclear Safety Inspectors Office

Units 1, 2 & 3 <BWR> (In operation),
Onagawa NPS, Tohoku Electric Power Co., Inc.

Ikata Nuclear Safety Inspectors Office

Unit 1, 2 & 3 <PWR> (In operation),
Ikata PS, Shikoku Electric Power Co., Inc
Kagoshima-city

Fig. 8-2 Establishment of Nuclear Safety Inspectors Offices
As of August 31, 2007

Article 9 Responsibility of Licensee
Each Contracting Party shall ensure that prime responsibility for the safety of a nuclear
installation rests with the holder of the relevant licence and shall take the appropriate
steps to ensure that each such licensee meets its responsibility.

The prime responsibility for the safety of a nuclear installation rests with the licensee, while
the regulatory body establishes relevant regulation to ensure the public safety and supervisesthat the licensee complies with the regulation.
The rules for the Reactor Regulation Law clarifies the licensee’s responsibility on quality
assurance and maintenance management, while the Electricity Utilities Industry Law clarifies
the licensee’s responsibility with the Periodic Licensee’s Check and the Audit of Licensee’s
Periodic Check System
9.1 Regulatory Measures for the Licensee to take the Prime Responsibility
The prime responsibility for the safety of a nuclear installation rests with the licensee and thelicensee shall comply with the regulatory requirements in each stage from planning throughoperation, which are stipulated in the Reactor Regulation Law, the Electricity Utilities
Industry Law, etc. Those regulatory requirements are described in the Article 7 of this report.
The following activities of the licensee are described in the respective articles of this report;• Education and Training of Operational Personnel etc. (Article 11),• Quality Assurance Activities (Article 13)• Periodic Safety Review (Article 14),
• Aging Management Review (Article 14),
• Emergency Preparedness (Article 16),
• Design and Construction (Article 18) and• Operation (Article 19)
In addition, in order to ensure safety, not only meeting with these regulatory requirements, thelicensees are continuously to make effort for improving the safety and reliability of their
nuclear installation by taking the following measures;• Education and training of operators and maintenance personnel and preparation ofeffective operation manual,• Collection, examination and exchange of the information related to the operating
experiences, ,• Study of the state-of-the-art technical insight and implementation of safety research,• Adoption of operating experience, etc. to design, operation and maintenance,• Implementation of quality assurance activities and• Preparation of the accident management, etc.

9-1

9.2 Supervision of Licensees by Regulatory Body
The basic mechanism to ensure the safety of nuclear installations is that NISA issues licenses,
orders the licensee to bear the prime responsibility for safety and supervises it within thelegislative and administrative framework.
The following is an overview of the above mentioned mechanism.
(1) Licensing
The Minister of METI issues a license for the establishment of a nuclear installation after
examining that the nuclear installation will not be used except for the peaceful purposes, that
there is no potential obstacle for accomplishing the planned development of atomic energy,
that technical capability and financial foundations of licensees are sufficient, and that the site,
the structure and the equipment of the nuclear installation may not cause any hindrance to theprevention of nuclear emergency. The regulation under the Reactor Regulation Law and theElectricity Utilities Industry Law in each stage from planning through operation is describedin section 7.3.
(2) Periodic Licensee's Check and Audit of Licensee's Periodic Check System
In addition to the confirmation by NISA at the Periodic Inspection, licensee’s self-controlled
inspection that was carried out voluntarily by licensees has been upgraded to the mandatory"Periodic Licensee's Check" through the amendment of the Electricity Utilities Industry Law
in 2003. Accordingly licensees shall inspect the nuclear installations subject to the TechnicalStandards and confirm the conformity and keep the records of the results. . The implementing
system of the inspections above is audited by JNES as the "Audit of Licensee's Periodic Check
System" and the audit results are reported to NISA. NISA evaluates the audit results andpublicizes its conclusion.
(3) Operational Safety Inspection and Nuclear Safety Inspector
NISA conducts “Operational Safety Inspection” periodically based on the Reactor Regulation
Law to confirm whether the compliance to the Operational Safety Program is assured. In
accordance with the Reactor Regulation Law, NISA stations the Nuclear Safety Inspectors at
each nuclear installation, who conducts the Operational Safety Inspection four times a year to
confirm the licensee’s compliance with the Operational Safety Program, and addresses
incidents if they occur.
(4) Quality Assurance and Maintenance Management Activities
In accordance with the ordinance based on the Reactor Regulation Law, the licensee shallestablish quality assurance system and maintenance management system and include them in
the Operational Safety Program. NISA confirms the compliance with the Operational Safety

9-2Program through the Operational Safety Inspection.
(5) The Senior Specialist for Nuclear Emergency
In accordance with the Special Law for Nuclear Emergency, NISA stations Senior Specialist
for Nuclear Emergency at each site of nuclear installations, who guides and advises thelicensee in preparing the Licensee’s Plan for Emergency Preparedness, and conducts duties
necessary to prevent nuclear emergency and mitigate the consequence should it occur.
(6) Periodic Safety Review
In accordance with the ministerial order based on the Reactor Regulation Law, licensees shallconduct the Periodic Safety Review for the operating reactor facilities every 10 years.
(7) Aging Management Review
In accordance with the ordinance based on the Reactor Regulation Law, licensees shallperform technical review on aging for the safety-related equipment and structures of nuclear
installations and to establish and implement the Ten-Year Maintenance Program in no later
than thirty years after the start of commercial operation. NISA reviews the technicalevaluation and maintenance program prepared by the licensees.
(8) Accident Management
The licensee prepares an accident management program according to the “AccidentManagement of Severe Accidents at Power Generating Light Water Reactor Facilities”, adecision by the NSC, 1992 (partly revised by the NSC in 1997), and submits it to NISA for
review. NISA reviews and evaluates the technical adequacy of it.
(9) Reports on accidents and failures
In accordance with the Reactor Regulation Law or the Electricity Utilities Industry Law, thelicensee shall report to NISA on accidents or failures.
(10) On-site Inspection
NISA conducts on-site inspection, if necessary, at the plants, offices, etc. of licensee or its
contractor (welders) in accordance with the Reactor Regulation Law or the Electricity Utilities
Industry Law(11) Revocation
Judging that the licensee violates regulation, the Minister of METI may take measures of

9-3

enforcement such as revocation of the license, suspension of operation, fine, etc., in accordancewith the Reactor Regulation Law or the Electricity Utilities Industry Law.

In accordance with the provisions of Article 35 of the Reactor Regulation Law, licensees shalltake necessary measures for 1) ensuring safe conditions of the nuclear facilities, 2)maintaining safe operation and 3) the safe transportation, storage and disposal of nuclear
fuel materials or materials contaminated with nuclear fuel materials. When the above isviolated, the Minister of Economy, Trade and Industry can order suspension of operation of
the nuclear facilities. Moreover, in accordance with the provision of Article 39 of the"Electricity Utilities Industry Law", licensees are obliged to meet the technical standards asfor electric facilities and when it is violated, the Minister of Economy, Trade and Industry canorder the suspension of operation of the electric facilities.
9.3 Communication with Licensees
NISA, for mutually promoting the understanding between NISA and licensees on the policy ofregulations, is trying to facilitate the opportunities to exchange opinions based on thetransparency.
•

In order to build mutual trust with licensees and to make a smooth communication at
inspection sites, the "Handbook for Inspectors", which describes the inspector's rules, isdistributed to all of NISA inspectors to be carried always with them.

•

Licensees, JNES and NISA forms the "Project Team for Operational Improvement of theInspection System" and is making efforts to make efficient and steady use of the newinspection system.
• The Director General of NISA, presidents of licensees, etc. have opportunities toexchange opinions freely and openly on the safety situation of licensees’ nuclear
installation and their future tasks from the viewpoint of quality assurance. Moreover
NISA staff visits the nuclear power station, and explains the current trend of safety
regulation and the concept behind to the field operators and exchange opinions withthem in order to promote the morale among the organizations and people concerned.
• In order to share nuclear related safety information concerning accidents, failures etc.
of nuclear installations, "Regular Meeting on Safety Management of Nuclear Power
Stations" is held about once every two months as an opportunity for opinion exchangeon the safety management among NISA, JNES, electric utilities, the Federation ofElectric Power Companies, and Japan Nuclear Technology Institute.

9-4

C. General Safety Considerations

This page intentionally left blankArticle 10 Priority to Safety
Each Contracting Party shall take the appropriate steps to ensure that all organizations
engaged in activities directly related to nuclear facilities shall establish policies that givedue priority to nuclear safety.
The development and utilization of nuclear energy has been promoted giving due priority tosafety in accordance with the Atomic Energy Basic Law.
The JCO criticality accident in 1999, etc. showed the importance of moral and education of
the employee in the organization, and resulted in the introduction of the Operational Safety
Inspection system and the allegation system by the employee.
However, TEPCO falsification issue revealed in August, 2002, showed, again, that negligenceof priority to safety among personnel gave rise to organizational falsification, and this led to
the renovation of the safety regulation by NISA. In the renovated regulation for organizationand management, it has introduced the quality assurance system into the licensee and haveestablished it in the regulation. The inappropriate quality assurance system and maintenancemanagement activities were causes of the Secondary Pipe Rupture Accident at Unit 3 of theMihama Power Station that occurred in August, 2004, and it was pointed out that thedeterioration of the "safety culture" lurked in it’s background. As the quality assurance andmaintenance management activity are positioned in the center of the corporate culture andorganizational climate concerning the nuclear safety (safety culture), the regulatory bodycomprehends licensees’ efforts for preventing deterioration of the corporate culture andorganizational climate at the periodic safety review, and is promoting their efforts bypositively encouraging good practices etc. On the other hand, licensees etc. reported to NISA
about the measures to prevent recurrence of about one hundred (100) cases of information
falsifications and procedural deficiencies in April 2007, and aiming at ensuring nuclear
safety, they strengthened their efforts for systematic improvement in the safety culture andstarted its enhancement. NISA, receiving the reports, is taking measures relevant to thepriority to safety at the licensees.
10.1 Basic Policy for Priority to Safety
Priority to safety is a basic policy in all nuclear energy development and utilization in Japan.
Article 2 of the Atomic Energy Basic Law states that priority should be given to ensure safety
in all related activities.
Also, Article 1 of the Nuclear Regulation Law states that “this law, in accordance with theAtomic Energy Basic Law, is enacted for the purposes of providing necessary regulations on
the establishment and operation of reactors, in order to ensure that the use of atomic energy islimited to peaceful purposes and carried out in a planned manner, and at the same time, toensure the public safety by preventing the hazards due to these materials and reactors.”

10-1

10.2 Efforts for Improvement in Safety Culture
Safety culture in the organization is so vital to ensure the safety of the nuclear installation,that the lack of safety culture there may result in serious consequences.
Each regulatory body and licensee makes diverse efforts to establish safety culture.
(1) Efforts by Nuclear Industry
1) Policies of the whole nuclear industry
Japan Atomic Industrial Forum Inc., consisting of about 480 business operators(electricity utilities, reactor manufacturers etc.) who are engaged in the nuclear business,
established a "Charter for Safety by Nuclear Power Industry" in October 2006. The
objectives of establishing the “Charter”, and the main text of the “Charter” aredescribed below. The top managements of all organizations are obligated to takenecessary measures so that the "Charter" penetrates to the all fronts of each organization,and is practiced positively as a voluntary and continuous effort, and is aimed to thelong-term continuation of safety achievement.
Objectives to Establish the "Charter"
Technologies for peaceful use of nuclear energy in Japan is on a globally high level, but
accidents and troubles which have occurred in the nuclear industry have affected thesocial confidence of the nuclear power industry. In order to be trusted by society withpublic confidence it is required of every person engaged in the nuclear industry to have asense of pride and a sense of responsibility, to raise the consciousness of "not causing
any accident by any means", and to establish the safety by taking action. The "Charter" isestablished as an action agenda accordingly.
Main Text of the "Charter"
Article 1
We have a sense of responsibility and an awareness of its duty, we give the priority toensuring safety over all, no matter what it may be in what status,
Article 2
We aim at thoroughness of safety measures by learning modesty from past faults andsharing safety information.
Article 3
We make effort to develop good working environments, where matters perceived to beunsafe can be discussed at any time, are produced.
Article 4
We always keep a "questioning attitude", without being self-conceited with good safety10-2achievements.
Article 5
We positively release error information as well as we listen sincerely to the voice ofsociety.
In addition, although the Charter is conducted by each member's independent effort, theJapan Atomic Industrial Forum Inc. has performed activities to promote theestablishment of the Charter by the visiting local governments by the President, visiting
members’ offices to purport explanation, the presentation of each member's independent
efforts at the member's liaison councils, etc.
On the other hand, in April 2005, the nuclear industry (including nuclear power
operators, nuclear fuel fabrication facility operators, plant manufacturers, etc.)
established the Japan Nuclear Technology Institute( JANTI), who inherits and enhances
the functions of the network of organizations for sharing and in the improvement of thesafety culture, "Nuclear Safety Network (NS Net)" and the Nuclear Information Center ofCentral Research Institute of Electric Power Industry and it has new additional
functions to study and develop standards and codes, aiming at further improvement in the
self-controlled operational safety activity of the nuclear industry. The activities
concerning the improvement in the safety culture of this association is as follows;a. Safety-culture dissemination activities••

Holding seminars concerning safety and lecture meetings, opinion exchange
meetings concerning safety for persons at licensees’ sites. Moreover, theactivities are opened to the public for transparency.
Investigating and studying trends of safety culture in and outside Japan, andplanning to support the licensees' self-controlled activities for safety culture.

b. Peer-review activities•

Cooperating with the Institute of Nuclear Power Operation (INPO) who hasabundant peer-review achievements in the U.S. and the World Association of
Nuclear Operators (WANO) who is developing the international peer review. It
has been enhanced from the conventional review of just confirming documents to
the review of focusing on field activities. Moreover, making effort to obtain goodforeign practices and making an international contribution, by dispatching
personnel to the peer reviews of WANO and the IAEA.

•

Disclosing peer-review results, aiming at the formation of social consensus.

c. Effective utilization of information
Utilizing the open library "NUCIA" on nuclear power generation and overseas
information, JANTI analyses and evaluates the information and provides the results to10-3

the licensees at the periodically held "Study Group on Operational Information" andat peer reviews. Licensees are supporting INPO, WANO etc. in exchange of
operational experience information with overseas as well.
d. Safety-culture assessment activities
Based on the efforts of fostering the safety-culture by the former NS Net, the status of
the safety culture of member's sites will be, on a basis of questionnaires, assessed by
JANTI as the third party, to support the self-controlled activities for fostering
member's safety-culture. These assessment activities will start in the 2007 fiscal year
as a trial and will be applied in a practical way in the 2008 fiscal year.
2) Policies of Licensees
All licensees have declared their principles to give due priority to nuclear safety at
nuclear installations, and have tried hard to improve not only in the safety culture but
also the corporate ethics or quality assurance. Under the policy to give priority to safety,
each licensee started system development so that the top management (president)
participates in ensuring safety under his direct responsibility.
But in July, 2006, NISA judged that the quality assurance system might not be functioning
sufficiently at Tohoku Electric Power Co., Inc., and directed the company to perform theintegrated check of the quality assurance system (refer to Section 13.2). Moreover, after
discovering data falsification at a hydroelectric power plant of the Chugoku ElectricPower Co., Inc. in October 2006, investigation at sections including thermal power andnuclear power was conducted, and of 316 cases in total, 98 cases of falsifications andprocedural deficiencies at the nuclear power sections, were discovered. (Refer to Section
6.2)
The causes were judged by the companies that efforts till then by the electric powercompanies, who are licensees, were not pervading thoroughly to the job sites, and the
support by top managements and managers were insufficient to lighten a burden of site
stuff. After discussing about the prevention of recurrence at the "Reliability Recovery
Committee" of the Federation of Electric Power Companies in March 2007, the electric
power companies, reexamined the action agenda of the Federation, and in May of thesame year, presented an action plan for prevention of recurrence to NISA, which includes
participation of the top management, thoroughness of training and education of
personnel, enhancement of sharing safety information, and this started the reconstruction
and fixing of the safety culture.
(2) Effort of the National Government
Although the safety culture should be developed in organizations of licensees, who take theprimary responsibility in the safety operation of nuclear installations, the national
government has appropriate attention on licensees' fostering safety culture and promoting it.

10-4

1) Efforts of NISA
As it is important for a manager of an organization to pervade a sense of value by giving
top priority to safety among site staff, regulatory bodies are required to look at the natureof the licensee’s management and promotion of safety culture. The regulatory bodies areencouraging the licensee's managements so as to make them incorporate the quality
management system and safety culture, for the time being, and they are requiring
licensees to follow their quality management systems strictly. As part of this activity,
NISA is designing institutional arrangements to clarify regulatory requirements
concerning the quality assurance so that the licensee's quality assurance system isestablished firmly. The establishment of the quality assurance system is described in
Article 13.
NISA together with JNES is, in order to promote the licensee's safety culture, taking thefollowing measures;•

Development of the quality assurance system
From October 2003, the licensee's quality assurance system is provided in the
Operational Safety Program, and it is verified during the Operational Safety
Inspection etc. that the quality assurance activities are functioning properly. Refer tothe Article 13•

Deterioration prevention of the corporate culture and organizational climate
Since the corporate culture and organization culture are fundamentals of various
activities for ensuring safety, licensees are taking measures for deterioration
prevention and the performing of the self-assessment at the periodic safety review.
NISA perceives the licensees' efforts during the periodic safety review and is
promoting their efforts by positively encouraging good practices as follows.
a) Viewpoints
NISA perceives the licensee's efforts from the following viewpoints.
(i) Effectiveness of the efforts for detection and prevention of organization climate
deterioration(ii) Self or external assessment on the effectiveness of the efforts for detection andprevention of organization climate deteriorationb) Encouragement(i) Especially effective efforts for detection and prevention of organization climate
deterioration(ii) Especially effective efforts to enrich measures for aging management in theorganizational climate. The viewpoints to be used to perceive efforts developed

10-5

by JNES are provided in Table 10-1.
•

Evaluation of safety culture
NISA together with JNES is developing a guideline to assess the licensee's safety
culture at the Operational Safety Inspection. In the development of this guideline,
IAEA's publication (INSAG-4 "Safety Culture", "ASCOT Guideline", etc.), ISO 9001
(2000) and foreign examples are referred to.
Note: ASCOT: Assessment of Safety Culture in Organizations Team•

Root cause analysis of accidents and failures
When an accident or a failure occurs, it is necessary to clarify not only the directcauses, but also the root causes taking account of their organizational factors.
Licensees are required to implement systematic and lasting measures. From theviewpoint of making licensee's root cause analyses effectual, a guideline for root
cause analysis is being prepared in the system of the quality assurance standards ofJapan Electric Association. NISA together with JNES is developing a guideline toassess effectiveness of the licensee's root cause analysis.

•

Comprehensive check of power generation facilities
Based on the comprehensive check (refer to Section 6.2) of power generation facilities,NISA decided to take measures for licensees to give due priority to safety furthermore.
These matters will be reflected in related legislations, etc.
(i)

Clarification of a compliance system in the Operational Safety Program
The following will be added as matters to be provided in the Operational Safety
Program.
・ Matters relating to the compliance system,
・ Matters relating to the system for fostering safety culture,
・ Matters relating to the root cause analysis, and・ Matters relating to the information reporting important-to-safety.

(ii)

Addition of measures for Safe Operation
・ To develop work procedures properly and to perform operational safety
activities following the procedures, and・ To take actions for the required procurement control in order to be able toshare manufacturer's information about safety technologies among
licensees.

(iii)

Development of an organizational system in which independency of the Chief
Engineer of Reactors is secured2) Efforts of the NSC

10-6

•

First-Series Roundtable Discussions on Safety Culture
The NSC, as one of measures taken after the JCO nuclear criticality accident
occurred in September 1999, held the " First-Series Roundtable Discussions on Safety
Culture " with unit managers and shift supervisors of twenty one (21) nuclear
facilities in Japan from July 2001 to December 2003. The contents were compiled andpublished in a document "Site interviews about Safety Culture –Discussions on sites
where the safety should be assured -" (January 2004). The summary is as follows;(a) Findings brought from the sites for developing safety culturea. Educate the staff to have their own pride and responsibility without hiding in an
organization.
b. Educate the staff to develop common sense and morals so that they can recognize
the necessity of safety by their own values.
c. Mistakes are not for learning about people’s faults. Mistakes are lessons forpreventing problems from occurring again and for assuring safety. It is necessary
to recommend staff members to report mistakes and managers should not punish
them for it.
d. Learning does not always consist of sitting and learning. There is actual
experience and learning by simulation.
e. You can find opportunities everywhere on-site for systematic training, lessons forpreventing problems from reoccurring, and for other things.
f. Successful experiences do expire. Do not adopt everything at once. Be sure toconfirm and consider them since some successful experiences may be out of dat.
g. Introduce knowledge in quality control and other areas to business management.
h. Be sure to cope with all circumstances, imagine every possible situation and makeit clear who is responsible for making the decisions.
i. There is no purpose in collecting information alone. It is more important whocollects it and how it is used.
j. Learn to see yourself and your organization from a broad point of view.

(b) Important topics brought from the sites regarding safety culturea. Educate every staff member to comply with the corporate morals.
b. Use adverse circumstances and pressures ( You can keep tension by the strict gazeof the surrounding society).
c. Communicate more with the media than usual and proceed with transparency of
information.
d. Educate every staff member to be accustomed to being conscious of the complete

10-7flow as a whole.
e. Let them realize that reporting and consulting are the best steps for assuringsafety.
f. Introduce engineer morals and morals for work accomplishment into educationand training programs.
g. Educate all staff members to have a questioning attitude about the safety of theirwork, actions, and activities.
h. Make the most use of knowledge in the general analysis system that supports thepassing on technology and the high technologies that explain vagueness.
i. Always appreciate the appropriate allocation of human resources. (Topmanagement)
j. Reinforce the quality assurance system and concentrate on software quality
assurance.
k. Maintain and improve the ability of the staff in accordance with the increase rateof direct constructions.
l. Sharing the values of safety with cooperating companies is a major key fordeveloping safety culture.
m. Use the latest knowledge about risk communication to communicate with outside
stake-holders.
n. It is necessary to observe our actions both from the subjective and objective
viewpoints and to ask ourselves if they are appropriate.

•

Second-Series Roundtable Discussions on Safety Culture

The NSC held a series of “Roundtable Discussions on Safety Culture”, from October
2004 to April 2005, as one of the measures after the secondary pipe rupture accident at
Unit 3 of the Mihama Power Station that occurred in August 2004, in order to exchange
opinions with top managers of nuclear power companies and major contractors. The
contents were compiled and published in a document "Fostering a Culture of Safety inJapan’s Nuclear Industry – Exchange of Views with Top Management" (June 2005), Thesummary is as follows.
(a) Management Safety Alertness and Activities
To conduct activities of nuclear energy, everybody in the organization must shareand implement the practices of valuing the concept of “safety first”, by constantly
questioning whether the current practices of activities are appropriate from theviewpoint of ensuring safety. To this end, the top management should takeleadership in such areas as organizational composition, resource allocation,
quality assurance system, technology, human resources and training.

10-8

(b) Productive Communications between Site Staff and Upper Management
To give safety-ensuring activities substantial significance, it is crucial thatsafety-related information smoothly permeates throughout the organization, andthat a system and means be provided to enable it. With full awareness of the
difficulties involved in creating smooth communications, management must
constantly and intentionally motivate their staff to improve the situation by
ensuring the bi-directional information channels, and appropriate and timely
remedial actions.
(c) Workplace Environment
Regulators and licensees must continue to make bilateral efforts to improve theeffectiveness of regulatory activities for productively improving safety assurance,
without being content with the formalities in meeting regulatory standards ineffect. It is necessary that the management of licensees and contractors will
maintain and promote a thoroughgoing cooperative relationship, while sharing astrong perception that ensuring safety is the prerequisite to everything else innuclear activities, and that it be the most efficient means of cost optimization.

10-9

Table 10-1 Viewpoints to Comprehend the Efforts to Prevent Deterioration ofLicensees' Organizational Climate
Points to confirm symptoms of theorganizational climate deterioration
1. Top management's commitment

Viewpoints of Comprehension
Pervasion of messages for priority to safety to
the tips of organization.

2. Senior management's clear policies
and behavior

Presentation and behavior of the policies of
ensuring safety

3. Improvement and fixing of the quality
management system (QMS)

Feedback of the knowledge obtained from thenonconformity management to QMS
improvement
System, encouragement and utilization ofreportingOperating experience feedback, efforts ofmaintaining and improving technicalcapabilities, communication in the operationalsafety activities
Efforts to improve in-company communication,communication with contractors

4. Reporting culture
5. Learning organization

6. Workplace with good communication
7. Exclusion of decision-making by amisjudgment
8. Compliance with rules
9. Accountability and transparency

10. Self (or 3rd party) assessment

Preventive measures for eliminating adecision-making by a misjudgment
Maintenance and management of rules, fixingof routine work
Timely information service, promotion ofmutual understanding among local residents
and regulatory bodies, improvement intransparency
The self-assessment (the 3rd party) method forpreventing activities to be just a formality

10-10Article 11 Financial and Human Resources
1. Each Contracting Party shall take the appropriate steps to ensure that adequate
financial resources are available to support the safety of each nuclear facility throughout
its life.
2. Each Contracting Party shall take the appropriate steps to ensure that sufficientnumbers of qualified staff with appropriate education, training and retraining areavailable for all safety-related activities in or for each nuclear facility, throughout its
life.
The financial basis of electricity utility rests on the understanding and recognition that
nuclear energy is the environmentally clean energy and reliable source for base load power,u nder the pressure of reduction of power rate due to the deregulation of electricity
utility industry.
In Japan the sufficient numbers of qualified staff are available through imposing requirements
of appointment of chief reactor engineers, persons responsible for operation, chief electricalengineers, etc. upon licensees and assessing the technical competence of the licensees. After
more than 30 years of operational experiences, many experienced employees are now retiring.
The licensees face the challenge of succession of technology and ensuring human resources.
The recruitment and training of personnel in various fields are now in progress.
11.1 Financial Resources of the Licensee(1) Confirmation at Issuing License
Before issuing license of a nuclear installation, the Minister of METI, in accordance with
Article 24 (Criteria for the license) of the Reactor Regulation Law, confirms that the applicant
for the license possesses necessary financial basis by requiring the applicant to submit“Amount of Funds Required for Construction and Finance Procurement Plan”, and also
consults with the AEC. (Refer to section 7.3 and Fig. 7-3)
(2) Applicant for the License of Nuclear Installations
Applicants for license of commercial power reactors are the general electric utilities, that is, 9
electric power companies and 2 wholesale electric power companies. The Minister of METI
issues license for electricity utility business only to those meeting certain criteria of financial
basis, technical capability, etc.
The nuclear power generation is recognized as a superior power source with characteristics
of reliable supply and useful energy to cope with global warming, and measures, such as
performing a preferential load dispatching, are taken. Such utilization of the nuclear energy
ensures stable flow of income for electric utilities.

11-1

On the one hand, since back end businesses, such as reprocessing of spent fuel generated by
nuclear power generation, requires large amount of expense over extremely long period oftime, in accordance with the "Law for Reserving and Management of Reserve Funds for
Reprocessing Spent Fuel in Nuclear Power Generation, etc." enacted in 2005, the electricutilities reserve funds for the expense in advance. As for final disposal of high level
radioactive wastes, the Nuclear Waste Management Organization of Japan, an implementing
organization for disposal, will perform geological disposal in accordance with the Law
concerning Final Disposal of Specific Radioactive Waste enacted in 2000, using funds
reserved by electric utilities etc. Furthermore, METI enacted the Ministerial order of Reserve
Fund for Dismantling Nuclear Power Facilities in accordance with provision of Article 35 of
the Electricity Utilities Industry Law, and the electric utilities deposit reserves fordecommissioning on the basis of this order.
On the other hand, the Japan Atomic Energy Agency, who owns R&D reactors of Monju andFugen, is established by a law, and financial basis necessary for its business operation isprovided by the national budget.
11.2 Human Resources concerning the Nuclear Installation of the Licensee(1) Confirmation of Technical Competence
Before issuing license of a nuclear installation, the Minister of METI confirms that theapplicant possesses technical competence necessary to establish a nuclear installation andoperate it adequately, and consults with the Nuclear Safety Commission (the NSC). The NSC
had established the “Regulatory Gude for Reviewing Technical Competence of Nuclear
Operators” on May, 2004, and based on this examination guide, the applicant's technicalcompetence for the following items is examined.
The examination items of technical competence;
1. Organization for design and construction,
2. Ensuring engineers for design and construction,
3. Experience related to design and construction,
4. Quality assurance activities concerning design and construction,
5. Organization for operation and maintenance,
6. Ensuring engineers for operation and maintenance,
7. Experience related to operation and maintenance,
8. Quality assurance activities concerning operation and maintenance,
9. Education and training for engineers,
10. Designation and staffing of qualified personnel etc.
The licensees are responsible for safety of the decommissioning and for preparing personnel
for it. In fact, the licensees have trained and prepared human resources and implemented
technological development programs through the decommissioning and verification test on a
research reactor (JPDR) in cooperation with national organizations, manufacturers andconstruction companies, and implemented decommissioning for the Tokai Power Station of the

11-2

Japan Atomic Power Company.
(2) Qualification, Training and Retraining of Personnel Engaged in Safety Activities
1) Staff Qualification
The licensees shall appoint a Chief Reactor Engineer to supervise safety operation of
nuclear installation, a Chief Electrical Engineer and a Chief Engineer of Boiler andTurbine to supervise safety during construction, operation and maintenance of electricfacilities. The licensee assigns the Persons Responsible for Operation from those whohave knowledge, skills, and experience required for operation of reactors, and who satisfy
requirements provided by the Minister of METI. As a method to judge conformity torequirements of a person responsible for operation, it is under study to incorporate an
accreditation system utilizing an independent organization. The person responsible for
operation observes operation in general and supervises operators. The numbers of Chief
Engineers of Reactors and Persons Responsible for Operation are 1254 and 421
respectively at the end of June, 2007.
Their duties are explained in Section 19.3 in detail.
As for the qualification of staff engaged in the dimensional measurement of defect in thefitness-for-service assessment of equipment, PD (performance demonstration)
qualification system for the ultrasonic flaw detection system (including the qualificationof staff) is employed by the Japanese Society for Non-Destructive Inspection, and 14 PD
technicians are certified (as of November 2006).
2) Staff Training and Retraining and Resources for Training
Licensees shall integrate education on operational safety of personnel in charge of
operation and management of a nuclear installation into the Operational Safety Program,and prepare and carry out long-term and short-term staff training programs to maintainand improve their skills and capabilities. Licensees, in addition to in-house operator
training course using simulators (Table 11-1); periodically send their operators toexternal operation training centers for retraining. There are two centers: the BWR
Operation Training Center (BTC) for BWRs and the Nuclear Power Training Center(NTC) for PWRs. A curriculum suitable for the ability/skill of each operator is preparedin these training centers.
Each licensee has established maintenance training centers (Table 11-2) for education andtraining of maintenance personnel. Various mock-up devices, inspection devices andtraining devices, etc, simulating plant facilities for training purposes, have been used tomaintain and improve the knowledge, skills and work management capabilities of
personnel involved in maintenance and inspection.

11-3

11.3 Efforts for Ensuring Infrastructure of Human Resources in the Regulatory Bodies(1) Training of Experts in NISA
Staff members, who are in charge of nuclear regulation in the Nuclear and Industrial Safety
Agency (NISA), are the Senior Specialist for Nuclear Emergency, the Nuclear Safety
Inspector, the Nuclear Facility Inspector, the Electric Facilities Inspector, and the Safety
Examiner. These are called "Nuclear Regulatory Staff" as shown below.
A Senior Specialist for Nuclear Emergency is stationed at each nuclear installation, guides andadvises the licensees in preparing its Plan for Emergency Preparedness, and conducts duties
necessary to prevent progression of nuclear emergency should it occur.
A Nuclear Safety Inspector is stationed at each nuclear installation, conducts the Nuclear
Safety Inspection to confirm licensee’s compliance with the Operational Safety Program,
address incidents if they occur, and supervises operation management of a nuclear installation.
A Nuclear Facility Inspector and /or an Electric Facilities Inspector is dispatched from NISA
head office, and conducts inspection activities, such as the Pre-Service Inspection and the
Periodic Inspection of a nuclear installation, and the Fuel Assembly Inspection, on the basis of
the Reactor Regulation Law or the Electricity Utilities Industry Law, respectively.
Safety Examiners conduct the Safety Examination of a nuclear installation.
A Nuclear Regulatory Staff is required to have expertise in nuclear technology. The system oflong term and multistage education and training programs necessary for improvement of
his/her expertise is developed, taking account of his/her experience and of the nature of the
facility to which he/she is assigned. Moreover, NISA started a Special Training Course on
Quality Assurance of Nuclear Installation in 2002.
In order to increase effectiveness of the training, the contents of the training being
implemented are also reviewed and improved suitably. The disposition of the personnel
engaged in securing safety of nuclear installations is improved through these training.
Summary of training for nuclear safety regulation is shown in Fig. 11-1.
NISA has appointed six Special Inspection Instructors in December 2003. They advise
inspectors for the Nuclear Safety Inspection, the Periodic Inspection, etc. in each power
station, instruct them to equalize the levels of inspections, and collect opinions and proposals
from inspectors and licensees for the purpose of opinion exchanges at the site.
Furthermore, NISA maintains and develops its regulatory capability, as well as contributes tointernational safety regulation, through exchange of technical experts and information on
safety regulation and safety technology, under bilateral arrangements with foreign regulatory
bodies and in the framework of multilateral cooperation (the IAEA and the OECD/NEA).
Moreover, besides training professional human resources as mentioned above, NISA recruits

11-4

professional human resources for nuclear safety from industries or other ministries andgovernment offices, and is making efforts to ensure infrastructure of human resources.
In the future, the training curriculum that makes staff be able to acquire the required
regulatory skills effectively will be established, and a training management system that keeps
and manages every personnel's training status and offers appropriate training at suitable time
will be developed. Moreover, measures, such as preparation of training facilities for trainingof practical inspection skills etc. through with mock-up will be taken.
(2) Training of Experts in JNES
Japan Nuclear Energy Safety Organization (JNES), as well as NISA, develops training courses
for its personnel, putting emphasis on inspection activities.
JNES’s inspection activities include the Electric Facilities Inspection, the Nuclear Facility
Inspection, the Welding Inspection, the Audit of Licensee's Periodic Check System, the Audit
of Licensee’s Welding Check System, the Safety Confirmation of Disposal Facility, the Safety
Confirmation of Radioactive Waste Package, the Confirmation of Transportation Packaging,and the Confirmation of Transportation Method. The Electricity Utilities Industry Law or theReactor Regulation Law stipulates that each of these activities be conducted by qualified
personnel. JNES prepares various training courses for staff members to get appropriatequalification in their respective activities. President of JNES assigns inspectors from thosequalified persons.
JNES encourages inspectors and examiners to be qualified in the disciplines related to theirduties. Moreover, JNES encourages inspectors and examiners to participate in school ofexternal bodies, scientific seminars etc. to enhance their expertise.
11.4 Maintaining Human Resources in Nuclear Fields in Japan
In order to ensure safety of nuclear power generation, highly capable human resources should
be maintained and ensured. In Japan, it has been an issue to keep human resources inappropriate condition, because of low birthrate and increase of elderly people, decrease inpopulation, retirement of skillful engineers, and decrease of construction opportunities ofnuclear installation.
(1)

Efforts by National Government

NISA has been studied strategic measures for training and ensuring human resources ofnuclear specialists at the Nuclear Safety Infrastructure Subcommittee under the Nuclear andIndustrial Safety Subcommittee established in 2006. Based on the availability of external
advisory specialists and the emphasis on ensuring safety in nuclear facilities and on
appropriate safety regulation, the subcommittee studies the strategy for training and ensuringof human resources with the understanding that it is necessary to clarify the technical fields(specific and basic) in which human resources should be allocated.

11-5

In parallel with this study, the Ministry of Education, Culture, Sports, Science and Technology
and the Ministry of Economy, Trade and Industry will implement the nuclear human resource
training programs focusing on the following items from 2007 fiscal year;i) Support of educational activities, such as basic nuclear education and research, internship,
and preparation of core curriculum for the nuclear power,ii) Implementation of nuclear human resource training programs for research activities
focusing on basic and infrastructure technology fields supporting the nuclear power, fortraining of human resources to ensure research successors.
(2) Efforts by the Nuclear Industry
The nuclear industry has grave concerns in the succession of technology, expertise andexperiences between the generations. The first generation experts who had many experiencesin commissioning test, operation, maintenance, and trouble shooting in abnormal events are in
the age of retirement.
Main activities currently performed for human resource training and technology succession in
the nuclear industry are shown in the following;
1) Training of on-site technicians and succession of skills
In the area where the nuclear installation is established, the training for qualification,
training of practical skills for maintenance and repair, OJT training at the power station, etc.
are implemented beyond the frame of an individual firm.
2) Study on qualification and certification system for private sectors
For the purposes of improved skills of maintenance-and-repair workers, appropriate staffing,and ensuring future human-resources, the common standards and qualification / certification
procedures for objective evaluation of skill level are being studied. These standards etc. will
be made to harmonize with the licensee's in-house qualification system.
3) Acquisition of advanced expertise
Licensees’ engineers acquire the education in the graduate school related to the nuclear power,
thus engineering specialists with advanced expertise are fostered.

Moreover, the Japan Atomic Industrial Forum, Inc., consisting of enterprises related to the
nuclear power has proposed policies as a private sector, in order to improve the effectivenessof "the nuclear human resource training program. It has investigated and compiled the current
state, issues of schools, such as universities, graduate schools and research organizations, for
improvement for human resource training. The study is continued.

11-6

(3) Efforts by Universities and Research Institutes
Recent years, it has been a trend to reorganize and to unify faculties or graduate schools. In
this trend, conventional nuclear department and majoring were unified with other ones, andtheir names were changed. And they were reorganized to departments and majoring that
address wider fields, including energy, environment etc. However, in recognition of animportance of nuclear education, for the purpose of training for engineering specialists withpractical capabilities and engineering theories in the nuclear field, the faculty, graduate
school and professional school have been established in nuclear research and development
complex, such as Fukui Prefecture and Ibaraki Prefecture.
Moreover, some research organizations and graduate schools incorporate a cooperation
system of graduate schools. Through the system, the facilities/equipment and human resourcesin the research organizations are shared, the contents of education/study at the graduate
school become more in details, the communication between researchers is promoted, and theeducation at graduate schools is activated. The Atomic Energy Society of Japan has founded asenior network (SNW) whose members are retired seniors from nuclear related organizations.
Senior people plays a central role in dialogues with college students and in other activities at
SNW to make succession of nuclear technologies to young people and students who areresponsible for next generation and to spread correct understanding of the nuclear power.
(4) Establishment of Professional Engineers System for Nuclear and Radiation Technologies
The Ministry of Education, Culture, Sports, Science and Technology who has jurisdiction overthe Professional Engineers System established a nuclear and radiation technology department
for professional engineers in 2004 fiscal year. The qualification examination has been
implemented every year, and a total of 153 people were qualified as the Professional Engineer
by the end of 2006 fiscal year. The purposes of the Professional Engineers System are such toenhance nuclear engineering capabilities, to utilize the capability in the nuclear safety
regulation, to strengthen the safety management system in each corporation.

11-7

Table 11-1

Organization

Operator Training Facilities of Nuclear Installations

Location

Simulator

BWR Operator Training Okuma-machi, Futaba-gun, Fukushima
Full scale; 3 units
Center Corp.
Prefecture
Full scale; 2 units
Kariwa Village, Kariwa-gun, Niigata
Prefecture
Nuclear Power Training Tsuruga, Fukui Prefecture
Full scale; 3 units
Center Ltd
The Japan Atomic
The Japan Atomic Power Company Training
Compact; 1 unit
Power Co.
Center (Tokai Village)
On site of Tsuruga Power Station
Compact; 2 units
Hokkaido Electric
On site of Tomari Power Station
Full scale; 1 unit
Power Co., Inc.
Tohoku Electric Power Nuclear Power Engineering Training Center
Full scale; 1 unit
Co., Inc.
(on site of Onagawa Nuclear Power Station)
Nuclear Power Engineering Training
Full scale; 1 unit
Building(on site of Higashidori Nuclear Power
Station)Tokyo Electric PowerOn site of Fukushima Daiichi Nuclear Power Full scale; 1 unit
Full scale; 1 unit
Co., Inc.
StationFull scale; 1 unit
On site of Fukushima Daini Nuclear Power
StationOn site of Kashiwazaki Kariwa Nuclear Power
StationChubu Electric Power
Nuclear Power Training Center (on site ofFull scale; 2 units
Co., Inc.
Hamaoka Nuclear Power Station)
Hokuriku Electric
Nuclear Power Engineering Training Center
Full scale; 1 unit
Power Co.
(on site of Shika Nuclear Power Station)
The Kansai Electric
On site of Mihama Power Station
Compact; 1 unit
Power Co., Inc.
On site of Takahama Power Station
Compact; 1 unit
On site of Ohi Power Station
Compact; 1 unit
The Chugoku Electric
Ohno Training Center (Ohno-machi)
Full scale; 1 unit
Power Co., Inc.
Shikoku Electric Power Nuclear Engineering Training Center
Full scale; 1 unit
Co., Inc.
(Matsuyama)
On site of Ikata Power Station
Kyushu Electric Power Nuclear Power Training Center (on site ofFull scale; 2 unit
Co., Inc.
Genkai Nuclear Power Station)
Nuclear Power Training Center (on site ofFull scale; 1 unit
Sendai Nuclear Power Station)
Japan Atomic Energy
On site of Fugen Power station
Compact; 1 unit
AgencyOn site of Monju Construction Office
Full scale; 1 unit
(As of the end of March, 2007)11-8

Table 11-2 Maintenance and Repair Training Centers of Licensees

Organization

Name

Location

The Japan Atomic
The Japan Atomic Power
Tokai Village, Naka-gun,Power Co.
Company Training Center
Ibaraki Prefecture
Hokkaido Electric
Nuclear Power Training Center On site of Tomari Power Station
Power Co., Inc.
Tohoku Electric Power Nuclear Power Engineering
On site of Onagawa Nuclear
Co., Inc.
Training Center
Power Station
Fukushima Nuclear PowerOn site of Fukushima Daiichi
Nuclear Power Station
Tokyo Electric Power Plant Training Center
Co., Inc.
Kashiwazaki Kariwa Nuclear
On site of Kashiwazaki Kariwa
Power plant Training Center
Nuclear Power Station
Chubu Electric Power Nuclear Power Training Center On site of Hamaoka Nuclear
Co., Inc.
Power Station
Hokuriku Electric
Nuclear Power Engineering
On site of Shika Nuclear PowerPower Co., Inc.
Training Center
StationThe Kansai Electric
Nuclear Power MaintenanceTakahama-cho, Ohi-gun, Fukui
Power Co., Inc.
Training Center
Prefecture
The Chugoku Electric Shimane Nuclear Power Station On site of Shimane Nuclear
Power Co., Inc.
Power Station
Engineering Training Center
Shikoku Electric
Nuclear Engineering Training Matsuyama City, Ehime
Power Co., Inc.
CenterPrefecture
Genkai Nuclear Power Station On site of Genkai Nuclear
Kyushu Electric Power Nuclear Power Training Center Power Station
Co., Inc.
Sendai Nuclear Power Station On site of Sendai Nuclear
Nuclear Power Training Center Power Station
Japan Atomic Energy General Training Facility for
On site of International Nuclear
AgencyFBR Cycle
Information/Training center(As of the end of March, 2007)

11-9

Fig. 11-1 Training on Nuclear-Safety Regulation
Training on nuclear safety regulation
Commercial power reactor

R&D stage reactor-Risk communication training for managers

Cross-cutting training
Nuclear fuel cycle facility- Public-relations training for Nuclear Safety Inspectors

Nuclear emergency preparedness,
Crisis management- Quality Assurance training

Meister
Senior expert

11-10- Nuclear emergency
preparedness, Advanced- Nuclear emergency
preparedness, on-site training- Off-site center desk-top drill- Emergency preparedness and−Nuclear power- Nuclear power generation
response(FBR) for experts
generation (BWR, PWR) for- FBR sodium technical- Off-site center management
experts
training- Off-site center functional groupt i i
Nuclearemergency- Special training course on QA of nuclear installation- Special training course on QA of nuclear
preparedness,Basicinstallation,
Follow up- Nuclear officers training

Expert- Nuclear Safety Inspector basic training-Electric Structure Inspector(nuclear power) training- Nuclear Facility Inspector basic training-Nuclear power station risk assessment technology-Nuclear reactor safety design, basic- Overseas training
Entry- Radiation safety- Basic Safety Regulation-Participation to the various basic lectures by the Japan Atomic Energy Agency

Article 12 Human Factors
Each Contracting Party shall take the appropriate steps to ensure that the capabilities
and limitations of human performance are taken into account throughout the life of a
nuclear installation.
The licensee takes human factors into consideration at the design stage of nuclear installation,
and, at the operation stage, prepares operational procedures, education and training course forits personnel and the management system for operation and maintenance. The regulatory bodyalso takes various steps for prevention and correction of human errors at design and operation
stage.
Since the last report, the requirements for prevention of misoperation in the main control
room are clarified and came to be reflected in the facility design.
12.1 Efforts by Regulatory Body
(1) Design Stage1) The Regulatory Guide for Reviewing Safety Design of Light Water Nuclear Power Reactor
Facilities requires that "the nuclear installation be designed to reflect appropriate preventive
considerations against operators’ misoperation", and its explanatory document requires that“In designing, attention should be given in consideration of ergonomics-oriented factors, topanel layout, operability of operating devices, valves, etc., instrument and alarm indication for
accurate and quick recognition of reactor status and prevention of errors during maintenanceand inspection.” and that “in designing, measures should be taken so that necessary safetyfunction is maintained without operator’s actions for a certain length after the occurrence ofan abnormal condition.” The guide also requires that “control room be designed that thesituation of operations and principal parameters of reactor and principal related facilities canbe monitored and that prompt manual control can be performed, whenever required, tomaintain safety.” In conformity to these requirements, the Regulatory Guide for Reviewing
Safety Assessment of Light Water Nuclear Power Reactor Facilities requires that “safety
analysis be performed in consideration of the following: in case that operator actions are
expected at the occurrence of abnormal situations, sufficient time and adequate information beavailable so that operator may be able to properly judge the situations and take necessary actswith a high degree of confidence.” JNES prepared a manual for evaluation of human factors in
the main control room, to confirm that these requirements are reflected in the design.
2) At the Approval of Construction Plan, the Technical Standards as a performance code
under the Electric Utilities Industry Law request that the main equipment necessary for safe
operation of nuclear installation can be monitored at a glance and necessary actions can betaken in the control room without any misopration.
3) The Nuclear and Industrial Safety Agency (NISA) and the Japan Nuclear Energy Safety
Organization (JNES) clarified the requirements for prevention of misopration, so that the

12-1

above mentioned performance code is reflected on the specification code developed as an
academic society and association standard. These requirements are shown on table 12-1.
(2) Operation Stage1) The Reactor Regulation Law provides that the licensee prepare the Operational Safety
Program, obtain approval of NISA on it and comply with it. The program includes preparationof operation management system, education on operational safety, operational procedures etc.
NISA confirms and approves the Operational Safety Program, and the resident Nuclear Safety
Inspectors confirm the compliance with it by the licensee in the Nuclear Safety Inspection.
2) The licensee reports failures of the installation to NISA in accordance with the laws.
Especially, in the case that the failures are identified to be caused by human errors, thelicensee reports to NISA measures addressing failures including improvement of equipment.
Licensees are required to correct any nonconformance of direct cause, such as a human error,
and NISA is preparing guidelines to evaluate licensees' self-supporting efforts for corrective
actions and to identify viewpoints to promote them in cooperation with JNES.
NISA also consults on the failure with experts when necessary and urges licensees to apply
lessons learned to other installations. JNES analyses human error-related cases in detail, andselects items to be reflected in the safety regulations. JNES prepares a summary report on
lessons learned, and accumulates them in the database.

12-2

Table 12-1 Requirements to prevent misoperation in a main control room
Item
1. Environmental
conditions of amain controlroom
2. Arrangement andworking space ofa main control
room

3. Arrangement ofdevices on control
panels
4. Display system(including alarmsystem)5. Control function

Requirements
Main control room should be in comfortable environmental conditions
taking into consideration temperature, lighting and noise so that
operators can operate appropriately.
(1) Consider that the following does not become too much burden at an
operator in any plant operating conditions.
(a) Duty allocation of human and machines shall be decided.
(b) Items that should be intensively supervised and operated in amain control room shall be defined, and the duty allocation
with local spots (including the panels installed at the back ofmain control panels) shall be decided.
(c) The equipment arrangement shall be designed so that
information sharing among operators to be effective.
(2) When manual operation is required for safety in a plant abnormal
condition, the operators’ working area shall be limited to the area
possible for surveillance and operation.
Alarm, display and control devices which are installed on control
panels shall be arranged with unified displaying manner so that an
operator's misoperation or erroneous recognition can be prevented.
(1) Information function(a) Information that is used for condition display of plant systemand equipment that is necessary for ensuring safety shall be
sufficient and be provided to operators being easy tounderstand at suitable position.
(b) It should be considered that communication defect or judgment
error should not arise, which is an important function ofcommunication and cooperation with on site emergency
station.
(c) Safety significant information should be displayed at theposition where operators in a control room can share among
them such information.
(2) Alarm function
When an anomaly arises in the plant equipment or process, it shallbe notified to operators so that suitable action can be made byoperators.
(3) Operation support
When an operation support system is provided, even when thesystem function is lost, plant facilities should be operable safely.
(1) The control equipment shall be easy to operate so that misoperation
becomes as small as possible.
(2) Systems and equipment controlled from a control room should bedesigned so that they cannot be operated unsafely not to impair plantsafety.
(3) During an automatic operation, operators should be able to checkthe progress of the automatic operation.

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12.2 Efforts by Licensee(1) Considerations in Design
Licensees take following considerations on human factors in designing a central control room.
The central control room is designed so that operating conditions of the reactor and otherimportant equipment and principal plant parameters can be monitored at a glance andnecessary actions can be taken in the room during normal operation and abnormal transients,
and in an accident of a nuclear installation.
For example, advanced BWRs (ABWRs) and present PWRs under design / construction are
designed re-examining the plant layout and applying computer technologies, and also adopt"advanced central control panels", which is improved in operability and parametermonitoring capability. For example, by adopting a large display screen, which is easy tooverlook power station conditions at a glance and to share information among operators, theopportunity of preventing and taking corrective action for an error is increased. Moreover,centralized supervisory operation panels, which can make operators concentrate on
observation and operation just sitting on chairs, are adopted, and automation scopes are
expanded more than those of existing nuclear installations, which reduced operator’s workloads for routine operation following a reactor scram. (See Fig. 12-1 and Fig. 12-2)
When remodeling control panels in the central control rooms of existing nuclear installations,extensive use of CRT has improved monitoring capability and operability of control panels.

Fig. 12-1 Main Control Panel of ABWR

Fig. 12-2 Main Control Panel ofLatest PWR (proto type)

A guide for digitalized main control panels was established as one of the academic society andassociation standards, JEAG 4615-2005, “Development and Design Guide for Human MachineInterface of Computerized Central Control Rooms for Nuclear Power Plants” by the Japan
Electric Association. This guide refers to related international standards and overseas

12-4requirements, and reflects past fruits of design development etc. incorporating the trend ofJapanese regulation and latest technical progresses. Specifically, requirements on functions
and designs of the central control room (information display, control and operationequipment, alarm device etc.) and standard development and design processes of the humanmachine interface are defined.
(2) Considerations in Operation Management
Licensees perform appropriate operation management during normal operation and in
accidents.
1) Operational management
a. Organizations for operation
The manager of power generation division, responsible for the operation of a nuclear
installation, controls operating shifts in charge of the operation and their supporting groups.
The shift supervisors have authority and responsibilities to take measures required in an
accident, and are selected from those who conform to the criterion specified by the Minister of
METI and have suitable experiences and suitable competence.
b. Shift of operators
Operators work in shifts. There are shifts devoted to education and training, in addition tooperating shifts, to maintain and improve operator’s capability. The education and training ofoperators is one of the important elements of human factors. Details are described in Section
11.2.
When turning over shift duties, the shift supervisor makes sure to pass on the logbook, thesupervisor logbook, keys, and precise description of operations to the succeeding supervisor.
Each operating staff also transfers information of plant operation to the succeeding operating
staff.
2) Preparation and amendment of operation procedures
Operation procedures are prepared for normal operation, failures and accidents and are
constantly amended by lessons learned from incidents and accidents or by alteration offacilities.
Symptom-based procedures for multiple failures are prepared in addition to scenario-based
procedures for design basis events. The symptom-based procedures enable prevention of
accident progression without identifying the cause of an accident. Also prepared were theprocedures addressing severe accidents exceeding design basis events, and accident
management guidelines for the staff group supporting shift operators. The effectiveness ofthese procedures is verified by comparison with the results of the analysis of plant transient by
the analysis code used in the application for licensing for establishment, and probabilistic12-5

safety assessment. Training course using simulator, based on a symptom-based procedure, isconducted at the operator training facility. Preparation of the procedures for emergency
situation is expected to be effective for mitigation of operators’ stress in an emergency.
3) Maintenance Management System
The maintenance department of a licensee controls the work of periodic inspection,
modification works, etc. of a nuclear installation carried out by the plant manufacturer andmany affiliated companies. A majority of human errors in the past occurred in the works
associated with maintenance and repair, which means that the maintenance management by
the licensee is very important.
The plant manager of a nuclear installation manages modification works, clarifying scope ofwork, scope of responsibility and authority. Maintenance of important equipment is carriedout with a prior mock-up test.
Chief engineers (Chief Engineer of Reactors, Chief Electrical Engineer, Chief Engineer of
Boiler and Turbine) perform verification and assessment of regulatory inspections byattending the regulatory inspections or confirming inspection records. They also perform
verification and assessment, as appropriate, of the plans and results of regular inspections ormodification works to prevent human errors in maintenance and management works.

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Article 13 Quality Assurance
Each Contracting Party shall take the appropriate steps to ensure that quality assurance
program are established and implemented with a view to providing confidence that specified
requirements for all activities important to nuclear safety are satisfied throughout theoperating life of a nuclear installation.

The regulatory body (Nuclear and Industrial Safety Agency, NISA), licensees, plant
manufacturers and equipment suppliers (hereinafter referred to as “manufacturers”), conduct
quality assurance (hereinafter referred to as “QA”) activities for nuclear installations in acoordinated way at each stage from design through operation and maintenance.
The basic concept of regulatory inspection has been continually renovated, seeking for more
effective and efficient inspection activities, and in line with international trend in regulatory
inspection. It has moved from a concept of system and component inspection to a concept
where NISA encourages licensees to improve their QA activities and confirms the adequacy ofthem.
A series of wrongdoings by licensees and manufacturers, such as the TEPCO falsification
issue, falsification of fuel assembly inspection results, improper construction work at a spent
fuel reprocessing facility, revealed the importance of transparency of licensee’s QA activities
and of regulatory supervision on them. In view of the circumstances, NISA accelerated study
on regulatory requirements concerning QA and the NSC reviewed the regulation concerning
QA operational safety activities and presented its views and opinions to NISA
Thus, in Japan, NISA established regulatory requirements for QA. NISA encouraged
establishment of academic society and association standards in conformity to international
standards and reviewed their technical adequacy and licensees apply newly establishedacademic society and association standards to their QA activities. NISA, when an inadequate
quality assurance system is discovered at the Operational Safety Inspection to licensees, etc.,
directs the integrated check.
Moreover, licensees and relevant parties are performing the self-controlled operational safety
activities based on the academic society and association standards.
13.1 Legislative Framework for QA of Nuclear Installation
NISA conducts regulatory activities from the design stage to the operation stage, such as alicensing for establishment, an approval of construction plan, and an equipment inspection, on
the basis of the Reactor Regulation Law and the Electricity Utility Industry Law.
In the in-service operational safety activities, the legislative requirements stipulate that thequality assurance system should be established and be included in the Operational Safety
Program, based on the Reactor Regulation Law, and NICS supposes to check the state of

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implementation for the licensee through the Operational Safety Inspection.
The aim of the mechanism mentioned above is for the licensee to establish comprehensive andsystematic QA programs, to be enabled to have firm belief in their operational safety activities
and to implement them and to carry out their accountability to the public to obtain confidenceof the public.
The key points of QA activities are; a) to involve top management, b) to be based oninternational standards on QA (ISO9001: 2000), c) to improve the activities continuously by
Plan-Do-Check-Act cycle for planning, performing, and evaluating the operational safety
activities, and d) to implement the full audit by the independent audit organization.
The Reactor Regulation Law stipulates that licensee's QA program should include 1)
organization governing the performance of QA, 2) plan of activities for operational safety
activities. 3) implementation of activities for operational safety activities., 4) evaluation ofactivities for operational safety activities,. and 5) improvement of operational safety
activities.
Note) The term, activities for operational safety activities., means activities necessary tomaintain safety, in maintenance work of facilities, operation of reactors, and transportation,
storage and management of nuclear fuel materials or materials contaminated by nuclear fuelmaterials.
Licensee prepare their QA program of the nuclear facilities and implement them for theoperational safety activities, according to JEAC 4111-2003, "Rules of Quality Assurance for
Safety of Nuclear Power Plants" (hereinafter referred to as JEAC 4111-2003) established by
the Japan Electric Association (JEA) in autumn of 2003 based on the ISO9001: 2000, and
NISA evaluated the rules and accepted them as the standards to meet the regulatory
requirements.
The contents of JEAC4111-2003 are shown in Table 13-1.
13.2 Verification of Quality Assurance by Nuclear and Industrial Safety Agency
NISA requires applicant for license at the each stage of safety regulation to submit appropriate
QA program, and verifies implementation of QA program as follows:(1) Review of Basic Policy for QA activities at Reactor Establishment StageNISA requires the applicant to submit the "Policy for Quality Assurance" attached to an
establishment license application document in the licensing for establishment of nuclear
installation.
(2) Confirmation of QA Program in Construction StageIn application of the construction plan of nuclear installation, NISA requires the licensee tosubmit the "Description on QA Assurance" which the licensee of the reactor establishment

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should implement through each stage, such as the design, manufacturing, installation andfunctional test. NISA confirms that the licensee has prepared appropriate procedures to audit
principal contractor's procedures such as the procurement quality control, the material control
as well as the audit of the principal contractor's QA program and their process control, also on
the licensee's responsibility.
(3) Verification of QA of Fuel Assembly
NISA requests the licensee of fuel assembly fabrication to submit the application form of theapproval of fuel assembly design which describes such as the performance, the strength andthe flow sheet of fabrication process for fuel assemblies, and "Description on Quality
Assurance" attached at the time. When conducting the Fuel Assembly Inspection, theinspector of NISA verifies not only the licensee's test results but also the adequacy oflicensee's test procedures by checking the extracted test processes without prior notice.
NISA requests the licensee to submit such as the description of QA program for theapplication for inspection of imported fuel assemblies. In addition, the fuel assembly
inspection is described in Section 14.2 in detail.
(4) Verification of QA Activities throughout Operating LifeNISA verifies the licensee's QA activities throughout in-service for the nuclear installation as
follows:
・ Description of provisions relating to the quality assurance in the Operational Safety
Program・ Verification of the Operational Safety Program through the Nuclear Safety Inspection
During outage of nuclear facilities, it is assessed that the periodic licensee's check isimplemented appropriately as Audit of Licensee's Periodic Check System.
NISA is performing inspection on licensee's quality assurance since the 2004 fiscal year
through the Operational Safety Inspection based on the Reactor Regulation Law. Specifically,
NISA confirms licensee's situation of implementing the quality management system in a timely
manner. In addition, in conducting the Operational Safety Inspection, nationwide inspectors
from Nuclear Safety Inspectors Offices meet together at “Nuclear Inspectors Meeting”, andparticipate in the effort aiming at equalization of nuclear-safety-inspection methods andinformation sharing by Nuclear Safety Inspectors, by performing "Operational Safety
Inspection model". The national government and JEA are improving the guidelines in order toadequately conduct root cause analysis focusing organizational factors. The details areprovided in Section 10.2.
In these Operational Safety Inspections and Audit of Licensee's Periodic Check Systems etc.,
following examples, which were caused because of inadequate preparation or functioning of
the licensee's quality management system, were found out, and NISA instructed to implement
the check for fitness of the quality assurance system and to take required measures to thelicensees concerned as shown in the following table;

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Licensee

Date

Background and instruction

The Kansai
Electric Power
Co., Inc.

September 27,
2004

Background
Inadequate preparation of the quality assurance system forensuring systematic "nuclear safety" (Direct causes of thesecondary system piping failure accident at the Mihama
Power Station Unit 3)

Tohoku ElectricPower Co., Inc.

July 7, 2006

Instruction;
・ Verification of the quality assurance system
・ Establishment of effective measures to prevent recurrence
Background・ Inadequate piping thickness control (prevention ofrecurrence of the secondary system piping failure
accident of the Mihama Power Station Unit 3)
・ Inadequate "nonconformity management" and"procurement control"
Instruction;
・ Integrated check of the quality assurance system
・ Establishment of effectual measures to prevent recurrence

13.3 Implementation and Evaluation of QA Program by Licensee
Outlines of QA activities of licensee as follows;(1) Establishment of QA Program
Licensee prepares QA program in accordance with JEAC4111-2003, and implement QA
activities based on it. These programs cover the procedures of document control, design
control, procurement control, management of inspection and testing, nonconformity
management, and audit, etc. The licensee submits "basic Policy for QA" and"Description on QA" to NISA based on this quality assurance program as described in
section 13.2 (1) through (4).

(2) QA activities at each stage of Design, Construction, Commissioning, and Operation(Maintenance)
QA activities are carried out by many organizations. The licensee clarifies the scopeand responsibility of the manufacturer in QA activities, and entrust it with QA
activities in its scope. In the same way, the manufacturer entrust its subcontractors
with QA activities in their scope.

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(3) QA audit, observation and measurement
As activities relevant to an evaluation of the licensee's quality management system, the top
management reviews the quality management system of the company including nuclear
installation periodically.
Moreover, as the observation and measurement (equivalent to Check of thePlan-Do-Check-Act cycle), the internal audit, observation of processes and measurement, andnuclear safety inspection and test are conducted.
As the internal audit, the independent audit that directly reports to the management is
implemented by sections other than the nuclear power division.
And for procurement, following activities are conducted for suppliers;(a) In selecting suppliers, their supply capabilities of procuring articles are evaluated.
(b) At the confirmation of procuring articles, in order to ensure that the procuring articles
conform to procurement requirements, required inspections and other activities are
conducted.
Moreover, the supplier audits subcontractors’ activities in addition to its own internal audit.

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Table 13-1 Contents of JEAC4111-2003,"Rules of Quality Assurance for Safety of Nuclear Power Plants"

0. Introduction
1. Objective
2. Scope of Application
3. Definition
4. Quality Management System
5. Responsibility of Management
6. Management Control of Resources
7. Planning and Implementation of Job
8. Evaluation and Improvement

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Article 14 Assessment and Verification of Safety

Each Contracting Party shall take the appropriate steps to ensure that:(i)Comprehensive and systematic safety assessments are carried out before theconstruction and commissioning of a nuclear installation and throughout its life.
Such assessments shall be well documented, subsequently updated in the light ofoperating experience and significant new safety information, and reviewed under
the authority of the regulatory body;(ii) Verification by analysis, surveillance, testing and inspection is carried out toensure that the physical state and the operation of a nuclear installation continue
to be in accordance with its design, applicable national safety requirements, andoperational limits and conditions.

The national government (governmental offices responsible for safety regulations) andlicensee perform and record the assessment and verification of the safety of nuclear
installations in accordance with the legislative framework provided in Article 7, at each stageof planning, establishment, construction and operation of nuclear installations. The national
government enacts and utilizes necessary regulatory guides for assessment and verification ofeach stage. When the regulatory framework is newly enacted or is updated, licensees carryout necessary assessments for the safety of nuclear installations and receive review by thenational government.
In the safety assessment at the stage before construction, the Nuclear Safety Commission(hereinafter referred to as the NSC) deliberates on the safety review and assessment resultsby the Nuclear and Industrial Safety Agency (hereinafter referred to as NISA) from the view
point of the licensee’s technical capability and non-hindrance in the prevention ofradiological hazards (implementation of double checking).
NISA confirms with the Periodic Inspection, the Operational Safety Inspection, the Audit ofLicensee's Periodic Check System, and the periodic safety assessment (the periodic safety
review and measures for aging management), that licensee’s activities for operational safety
are continuously performed appropriately to satisfy the safety design requirements, limiting
conditions of operation for the facility. In addition, when guidelines and/or technicalstandards etc. are revised reflecting new knowledge etc., NISA directs the licensees to confirm
the safety of the operating facilities, as necessary. The NSC supervises and audits theappropriateness of NISA’s regulatory administration in the construction and operation stages
after issuance of the license, from the view points of rationality, effectiveness andtransparency.
In recent years, the probabilistic safety assessment methodology has been developed as a
useful tool in supporting the conventional deterministic safety evaluation and the resultant
risk information are used in regulatory activities such as development of accident

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management and the Periodic Safety Review (PSR) in Japan. Also introduction of safety goals
and performance goals (proposal) that define the acceptable risk levels are underdevelopment.
New reporting items since the last report are as follows:
NISA developed the "Standard Review Procedures for Aging Management of Commercial
Power Reactors" in order to review the technical evaluation report of aging and the long-term
maintenance program submitted by licensees in accordance with the regulatory requirements.
NISA published the “Basic guidelines for use or risk -information in safety regulation of
nuclear installations (trial use)”, and “The quality guidelines for probabilistic safetyassessment (PSA) of nuclear installations (trial use)”, and is actively making efforts forutilizing such "risk information" in regulation.
In response to the revision of the Regulatory Guide for Reviewing Seismic Design of Nuclear
Power Reactor Facilities by the NSC in September 2006, NISA required licensees toimplement the seismic safety re-evaluation for the existing nuclear installations inaccordance with the revised Guide.
14.1 Safety Assessment prior to Construction
A person who intends to install a nuclear installation submits license application documents
including the results assessed for the safety of basic design of the nuclear installation to NISA
pursuant to the Reactor Regulation Law.
NISA examines whether the application conforms to the licensing criteria prescribed in theReactor Regulation Law.
The Minister of Economy, Trade and Industry consults with the NSC in order to hear opinions
about the results of examination. The NSC deliberates on NISA’s examination results from theview point of the licensee’s technical capability and non-hindrance in the prevention ofradiological hazards, with taking into consideration the opinions received at the public
hearing.
The outlines of the safety assessment for Establishment License submitted and regulatory
criteria are provided below, with a commercial power reactor (light water nuclear powerreactor facility) used as an example.
(1) Documents for Establishment License Application
An application for establishing a commercial power reactor consists of a main text andattached documents detailing safety design, safety analysis, siting assessment, etc. inaccordance with the provisions of the Reactor Regulation Law and the related legislation. The
application document describes basic design with sufficient information for examining thesafety.

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(2) Method and Criteria of Assessment
1) Siting Assessment
The siting assessment of nuclear installations is conducted pursuant to the Regulatory
Guide for Reviewing Nuclear Reactor Siting Evaluation and application Criteria(hereinafter referred to as the “Guide” in this paragraph). The Guide requires thatdesign, construction, operation and maintenance shall be implemented so that an
accident may not occur irrespective of the place in which a nuclear reactor is established.
In addition, in order to ensure the public safety in case of an accident, the following
siting conditions are necessary in principle; a) there has been no event (natural disaster)in the past to induce a large accident and no such event is expected to occur in the future,and there are few events that escalates a disaster, b) nuclear reactors shall be, in
relation to its engineered safety features, located at a sufficient distance from the public,
and c) the environment of the nuclear reactor site including its immediate proximity shallbe such that appropriate measures for the public can be implemented as required.
The conditions for judging the suitability for siting conditions are provided in "theguides for siting examination" of the Guide.
2) Assessment of Safety Design
The basic design and/or design concept of nuclear installations are confirmed that itconforms to the Regulatory Guide for Reviewing Safety Design of Light Water Nuclear
Power Reactor Facilities (hereinafter referred to as the “Regulatory Guide for Safety
Design”) at the safety review and assessment conducted by NISA, and the safety is
discussed and evaluated as a whole in accordance with the provisions of the Regulatory
Guide for Safety Assessment. The evaluation method and the criteria using theRegulatory Guide for Safety Design and the Regulatory Guide for Safety Assessment areprovided in Section 18.1 through Section 18.4.
14.2 Assessment and Verification of Safety Prior to the Commissioning
The licensee shall develop a construction plan for the installation of electric structures,and shall obtain the approval of NISA before starting construction in accordance with theElectricity Utilities Industry Law. After obtaining the approval of construction plan, thelicensee shall undergo the pre-service inspection by NISA at every stage and completion of
construction. For fuel assemblies to be loaded into the nuclear reactor, the licensee shallobtain the design approval and undergo the fuel assembly inspection by NISA. For thewelding of pressurized parts, containments, etc., the licensee shall conduct the Licensee’s
Welding Check, and shall undergo review of the implementation system of the Licensee’s
Welding Check (Audit of Licensee's Welding Check System) performed by theincorporated administrative agency, Japan Nuclear Energy Safety Organization(hereinafter referred to as “JNES”). Moreover, before starting an operation of reactors, inorder to operate the nuclear installation safely, the licensee must obtain the approval ofOperational Safety Program that describes the activity, commitments etc. (refer to Section

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19.3 of Article 19 for the contents).
Approval of construction plan and safety verification at the pre-service inspection,
approval of fuel design and safety verification by the fuel assembly inspection and safety
verification by the Audit of Licensee's Welding Check System are described in thefollowing.
(1) Approval of Construction Plan and Safety Assessment and Verification at the
Pre-Service Inspection
The licensee shall develop a construction plan for installation of electric structures and shallobtain the approval by NISA after obtaining the establishment license and before starting
construction in accordance with the Electricity Utilities Industry Law. NISA reviews it toconfirm that the detailed design of electric structures is not contradictory in the basic design
or fundamental design policies at the establishment licensing stage, and that it is conforming
to the technical standards in accordance with the Electricity Utilities Industry Law for theapproval of construction plan concerned.
Licensees, after obtaining the approval of construction plan, undergo pre-service inspection
by NISA at each construction stage and at the completion of all construction work, to verify
that the construction is completed in accordance with the approval of the construction planand that it is conforming to the technical standards. The pre-service inspection includes
inspections on structure, strength and leak-tightness of each component and inspections onfunction and performance of the overall system of a nuclear installation. Details are shown in
Table 14-1. The inspection at the time of criticality and completion of construction work in
the table are so-called commissioning tests. Since October 2003, JNES conducts part of theabove mentioned pre-service inspection.
(2) Approval of Fuel Design and Safety Verification by the Fuel Assembly Inspection
A person who intends to use fuel assemblies undergoes the Fuel Assembly Inspection,
pursuant to the Electricity Utilities Industry Law, after obtaining the approval of fuelassembly design. NISA, in issuing the approval, verifies that the proposed fuel design takes
into consideration the thermal characteristic, performance in radiation and corrosion
resistance corresponding to operating conditions, and that it maintains sufficient strength
through the years in service. NISA confirms this at inspections for each stage of fabricationthat fuel assemblies are fabricated in accordance with the approved design and technicalstandards. The fuel assembly inspection is also required for reload fuel, regardless of whether
or not there have been design changes. Since October 2003, JNES conducts part of the above
mentioned fuel assembly inspection.
Imported fuel assemblies are also required to undergo and pass the fuel assembly inspection
by NISA.
(3) Audit of Licensee's Welding Check System

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Licensees perform Licensee’s Welding Check on welds of pressurized parts and containment,
and the management system of Licensee’s Welding Check which undergoes review by JNES.
In addition, the Working Group for Welding Control and Inspection of the Electric Power
Safety Subcommittee under the Nuclear and Industrial Safety Subcommittee prepared a report"Application Improvement in an Audit of Licensee's Welding Check System" concerning
appropriate verification procedures of the Licensee’s Welding Check system, welding
specialists and welding method in November 2006.
14.3 Assessment and Verification of Safety during Operating Life TimeThe licensee performs a comprehensive safety verification of nuclear installations by theperiodic safety assessment, Periodic Licensee's Check, surveillance pursuant to the
Operational Safety Program, evaluation and investigation at the time of an accident or failure
and measures to prevent recurrence, and undergoes the spot entry inspection at any time inaddition to the Operational Safety Inspection, Periodic Inspection, and the Audit of Licensee's
Periodic Check System during the operating lifetime.
Safety verification by inspection and periodic safety assessment are described in thefollowing:(1) Verification of the Safety by Inspection
NISA performs the Operational Safety Inspection for verification to confirm theappropriateness for activities of the licensee to ensure operation safety and the Periodic
Inspection to confirm activities to ensure the integrity of equipment of nuclear installations.
Furthermore NISA performs the comprehensive assessment of the system, method etc. of the
Periodic Licensee's Check in response to the notification of the review results of the Audit ofLicensee's Periodic Check System performed by JNES. These inspections are performed inaccordance with related legislations and regulations as follows.
NISA has set-up resident Nuclear Safety Inspectors at nuclear installations and performs four
inspections per year (Operational Safety Inspection) on the observance of Operational Safety
Program in accordance with the Reactor Regulation Law to determine the status of complianceto various regulations for the safety and the status of the activities for safe operation
performed by the licensee.
In accordance with the Electricity Utilities Industry Law, NISA and JNES perform the
Periodic Inspection of structures and components important to safety within a time intervalthat does not exceed 13 months after the day of commissioning and the final day of the last
Periodic Inspection.
The inspections which used to be conducted by the licensee as a self-controlled inspection was
redefined in October 2003, by the amendment of the Electricity Utilities Industry Law as a
Licensee’s Periodic Check, and JNES performs the audit type inspection (Audit of Licensee's
Periodic Check System) to review the implementing system, planning and management of

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Licensee’s Periodic Check.
After two years of experience, the firm establishment of these inspection systems has beenpromoted considerably. However, an increase is expected in the number of nuclear
installations that have been operating for many years, and it is necessary to improve measures
for aging management further, the Task Force on Inspection System had resumed in November
2005. Section 19.3 provides the status of the Task Force activities.
(2) Periodic Safety Assessment
Pursuant to the request of the Ministry of International Trade and Industry (at present,
Ministry of Economy, Trade and Industry), licensees had performed the periodic safety review
voluntarily since 1992. However NISA has decided that it is necessary to define the position of
the periodic safety review as a part of reconsideration of nuclear safety regulations.
In accordance with the Reactor Regulation Law, NISA decided in October 2003 to obligate
licensees to implement the "Periodic Safety Review (PSR)"after every 10 year interval since
the first review at the time not exceeding ten years after commissioning and the "Aging
Management Review" within 30 year limit after commissioning.
1) Periodic Safety Review (PSR)
PSR is a licensee's effort to evaluate his operational safety activities performed since thecommissioning of the nuclear installation about every ten years and to obtain the prospectof capability for the nuclear installation to continue safe operation thereafter withkeeping a higher level than or equivalent to the newest nuclear installation.
In December 2005, NISA revised the Rules for the Installation, Operation, etc. ofCommercial Power Reactors under the Reactor Regulation Law. And the details of thelicensee’s activities that should be performed at the periodic safety review were clarified
in "Evaluation on the Status of Implementing the Operational Safety Activities at Reactor
Facilities" and "Evaluation on the Reflection Situation of the Latest Technical Knowledge
for Operational Safety Activities Performed at the Reactor Facilities".
In order to understand the licensee’s status on the improvement in aging management andon the prevention of deterioration in the organization culture, both in short term and inlong term, NISA decided to verify the licensee’s maintenance management activities andthe status of the organization culture at a usual Operational Safety Inspection (for shortterm understanding) from January 2006. Furthermore NISA also decided to verify thatthe licensee understand the status appropriately and is taking measures, such as
intensification of surveillance, at the periodic safety review (for long term understanding)
implemented every ten years after the start date of the operation of reactor.
2) Aging Management Review
In October 2003, NISA added the Aging Management Review in the provisions of

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"Periodic Assessment of Reactor Facilities" to the "Rules for the Installation, Operation,etc. of Commercial Power Reactors" and provided it as one of the requirements in the
Operational Safety Program to implement Measures for Aging Management.
Matters to be implemented as the Aging Management Review are: a) to analyze theimpacts of technically conceivable aging phenomena on components and structures ofnuclear power stations with safety functions at a time within 30 years aftercommissioning, and to technically evaluate the possibility for prevention of the loss offunction of the components and structures due to aging phenomena under the present
maintenance activities provided to them, b) to extract new maintenance measures fromthe technical evaluation results to develop the ten-year maintenance program. c) to
re-evaluate the ten-year maintenance program with a ten-year interval. And NISA
decided to confirm the status of the implementation of these items which shall be
implemented as a part of the quality assurance systems of the Operational Safety
Program.
The Aging Countermeasure Examination Committee under the Nuclear and Industrial
Safety Subcommittee, issued the "Enhancement of Measures for Aging Management at
Commercial Nuclear Installations" in August 2005. In response to the report, NISA
prepared the “Guidelines in Implementing Measures for Aging Management at
Commercial Nuclear Installations” and the “Standard Review Procedures on Measures
for Aging Management of Commercial Power Reactors” in December 2005. And, JNES
issued the "The compilation of Technical Information concerning Measures for Aging
Management", which disclosed the standards, view points and evaluation points when
NISA and JNES assess and review licensee's technical evaluation reports and long-term
maintenance programs.
The aging technical evaluation reports for Unit 3 of the Fukushima Daiichi Nuclear
Power Station, Tokyo Electric Power Co., Inc., Unit 1 of the Hamaoka Nuclear Power
Station, Chubu Electric Power Co., Inc., and Unit 3 of the Mihama Power Station, theKansai Electric Power Co., Inc. are submitted at present.
NISA assessed the technical evaluation report and the long-term maintenance programfor Unit 1 of the Hamaoka Nuclear Power Station, Chubu Electric Power Co., Inc. on
January 31 2006, and the result was reported to the NSC in May 2006 as follows.
Summaries of the assessment results; 1) the system for implementing the technicalevaluation is appropriate, 2) the technical evaluation implemented for aging, thetechnical evaluation for ensuring seismic safety, and conservation measures areappropriate, and 3) the long-term maintenance program based on the technicalevaluation is appropriate.
The same assessments were carried out for the other two nuclear reactors mentioned
above. NISA will study issues for improvement extracted from the actual evaluation
results of the three nuclear installations so that effective improvement is achieved.

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Also, in response to the "Completion of Measures for Aging Management at Commercial
Nuclear Installations" issued in August 2005, licensees reported to the Aging
Countermeasure Examination Committee, the status of efforts for the measures for aging
management for the following matters; (1) ensuring transparency and effectiveness, (2)preparing technical information infrastructure, (3) preventing deterioration in corporate
culture and organization culture, and keeping and improving technical capabilities, and(4) steadily achieving the accountability on measures for aging management. From now
on, the licensees will carry out confirmation of the situation of measures for aging
management at the Periodic Licensee's Check, and NISA decided to verify the licensees’
implementing situation through the Periodic Inspection, Audit of Licensee's Periodic
Check System, Operational Safety Inspection, etc.
NISA launched the Coordination Committee on Technical Information in JNES, in orderto reinforce the measures for increasing aged plants, and in order to share domestic andoverseas technical information for effective utilization among the industrial world,
academic and governmental institutions.
Furthermore, an ad-hoc committee that consists of NISA, JNES, universities, researchorganizations, electric utilities, nuclear plant manufacturers, plant engineering
companies, etc. was established under the Atomic Energy Society of Japan. From July
2004 through March 2005, the ad-hoc committee had prepared a load map on measures
for aging management and long life-time safe operation of light water reactors.
In addition, NISA regards the aging of nuclear installations as one of the more importantresearch subjects on safety. Clarification of the aging phenomenon and prediction of theaging process, development of early detection and detailed measurement methods ofcracks and deteriorations, and development of the structural integrity evaluation method
are the subjects of concern. The research on irradiation assisted stress corrosion
cracking (IASCC) using the material testing reactor (JMTR) of the Japan Atomic Energy
Agency, the probabilistic fracture mechanics research, and JNES’s data-base andtechnology development for flaw detection and sizing, etc. are currently underway.
(3) Assessment at the Occurrence of an Accident or a Failure, Survey on Accidents andFailures and Measures to Prevent Recurrence
Activities of licensees and regulatory bodies for assessment at the occurrence of an accident
or a failure, survey on accidents and failures and measures to prevent recurrence are provided
in Section 19.6 and 19.7. Furthermore, Section 6.2 provides actions taken for accidents andfailures at existing nuclear installations.
14.4 Utilization of Probabilistic Safety Assessment in Regulation(1) Utilization of Probabilistic Assessment in the Actual Regulation
1) Judgment on the Necessity of the Protection for External Events

14-8

In the safety review and assessment for an application for reactor establishment
license, the protection is required for external events (including natural events andhuman induced events which may occur inside or outside a facility), when thepossibility of occurrence of events that affect safety related facilities and equipment
exceeding a certain definite value, and the adequacy of the protection design is asubject at the safety review and assessment..
For example, at a turbine missile event affecting the reactor facility, the probabilityof damage to the reactor facility (coolant pressure boundaries of the containment andthe spent fuel pool) has thus been so far evaluated. If the result exceeds the criterion,
it is required to consider it in the design as “a missile assumable to generate inside
the reactor facility.”
As for an aircraft drop to the reactor facility, the probability of an aircraft drop to
the reactor facility has been evaluated. And if the result exceeds the criterion, it isrequired to consider it in the design as "a human induced external event." In addition,
NISA enacted the "Criterion for Evaluation of the Aircraft Drop Probability to a
Commercial Nuclear Installation (by-law)" in July 2002.
2) Evaluation on Effectiveness of Accident Management Measures
The licensees have implemented the PSA for nuclear power reactors under operation orconstruction to evaluate the soundness of the core and the containment in the case of asevere accident, and utilized the result for effectiveness assessment of their accident
management (AM) measures. Internal events during operation were subject to theiranalysis, and the results were utilized to discover AM measures and to evaluate theireffectiveness.
3) Utilization of the Probabilistic Safety Assessment (PSA) at the Periodic Safety Review
As provided in 14.3 (2), licensees have implemented the PSA for internal events during
power operation using the latest data for the periodic safety review for commercial
nuclear installations since 1992. And the results were utilized for safety assessments,
understanding the features on the safety of nuclear installations concerned, andverification for the effectiveness of accident management measures etc. Furthermore, atthe periodic safety review after 2001, the PSA for internal events during the shutdown
condition has been additionally implemented. The position of the periodic safety review
in legislations was identified in October 2003, but implementation of the PSA was not
obliged by the law. However licensees are requested to implement the PSA as part oftheir independent activities since then.
4) Assessment of Impacts and Measures of Operating Experience Etc.
In the study on the measures to prevent recurrence of the pipe rupture accident which
occurred in November 2001 on the steam condensing line (SCL) of the residual heatremoval system (RHRS) of the Unit 1 of Hamaoka Nuclear Power Station, the PSA, with

14-9

considerations of the pipe break accident specifically to the SCL of the RHRS, wasimplemented on the core integrity, and it resulted that the pipe rupture of the accident
concerned does not significantly increase the risk. It was also evaluated that several
proposed measures were effective in reducing the risk.
In addition, the effectiveness of provisional measures for functional impediment of theemergency core cooling system strainer of BWRs and the screen ofcontainment-recirculation sump of PWRs was confirmed by the PSA.
(2) Activities for Introduction of the Safety Regulation Utilizing Risk Information
In order to establish a more effective and efficient regulatory system, many countries aremaking efforts to utilize risk information for safety regulations, though the extent of such
efforts is different.
Also in Japan, the NSC decided the “Basic Policies on Introduction of Nuclear Safety
Regulation Utilizing Risk Information” in November 2003.
NISA together with JNES, receiving the decision, published the "Fundamental Concepts of
Utilizing the "Risk Information" at Nuclear Safety Regulations" in May 2005, and provided
the way of thinking for utilizing the "risk information" at nuclear safety regulations. In
addition, the "Present Implementation Plan of Utilizing the "Risk Information" at Nuclear
Safety Regulations" was published, and the study on utilizing the "risk information" waspromoted in accordance with the implementation plan. The implementation plan was revised
based on the progress since then in January 2007.
NISA published the "Guideline for Utilizing the "Risk Information" at the Safety Regulationsof Nuclear Power Stations" (trial use)", and provided the guidelines for utilizing "riskinformation" at the safety regulations of nuclear power stations in April 2006. Furthermore asa guideline for ensuring quality assurance of the PSA in utilizing the "risk information", the"Guidelines for the Quality of the PSA for Nuclear Power Stations (trial use)" was published.
In addition, the NSC established “Task Force for Introduction of Safety Regulations Using
Risk Information (RIR)”, in this Task Force, the current status of relevant organizations’approaches to the application of RIR and the issues to be solved for the wider utilization ofrisk information were compiled.
(3) Introduction of Safety Goals and Performance Goals
The NSC issued the "Interim Report on the Investigation and Review on Safety Goals" inDecember 2003, and the performance goals were established for nuclear installations in March
2006. The outlines of the goals are as follows.
1) Safety goals
The safety goal should be established for all activities in the utilization of nuclear

14-10

energy that may have an adverse influence of radiation exposure on the public.
The objectives to establish the safety goals are as follows;• To make it possible to assess regulatory activities in utilizing nuclear energy at
various fields with same standards for reasonable and consistent evaluation amongthem,• To make it possible to exchange opinions on the way of nuclear regulatory activities
of national governments, such as establishment of guidelines and standards, among
the national government and people more effectively and efficiently, and• Make it possible for licensees to implement their independent risk management
activities more effectively and efficiently to meet the expectation of the regulatory
authority.
Along with these objectives, first of all, the safety goals are applied as reference to makejudgment on the whole regulatory activity, in terms of rationality and consistency, and it
is considered as appropriate to start with more general applications, on a specificfacility, after abundant experience for the safety goals are accumulated.
The safety goal is of two fold. One is the qualitative goal, which is a controllable level ofrisk due to an accident that licensees must observe under the nuclear safety regulations.
The other is the quantitative goal that specifies the numerical value corresponding to theacceptable level of the risk. In this context, the risk during the normal operation of
nuclear power reactor facilities is excluded. And as the indices for quantitative goals,
the death risk of the average individual of the public who lives in a certain range isused.
The proposal on safety goals are made of the following configurations.
a. Proposal to Qualitative Goal
The possibility of health damage to the public by emission of radiation or release ofradioactive materials accompanied with activities for utilization of nuclear energy
should not meaningfully increase the risk of damage to the public’s health in daily life.
b. Proposal to Quantitative Goal
The mean value of acute fatality risk by radiation exposure resultant from an accidentof a nuclear installation to individuals of the public, who live in the vicinity of the siteboundary of the nuclear installation, should not exceed the probability of about 1x10 -6
per year. And, the mean value of fatality risk by cancer caused by radiation exposure
resulting from an accident of a nuclear installation of individuals of the public, wholive in the area but some distance from the nuclear installation, should not exceed theprobability of approximately 1x10 -6 per year.

14-112) Performance Goal
It is reasonable to review and indicate the level that will be understood as theperformance goal to conform with the safety goal, according to the characteristics ofeach accident that could occur at nuclear installations.
The preparation of the PSA method is advancing and the following performance goals
are proposed for nuclear power generation facilities that have experience in utilizing
the risk information.
a. Indices for the Performance Goal
The following indices are also used, because they well represent the facility
performance on the integrity of a reactor core and the integrity of the confinement
function of a containment unit, and are clearly defined and appropriately quantified.
Index 1: Frequency of core damage (CDF)
Index 2: Containment loss-of-function frequency (CFF)
b. Proposal to Indices Value
The knowledge obtained by the PSA for domestic nuclear installations implemented
by the national government, research organizations, licensees etc. and the PSA
results in the U.S. etc. were studied as references, and the following indices values
were proposed as a proposal to the performance goals corresponding to the proposal
to the safety goals.
Index value 1: CDF: 10 -4 / reactor-year approximately.
Index value 2: CFF: 10 -5 / reactor-year approximately.
These indices values shall be satisfied concurrently.
From now, studies for preparing a framework for use of performance goals in safety
regulations, application to nuclear installations other than commercial nuclear
installations, and a high safety level in future reactors are required.
14.5 Assessment of Seismic Safety(1) Assessment of the Seismic Safety of Existing Nuclear Installations Following the Revisionof the Regulatory Guide for Reviewing Seismic Design of Nuclear Power Reactor Facilities
The NSC revised the Regulatory Guide for Reviewing Seismic Design of Nuclear Power
Reactor Facilities, etc. in September 2006 (refer to the report for Article 18).
Following the revision, NISA required licensees to assess the seismic safety for existing

14-12

nuclear installations etc. in accordance with the revised Regulatory Guide for Reviewing
Seismic Design of Nuclear Power Reactor Facilities, and to report the results.
Each licensee etc. submitted his implementation plan for the seismic safety evaluation to NISA
in October 2006.
Each licensee submitted the process to implement the safety assessment within two to three
years, and the implementation plan describes that the assessment will be implemented in theorder of geological and active faults investigation, determination of the design basis
earthquake ground motion, and evaluation for seismic safety.
NISA decided to verify the adequacy of the contents of the licensees’ reports for the seismic
safety assessment, and then to report the verified results to the Aseismatic and Structural
Design Subcommittee of the Nuclear and Industrial Safety Subcommittee of the Advisory
Committee for Natural Resources and Energy.
As of the end of June 2007, the assessment results for Hamaoka Unit 3 and 4 were submitted
from the licensee etc. (February 21 and January 25, 2007, respectively). NISA will strictly
verify the adequacy of the reported results by means of the study at the Aseismatic andStructural Design Subcommittee of the Nuclear and Industrial Safety Subcommittee, etc.
The NSC established “ Investigation Project Team on Seismic Safety of Nuclear Facilities” inJuly 2007. The mandate of this Project Team is to review the NISA’s review results on the
Seismic Safety Re-evaluation of the existing nuclear facilities that will be done in near future.
(2) Assessment of Seismic Safety for the Nuclear Installation Experienced a Major Earthquake
1) Assessment of Seismic Safety for the Onagawa Nuclear Power Station at the Earthquake
Occurred at Miyagiken-oki in August 2005
NISA directed Tohoku Electric Power Co., Inc. to analyze the factors of the earthquake
ground motion that exceeded the design basis earthquake ground motion at the Onagawa
Nuclear Power Station that was confirmed at the earthquake of Miyagiken-oki on August
16 2005, and to assess the seismic safety of the equipment important to safety. The
seismic safety assessments (reports) for Onagawa Unit 2 and 3 were submitted on
November 25, 2005 and January 20 2006, respectively. NISA reported these seismic
safety assessments (reports) to the experts of the Aseismatic and Structural Design
Subcommittee of the Nuclear and Industrial Safety Subcommittee. And NISA asked JNES
to make a crosscheck assessment and study the seismic safety. NISA determined that thelicensee's study results on the seismic safety was appropriate, and informed theconclusion to Tohoku Electric Power Co., Inc.
Furthermore, the seismic safety assessment (report) for the Onagawa Unit 1 was
submitted by Tohoku Electric Power Co., Inc. on May 19, 2006. The additional report forthe seismic safety for the reactor building foundation ground based on an investigation
result of additional boring was submitted on June 12. The amendment report including a

14-13

piping wall-thinning survey implemented after the report for Onagawa Unit 1 and theanalytical evaluation results were submitted on August 22.
NISA held the Aseismatic and Structural Design Subcommittee of the Nuclear andIndustrial Safety Subcommittee, and received the expert's opinions on the reports, and
NISA asked JNES to make a crosscheck assessment and study the seismic safety for Unit
1. NISA determined that the licensee's study results for seismic safety was appropriate,
and reported the conclusion to Tohoku Electric Power Co., Inc. on September 13, 2006.
2) Assessment of the Seismic Safety for the Shika Nuclear Power Station at Noto-Hanto
Earthquake in March 2007
Unit 1 and 2 of the Shiga power station were in shutdown condition at the time of theNoto-Hanto earthquake on March 25, 2007. As the ground motion during the Noto-Hanto
earthquake exceeded some of the design basis earthquake ground motions, the licensee
re-evaluated the seismic safety of Unit 1 and 2, and reported the results to NISA on April
19, and additional report and amendment were also submitted on June 1 and August 20,
2007respectively. NISA decided to confirm the reports that were submitted by the licensee,
and NISA held the Aseismatic and Structural Design Subcommittee of the Nuclear andIndustrial Safety Subcommittee, and received the expert's opinions on the reports. NISA
determined that the licensee's study results for seismic safety was appropriate, andreported the conclusion to Hokuriku Electric Power Co., Inc. on August 27, 2007
receiving experts' opinions at the Aseismatic and Structural Design Subcommittee of the
Nuclear and Industrial Safety Subcommittee of the Advisory Committee for Natural
Resources and Energy.
14.6 Promotion of Nuclear Safety Research(1) Prioritized Nuclear Safety Research Program by the NSC
The NSC proposed nuclear safety research (prioritized nuclear safety researches) that shouldbe performed selectively for about five years from the 2005 fiscal year. In the proposal, thefollowing important research areas were proposed as safety researches for nuclear
installations; 1) the regulatory system area (example: use of risk information, assessment ofroot-causes of failures and accidents), 2) light water reactor area (example: safety analysis,material degradation and aging management, seismic safety technologies) and nuclear
reactor disaster prevention technologies.
In order to reflect the latest technical knowledge obtained from the results of safety researches
upon safety regulations and improvement of safety regulations, regulatory authorities andresearch organizations should come to a mutual understanding, and the regulatory authorities
should make efforts to identify the outcome of safety research required for safety regulations
and to show clearly how the results of safety researches are to be utilized. Furthermore,
research organizations are required to arrange and present the issues and results for safety
research that meet the requirements of regulatory authorities appropriately. Therefore, theNSC understands and evaluates the status of research in related research organizations.

14-14

Furthermore, the NSC holds a debrief session on the results of safety research every year.
From the 2006 fiscal year, in order to support safety research to be implemented smoothly, the
NSC started Nuclear Safety Research Forum. And the relevant persons coming from industries,
academia and the government meet together, and they deliberate on regulatory required safety
research and utilization, the promoting system, and international contributions and theyexchange information.
(2) Safety research by NISA
Since ensuring safety of nuclear installations is one of licensee’s responsibilities, the licensee
must implement all safety research necessary for ensuring safety. On the other hand, theregulatory body is required to promote appropriate safety research in order to make a morescientific judgment.
NISA has performed extensive safety research and has implemented the following activities
during the report period.
1) Review of Plans and Results of Research by the "Fundamental Research Subcommittee
for Ensuring Nuclear Safety"
NISA established the Fundamental Research Subcommittee for Ensuring Nuclear Safety
under the Nuclear and Industrial Safety Subcommittee in September 2006, and decided toplan, implement and assess research for the nuclear safety infrastructure with a load map
to be developed for building a framework systematically and efficiently aiming atensuring safety by industrial societies and regulatory authorities etc. in the fields thatshould be focused on from now on. Then, NISA will plan, implement and assess thefundamental research for ensuring nuclear safety.
2) Continuation of Research Facilities and Preparation for Strategic Research Organization
Recent years, research facilities for nuclear safety are under the threat of closureinternationally. Therefore, NISA reviewed the continuation of research facilities at theFundamental Research Subcommittee for Ensuring Nuclear Safety. The subcommittee
proposed that the Japan Material Testing Reactor (JMTR) of the Japan Atomic Energy
Agency should be positioned as the strategic fundamental safety research facility.
In addition, after the pipe rupture accident at Unit 3 of the Mihama Power Station, it was
decided to prepare a strategic organization for nuclear safety research in Fukui
Prefecture, where the Mihama Power Station is located. NISA defined the safety research
to be performed, and contributed to the strategic research plan.
3) Promotion for International Joint Studies
NISA and JNES have been promoting the international cooperation research actively.
Especially in the reporting period, the cooperation with OECD Nuclear Energy Agency(NEA) was promoted. Specifically, NISA supported the OECD/ROSA plan in the field of

14-15

thermal hydraulic safety research that the Japan Atomic Energy Agency will perform as
the first Japanese host organization.
In the fuel material research, NISA participated to the OECD/CABRI plan and theHalden plan for some time. NISA started the SCAP plan that is the project specially
funded by Japan. In the project, stress corrosion cracking (SCC) and cable aging arebeing studied in the NEA since measures for aging management are very important. In
addition, NISA has been continuously participating in many of the OECD projects
performed abroad and participating in the IAEA’s CRP (Coordinated Research Activities),etc.
(3) Safety Research by JNESJNES, as a specialized agency supporting NISA for technical fundamentals, promotes safety
research necessary for providing the scientific knowledge that should be reflected in the safety
regulations, such as preparation of safety standards and criteria necessary for the safety
regulation of nuclear installations etc.,.
(4)

Safety Research by the Japan Atomic Energy Agency

The Japan Atomic Energy Research Institute and the Japan Nuclear Cycle Development
Institute unified in October 2005, and the Japan Atomic Energy Agency was established as anincorporated administrative agency, which performs comprehensive research anddevelopments for nuclear energy. The Agency possesses a large number of facilities necessary
for implementing the safety research, and has human resources over broad area of expertise.
Therefore, the Agency plays the central role in implementing safety research, and is required
to play the role to support the safety regulation technically utilizing integrated nuclear energy
research facilities.
(5) Safety Research by Licensees etc.
In order to cope with longer life and sophistication in use of light water reactors, licensees
implement research that is needed for improvement in safety, reliability and economical
efficiency. The outcome of the research is reflected in preparation and sophistication of theprivate sectors’ standards as needed, which contributes in ensuring safety.

14-16

Table 14 - 1 Outline of Pre-Service Inspections
Time ofInspection(1) At the timeof installation ofeach structure
and component(2) At the timeof installation of
steam turbine
and auxiliary
boilers(3) At the timeof fuel loading(4) At the timeof criticality(5) At the timeof completion of
construction

Contents of Inspection
Test of structure, strength and/or leak tightness of reactor, reactor coolingsystem, instrumentation and control system, fuel handling system,
radiation management system, waste processing system or reactor
containment structure is performed, when each item is installed and readyto be tested.
Specifically, material inspection, structure inspection, pressurized leak
test, inspection on foundation and support structure are performed
Test of steam turbine structure is performed when installation of bottom
half part of turbine casing is completed.
Test of structure, strength and/or leakage on auxiliary boiler is performed
when its main part is completely assembled.
When the reactor is ready for fuels to be loaded, inspections of systems
around reactor, items required ensuring safety before fuel loading, anditems for which inspection would be difficult after fuel loading are
performed.
In the case of BWR, inspection of main steam bypass valves, inspection offunction and performance of those systems as control rod drive system,core spray system, residual heat removal system, etc. and functional
inspections of safety protection system, etc. are performed.
When the reactor attained criticality, inspections are performed on nuclear
characteristics of reactor core, and overall function and performance of
nuclear installations which can be performed only after fuel loading.
In the case of BWR, an inspection to confirm shutdown margin at full fuelloading, inspections of control rod full stroke test, effective multiplication
factor at the first criticality and moderator temperature coefficient tests
are performed.
When all construction work under the Construction Plan has been
completed, inspections are performed on performance of systems aroundreactor, overall functions and performances of nuclear installations thatcan be confirmed after fuel loading, and functions and performance ofsystems other than those around reactor.
In the case of BWR, inspections are performed on one control rod scram
test, loss of external power-supply test, generator load interception
inspection, plant trip inspections, and load inspections.

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14-18

Article 15 Radiation Protection
Each Contracting Party shall take the appropriate steps to ensure that in all operational
states the radiation exposure to the workers and the public caused by a nuclear installation
shall be kept as low as reasonably achievable and that no individual shall be exposed toradiation doses which exceed prescribed national dose limits.

The standards of radiation protection for the general public and personnel engaged in radiation
work in Japan are prescribed in the laws and legislations, such as the Reactor Regulation Law,
the Electricity Utilities Industry Law and the Industrial Safety and Health Law, etc. The 1990
recommendations of the ICRP are incorporated into their provisions of radiation protection
with due considerations. Consequently, licensees have kept the radiation exposure doses of
personnel engaged in radiation work below the dose limit, as a matter of course, and haveattempted to reduce the exposure based on the ALARA policy.
15.1 Summary of Laws and Requirements on Radiation Protection
The national standards of radiation protection for a nuclear installation are provided in theReactor Regulation Law, the Electricity Utilities Industry Law and the Industrial Safety and
Health Law, etc. and related ordinances, ministerial orders and notifications based on theselaws, and guidelines. The 1990 recommendations of the ICRP are given due consideration andhave been incorporated into legislation and regulation. At present the revision of the 1990
recommendations of ICPR is in progress at ICRP, Japan will incorporate the future revision,
if necessary, based on discussions at IAEA etc.
As the clauses on radiation protection, a ministerial ordinance, ‘The Rules for Commercial
Nuclear Power Reactors concerning the Installation, Operation, etc.’ under the Reactor
Regulation Law is established, which prescribe area control for radiation protection, radiation
control of workers in the radiation controlled areas, measurement and surveillance of radiationlevels, monitoring of discharged radioactive materials, and maintenance of radiation control
equipment. Also the Dose Limits Notification are enacted within the said ministerial order,
which prescribe dose limits and concentration limits of radioactive materials both inside thecontrolled area and outside the peripheral monitoring area, dose limits and concentration
limits of radioactive materials for radiation workers, and dose limits for workers in emergency
activities.
In order to ensure those rules are complied with, each licensee, is required to prescribe in the
Operational Safety Program, 1) radiation control area, access controlled areas, and peripheral
monitoring area and access control to these areas, 2) monitoring equipment at air ventilation
and water discharge, 3) monitoring of the dose, the dose equivalent, the concentration ofradioactive materials and the density of the surface radioactive materials of objects
contaminated by radioactive materials, and the decontamination, 4) maintenance of radiation
monitoring equipment.

15-1

The ministerial order “Ordinance of Establishing Technical Standards for Nuclear Power
Generation Equipment” based on the Electricity Utilities Industry Law, provides technicalstandards for radiation control equipment (biological shielding walls, ventilation facilities,
instrumentation devices, alarm devices, and waste processing equipment, etc.) at nuclear
installations. NISA confirms that such radiation control equipment conforms to the ministerial
order at issuing approval of the construction plan and when they conduct an inspection of theequipment.
The Industrial Safety and Health Law provides that licensees (employers of laborers) takemeasures to prevent damage to the health of radiation workers, including radiation exposure,throughout their period of employment, and it requires that they be educated on issues ofhealth and safety, work environment monitoring and medical examination of workers. On thebasis of the law, the Ministry of Health, Welfare and Labor has enacted a ministerial order,‘the Rules for Prevention of Hazards from Ionizing Radiation’, which prescribes controlledareas, dose limits and measurement, protection from external radiation, and prevention ofradioactive contamination.
Radioisotopes etc. used in nuclear installations are also regulated in accordance with ‘the Law
Concerning Prevention from Radiation Hazards due to Radioisotopes, etc.’ in the samemanner as regulated by the Reactor Regulation Law.
In examining the license to establish a nuclear installation, it is confirmed that the application
conforms to the Regulatory Guides established by the NSC in addition to the legislation andtechnical standards mentioned above. The Regulatory Guide for the Annual Dose Target for
the Public in the Vicinity of Light Water Nuclear Power Reactor Facilities gives dose target
guide to reduce the discharge of radioactive materials from a nuclear installation into
environment and the dose of the public as low as reasonably achievable (ALARA).
Each licensee has defined the release control value of liquid wastes and gaseous wastes in the
Operational Safety Program based on this Regulatory Guide for the Annual Dose Target for
the Public in the Vicinity of LWR.
The 1990 Recommendation of the ICRP(Publication 60) has been, after examination by the
Radiation Council, incorporated into national legislations and regulations on radiation
protection, by revision of related ministerial orders and notifications in April 2001 with thefollowing additional considerations. First, the radiation controlled area is defined where thedose may exceed 1.3 mSv / 3 months, corresponding to 5 mSv/year which is a special dose
limit to the public. Second, the occupational dose limit for female workers is set at 5 mSv / 3
months, an allocated value for a shorter period, reducing possible dose of a potential embryo.
The dose limits in emergency work remain 100 mSv/ year as before, considering the IAEA
BSS.
The Radiation Council is an organization established under MEXT for the purpose ofcoordinating technical standards on prevention of radiation hazards. The Radiation
Council submits reports related to inquiries from related administrative organizations, or

15-2advises them as necessary.

15.2

Laws and Requirements and Response of Licensees(1) Dose Limits
1) Dose Limits for Controlled Areas
The Rules for Commercial Power Reactors and the Dose Limits Notification requires
licensees to establish a radiation controlled area including the reactor room, spent
fuel storage facilities and radioactive waste disposal facilities, where the dose of
external radiation may exceed 1.3 mSv for three months, or where the concentrationof radioactive materials in the air or the surface density of radioactive materials may
exceed the values specified in the Notification, respectively, and to establish
necessary measures to be taken in these areas.
2) Dose Limits for Occupational Exposure
The dose limits for radiation workers are specified in the Dose Limits Notification aslisted in Table 15-1
Table 15-1 Dose limits for radiation workers
Item
1. Effective dose limitsa) Radiation workersb)Femalec)Pregnant female

2. Equivalent dose limitsa) Eye lensb) Skinc) Female abdominal region

3. Dose limits in emergencya) Effective doseb) Equivalent dose for eye lensc) Equivalent dose for skin

Limit
100 mSv / 5 years, but do not exceed 50 mSv
for any year
100 mSv / 5 years, but do not exceed 5 mSv
for any 3 months
100 mSv / 5 years, but do not exceed 5 mSv
for any 3 months and do not exceed 1 mSv
from internal exposure during pregnancy

150 mSv/ year
500 mSv/ year
2 mSv from notification of pregnancy todelivery

100 mSv/ incident
300 mSv/ incident
1 Sv/ incident

15-3

Licensees have paid much attention not only to comply with the dose limits but also
to reduce doses in line with ALARA concept by incorporating the following
activities:- reducing the radiation source in systems and components of a nuclear
installation,- keeping distance from or setting shields against radiation sources,- reducing working time in a radiation environment.
Consequently, the exposure doses of the radiation worker, etc. have beensuccessfully reduced to the level as shown in Annex 2.
Exposure doses of radiation worker in a commercial nuclear installation during thereporting period are summarized below;
a. Individual dose at commercial nuclear installations
The average of individual dose at commercial nuclear installations for the reporting
period was 1.0 to 1.4 mSv / year, and it shows the gradual decrease trend for the
past several years. Thus, the doses are well within the dose limit prescribed in the
notification.
In fiscal year 2006, the average of annual dose of radiation workers was 1.0 mSv
and the maximum annual individual dose experienced per nuclear installation was
19.7 mSv, these numbers were within the dose limit of the notification, which aresimilar to the value of 1.0 mSv and slightly lower than the value of 19.8 mSv for theprevious year of 2005, respectively. No worker who had worked in multiple nuclear
installations exceeded 20 mSv, which is well below the dose limits. The number ofradiation workers who had been exposed of 15 to 20 mSv was 243, and this number
showed a slight increase from 216 of the previous year.
Table Collective Dose, Average Individual Dose and Number of Workers
at Commercial Nuclear Installations
Fiscal year
Collective Dose(man-Sv)
Average annual
individual dose(mSv)
Total number ofworkers

2000

2001

2002

2003

2004

2005

2006

78.83

78.05

84.03

96.41

77.86

66.91

67.43

1.2

1.2

1.3

1.4

1.2

1.0

1.0

65,90067,80063,80066,60066,70066,30066,90015-4

b. Performance of Collective Dose at commercial nuclear installation
In Japan at the end of June, 2007, total of 55 units, namely 32 BWRs and 23 PWRs wereoperating.
The collective doses per reactor year have been slightly decreasing in recent years. The
data for operating BWRs were 1.58 man-Sv in 2004 FY, 1.39 man-Sv in 2005 FY, and 1.33
man-Sv in 2006 FY respectively. The data for operating PWRs were 1.25 man-Sv in 2004
FY, 0.97 man-Sv in 2005 FY and 1.08 man-Sv in 2006 FY respectively. The reasons for thedecrease were mainly due to decreases of periodic inspection duration and amount of
improvement and modification work.
Table Collective Dose per Unit / Reactor-Year
Fiscal year

2000

2001

2002

2003

2004

2005

2006

BWR
(man-Sv)

1.96

1.68

2.10

2.40

1.58

1.39

1.33

PWR
(man-Sv)

1.03

1.27

1.00

1.07

1.25

0.97

1.08

Trend of Collective Dose per Unit / Reactor-Year is shown below.

15-5

man-Sv/Ry 3.0

BWR
PWR

2.5
2.0
1.5
1.0
0.5
0.0
2000

2001

2002

2003

2004

2005

2006
FY

FigureCollective Dose per Unit / Reactor-Year

3) Dose Limits for the Public
The dose limits for the public are also given in the Dose Limit Notification as listed inTable 15-2.
Table 15-2 Dose limits for the public
Items

Limit

Dose limits outside the peripheral monitoring area
Effective dose
Equivalent dose for eye lensEquivalent dose for skin

1 mSv/ year
15 mSv/year
50 mSv/ year

(2) Conditions for Discharge of Radioactive Materials
1) Dose Target and Discharge Control to Reduce Dose of the Public in the Vicinity(ALARA)In the Regulatory Guide for the Annual Dose Target for the Public in the Vicinity ofLight Water Nuclear Power Reactor Facilities, the NSC has prescribed a numericalguide of 0.05 mSv, one twentieth of the dose limit to the public, in order to reduce thedose for the public due to discharge of radioactive material to the environment during
normal operation of a nuclear installation as low as reasonably achievable.

15-6

The licensee, in order to achieve the target, establishes an annual numerical discharge
control guide, which corresponds to the annual discharge amount evaluated at thesafety review and assessment, and makes the effort to keep the discharge of radioactive
effluents below the numerical discharge control guide. NISA acknowledges thenumerical discharge control guide and receives the report from the licensee.
2) The Discharge Data and the Measures Taken to Reduce the Amount of the Discharge
The discharge records of radioactive gaseous and liquid waste from the nuclear
installations (BWR and PWR) over the past seven years are shown in Tables 15-3 to
15-5. The tables clearly show that the discharge quantities are substantially below thenumerical discharge control guide, the noble gas discharge from the PWR being onlyone 1000th of the dose target. This is due to the fact that the licensees have carried out
the radiation management of the nuclear installation in line with the ALARA principle,
including the following measures.
Gaseous waste is discharged from the ventilation port, while being measured andmonitored, after particles are removed by a high efficiency filter, noble gas and iodine
are decayed in a holdup tank or activated carbon type noble gas hold-up device.
All liquid waste is collected in a disposal facility, and the equipment drain is recovered
after being processed in an equipment filter or demineralizer. The floor drain isrecovered after being processed in a concentrator and demineralizer. Floor drain isreused in general, though part of it may be discharged through the discharge outlet
after the concentration is measured. The recovered liquid waste from the resin is reused
after being treated in a concentrator and demineralizer. Concentrated liquid waste
generated in this process is treated as solid waste. Low-level laundry wastewater, etc.
are usually drained into the environment after being treated through a filter and then it
is monitored.
In addition to the measures shown in the paragraphs before, a very low level of gaseousdischarge and liquid radioactive waste were the results of the following efforts, thesubstantial reduction of the possibility of a fuel leak by the improvement of fuels, (soonly three cases with four fuel assemblies of fuel leak arose during the period ofreporting), filtering ventilation during periodic inspections through local highefficiency filter.
(3)

Environmental Radiation Monitoring
The licensee is required to install environmental radiation monitoring equipment
during the normal operation of a nuclear installation. This equipment includes
monitoring devices of the dose inside the radiation control area and outside theperipheral monitoring area and automatic devices to alarm any abnormal increase ofconcentration of radioactive materials or dose rates.

15-7

The licensee conducts radiation monitoring at the site vicinity during normal operation,assesses the impact upon the environment of the discharge of radioactive materials
from the nuclear installation, and feedbacks the results in improving discharge control
and facility management. Local governments (prefectures where nuclear installations
are located) also monitor the radiation level independently at the site vicinity to protect
public health and safety.
The NSC decided the fundamentals of planning and implementation of the monitoring
and the evaluation of radiation dose in the Guide for Environmental Radiation
Monitoring, in order to improve and to standardize the monitoring technology. Local
governments and licensees implement monitoring in accordance with this guide.
15.3
(1)

Regulatory Control Activities
Discharge Control of Radioactive Materials
By the Rules for Commercial Power Reactors, the licensee is required to report
immediately to NISA when a concentration of radioactive materials in the air outside
the peripheral monitoring area exceeds the allowable limit in discharging gaseousradioactive waste, or when the concentration of radioactive materials in the water at theouter boundary of the peripheral monitoring area exceeds the allowable limit indischarging liquid radioactive waste, and report the status of the event and measurestaken against it within ten days.

(2)

Control of Personal Exposure
The Rules for Prevention of Hazards from Ionizing Radiation require the licensee tomeasure the dose due to external and internal exposure of workers who are engaged inradiation work or in emergency work, or enter temporal access into the radiation
control area. The rules require that the licensee monitor and check daily the dose due toexternal exposure, if it is expected to exceed the specified value of 1 mSv at 1 cm dose
equivalent, and to calculate, without delay, the dose of the personnel engaged inradiation work using the method prescribed by the Minister of Health and Labor, and tokeep these records for a period of thirty years.
The Radiation Workers’ Registration Center of the Association of Radiation Impact
was established in November 1977, to address the difficulty of controlling the personaldose of each worker who works in more than one radiation environment. The Center
unitarily collects and controls such personal radiation control data of each worker who
works under the Reactor Regulation Law, with personal identification control, apersonal radiation control booklet, periodical dose registration and transfer andcustody of personal radiation dose record.

15-8

(3) Control of Collective DoseThe trend of collective dose in Japan after 1990s has generally continued to be flat ora gradual increase, while that of Western countries has gradually decreased. Though
the decreasing trend is observed after 2004 with the decrease of modification work, thetrend through the whole period still remains at a higher level as compared with that ofWestern countries.
In view of the recent situation, a study of radiation exposure at nuclear power plants inboth Japan and Western countries was carried out and the effort for dose reduction
based on the ALARA principle were also investigated in order to clarify the reasons ofdifference between Japan and Western countries and also to identify the issues for dose
reduction (from 2004 Fy to 2007 Fy).
As the results and reasons for differences of the collective dose between Japan andWestern countries were clarified, differences in workload during plant outage (amountof construction/modification work and the number of workers), length of operation
cycle and maintenance criteria, e.g. a long term operation of 18 to 24 months ispermitted at some plants in Western countries and 13 months operation is permitted as
the maximum in Japan, and the plant outage period is about 2 to 3 months in Japan,
which is about two times longer than that of Western countries. And, the results ofinvestigation on the efforts for dose reduction based on ALARA principle identified thesubjects for the optimization of collective dose management, such as medium and longterm dose reduction strategy, information sharing between licensees and for ALARA,
way of regulatory intervention to the efforts of licensees etc..
Though at present the collective dose level itself (about 1.1 man-Sv) is not a matter ofconcern, at present the individual dose (average annual dose of recent years: about 1
mSv, maximum individual dose: about 20 mSv) is below the dose limits (100 mSv / 5
years and 50 mSv / year), it is important to promote activities for collective dose
reduction continuously based on the ALARA principle.
The regulatory body will precede the study of solid control measures of collective dose,such as development of the diagnostic evaluation method focused on the dose reduction
processes in order to stimulate the licensees’ activities.

15-9

Table 15-3 Annual discharge of radioactive noble gas in gaseous waste(unit: Bq / year)
Year
1998

1999

2000

2001

2002

2003

2004

2005

2006

N.D.*

N.D.

N.D.

N.D.

N.D.

N.D.

N.D.

N.D

N.D

N.D

6.7 × 10 15

6.2 × 10 9

2.9 × 10 9

3.7 × 10 15

StationStation - A

4.3×10 11 6.1×10 11 1.2×10 11 5.7×10 10 1.5×10 10 2.8×10 10 1.8×10 11 4.1 × 10 11

Station - B

Numerical
Discharge
Control
Guides1997* N.D. indicates a value below the detection limit concentration of 2 x 10 -2 Bq/cm 3 .
Table 15-4 Annual discharge of radioactive iodine (I-131) in gaseous waste(unit: Bq / year)
Year

Numerical
Discharge
Control
Guides

1997

1998

1999

2000

2001

2002

2003

2004

2005

2006

Station - A

N.D.*

N.D.

N.D.

N.D.

N.D.

N.D.

N.D.

N.D

N.D

N.D

2.3 × 10 11

Station - B

8.6×10 5 1.2×10 5 1.6×10 5 1.1×10 6 2.7×10 5

N.D

N.D

1.9 × 10 8

N.D

N.D

1.0 × 10 11

Station*: N.D. indicates a value below the detection limit concentration of 7 x 10 -9 Bq/cm3 .

Table 15-5 Annual discharge of radioactive materials (excluding 3 H) in liquid waste(unit: Bq / year)
Year

Numerical
Discharge
Control
Guides

1997

1998

1999

2000

2001

2002

2003

2004

2005

2006

Station - A

N.D.*

N.D.

N.D.

N.D.

N.D.

N.D.

N.D.

N.D.

N.D.

N.D

2.5 × 10 11

Station - B

N.D.

N.D.

N.D.

N.D.

N.D.

N.D.

N.D.

N.D.

N.D.

N.D

1.4 × 10 11

Station*: N.D. indicates a value below the detection limit concentration of 2 x 10 -2 Bq/cm 3 .
(Represented by 60 Co)(Note) Station - A: Kashiwazaki-Kariwa NPS (BWR), Station - B: Ohi PS (PWR)

15-10

Article 16 Emergency Preparedness
1. Each Contracting Party shall take the appropriate steps to ensure that there are on-site andoff-site emergency plans that are routinely tested for nuclear installations and cover the activities
to be carried out in the event of an emergency. For any new nuclear installation, such plans shallbe prepared and tested before it commences operation above a low power level agreed by theregulatory body.
2. Each Contracting Party shall take the appropriate steps to ensure that, insofar as they arelikely to be affected by a radiological emergency, its own population and the competentauthorities of the States in the vicinity of the nuclear installation are provided with appropriate
information for emergency planning and response.
3. Contracting Parties which do not have a nuclear installation in their territory, insofar as they
are likely to be affected in the event of a radiological emergency at a nuclear installation in thevicinity, shall take the appropriate steps for the preparation and testing of emergency plans fortheir territory that cover the activities to be carried out in the event of such an emergency.
On emergency preparedness, the Special Law of Emergency Preparedness for Nuclear Disaster(hereinafter referred to as the “Special Law for Nuclear Emergency”) was enacted in December 1999,
incorporating the lessons learned from the JCO criticality accident.
Considering the special characteristics of a nuclear emergency, measures for nuclear emergency
preparedness have been defined to cope within the existing legal framework established by the Basic
Law on Emergency Preparedness, which has defined such preparedness for disasters as earthquakes,
typhoons, and conflagrations etc.
In May 2007, the Nuclear Safety Commission (hereinafter referred to as the NSC) revised the“Emergency Preparedness for Nuclear Installations” (hereinafter referred to as the “Emergency
Preparedness Guides”), which specified technical and special matters of nuclear emergency measures,
based on international trends such as safety requirements GS-R-2 published by the IAEA.
In addition, as the Special Law for Nuclear Emergency provides that its enforcement situation is
subject to review five years after its enforcement, the enforcement situation was investigated by MEXT
and METI. The investigation results were reported to the Special Committee on Nuclear Disaster, the
NSC.
Moreover, during the reporting period, the so called "Civil Protection Law" to protect the people inarmed attack situations etc. was enforced in September 2004. Since nuclear installations are also
included in the target-of-attack facilities, some training incorporating the provisions of the law wereconducted during some nuclear emergency exercises.
16.1. Development of Laws and Rules for Nuclear Emergency Preparedness
For Japan who was promoting the utilization of nuclear energy under the basic premise of securingsafety, the JCO criticality accident which occurred in September 1999 was the first serous accident of
its kind, and it was so serious that local residents were instructed for sheltering or evacuation. Lessons
learned from this accident clarified the special characteristics of a nuclear emergency, which would
16-1

require quick initial responses, close coordinated cooperation between the national government andlocal governments, strengthening of the national emergency system and the clarification of licensee's
responsibilities. The Special Law for Nuclear Emergency was enacted in December 1999 and it was
enforced in June 2000, addressing the special characteristics of nuclear emergencies as mentioned
above. The Special Law for Nuclear Emergency was enacted so as to harmonize with the existing legal
framework established by the Basic Law on Emergency Preparedness, which had defined the roles of
the national government, local governments, etc. in emergencies such as earthquakes, typhoons, andconflagrations.
The “Nuclear Emergency Preparedness” in the Basic Plan for Emergency Preparedness based on the
Basic Law on Emergency Preparedness, was extensively revised in accordance with the Special Law
for Nuclear Emergency, clarifying roles and responsibilities of the national government, local
governments, and licensees etc. The NSC, in May 2000, also taking into consideration of the Special
Law for Nuclear Emergency and the lessons learned from the JCO criticality accident, revised the"Emergency Preparedness Guides" on technical and special matters of nuclear emergency measures, toinclude:−Research reactors and nuclear fuel cycle facilities in addition to commercial powerreactors; and,−Accidental release of nuclear fuel material during transport of nuclear fuel, etc. inaddition to release of noble gas and iodine from NPS etc.

After that, the Emergency Preparedness Guides have been enhanced through the following multiple
revisions by the NSC:−In March 2001, the dose coefficient (Sv/Bq) for internal exposure was changed alongwith the term, in response to the amendment of the relevant legislations such as theReactor Regulation Law etc. based on the adoption of the 1990 Recommendation of the
ICRP;

−

In June 2001, provisions of the emergency exposure medical treatment for exposed
patients was revised to be more effective and responsibilities of the national and local
governments and nuclear licensees were clarified based on the experience of thecriticality accident;−

In April 2002, protective measures concerning the taking of stable iodine tablets as aprevention were established based on the scientific knowledge acquired from thelong-term follow-up survey to atomic bomb sufferers and the investigation results of
the Chernobyl Power Station accident, etc;−

In November 2002, measures for mental health care in a nuclear emergency were
established based on the experience of JCO criticality accident, experiences of natural
disasters such as seismic disasters, etc; and,−In July 2003, the designation of a regional emergency exposure medical treatment
system was established.

16-2

In May 2007, the emergency measures were reviewed with reference to the IAEA Safety Requirement
GS-R-2, "Preparedness and Response for a Nuclear or Radiological Emergency", Safety Guide
GS-G-2.1, "Arrangements for Preparedness for a Nuclear or Radiological Emergency" (hereinafterreferred to as the "IAEA documents"), etc., and the following six items were revised:−

Characterization of the Emergency Preparedness Guides was clarified as "guides ontechnical and special matters specified by the NSC for the national and local
governments and nuclear licensees in preparing plans related to nuclear emergencies
and in taking protective measures during an emergency;"−

Situations for which the Emergency Preparedness Guides are applied were clarified as"nuclear emergencies at reactor facilities except the reactors for nuclear ship, fuelfabrication facilities, processing facilities, utilization facilities (limited to facilities thatuse nuclear fuel material equal to or exceeding the critical mass), waste disposal
facilities and waste storage facilities, and transportation of nuclear fuel materials etc;"−

The following four goals of protective measures were clarified referring to the IAEA
documents, and it was also clarified that in taking these protective measures it isimportant to assess them with the principles of “Justification of intervention” and“Optimization of intervention”;
1) To prevent the occurrence of deterministic health effects in residents in the vicinity,
to nuclear-installation workers, in those relevant in emergency preparedness, etc.,
2) To render first aid and to manage the treatment of radiation injuries,3) To prevent, to the extent practicable, the occurrence of stochastic health effects inthe population, and,
4) Reducing anxiety on the health of residents in the vicinity, workers, and to those
relevant in emergency preparedness.

−

The IAEA documents specify the precautionary action zone (PAZ) and the urgent
protective action planning zone (UPZ) as off-site emergency zones for whicharrangements shall be made for taking urgent protective action. In the results of thespecialists’ study, it was clearly written that this was also effective, in the emergency
measures of Japan, to implement precautionary/protective measures before or
immediately after the release of radioactive materials instead of setting up a new
specific zone as PAZ. It was confirmed that the setting up of UPZ was for the same
purpose as EPZ which had already been set up in the Emergency Preparedness Guidesof Japan as an area where protective measures should be focused on in theimplementation;−

The IAEA documents provide guidelines for the protective measures corresponding toavertable doses. In Japan, projected doses are used when implementing protective
measures. It was confirmed that making judgment using projected doses rather thanavertable doses as guidelines for the protective measures served as a response on thesafe side; and,
16-3

−

It was clearly written that the effect of taking stable iodine tablets as prevention and aprotective measure is appropriate only for the internal exposure by radioactive iodine
and the measure will complement protective measures, such as sheltering andevacuation.

As the Special Law for Nuclear Emergency provides that its enforcement situation is subject to review
five years after its enforcement, the enforcement situation was investigated by MEXT and METI.
Results of the investigation were reported to the Special Committee on Nuclear Disaster, the NSC in
March 2006.
The Nuclear and Industrial Safety Agency (NISA) checked the enforcement situation concerning four
issues that were presupposed to respect when the Special Law for Nuclear Emergency was enacted,
and reported the following:−

Concerning the speeding up of the initial response, non-scenario-based training should
be carried out, and the effort should be continued;−

Concerning enhancing the cooperation among the national government and local
governments, the "Integrated Nuclear Emergency Preparedness Network", which is alarge-scale system and preparation of a fast unified network of communication among
them, should be made;−Concerning enhancing the emergency response system of the national government,
necessary renewal of materials and equipment of the Emergency Preparedness Center
should be promoted; and,−In relation to clarification of the licensees’ duties, the effectiveness should be verified
and improved so that nuclear emergency specialists may achieve their required
functions in an emergency.

16.2. Nuclear Emergency Preparedness and the Measures
The Special Law for Nuclear Emergency has prescribed measures in a nuclear emergency at powerreactors, research reactors, nuclear fuel cycle facilities, etc. Emergency measures of commercial
nuclear installations are given below.
(1) Responsibilities of Related organizations concerning Nuclear Emergency Preparedness
1) Responsibility of the National Government
The national government prepares the necessary emergency preparedness and is ready to takemeasures in an emergency:−

METI stations a Senior Specialist for Nuclear Emergency in the vicinity of eachnuclear installation, who guides and advises the licensee in preparing his emergency
action plan and, in an emergency, takes the necessary measures in preventing expansionof the emergency;−

The NSC is mandated to give technical advice to the chief of the Nuclear Emergency
16-4

Response Headquarters (Prime Minister) on designation or alteration of regional areasthat necessitates emergency measures to be taken, and technical matters on theimplementation of emergency response measures and dissolution of a nuclear
emergency. For that purpose, the NSC organizes the “Technical Advisory Organizationin an Emergency” which consists of the NSC Commissioners and the Investigators for
Emergency Response;−

The Minister of METI designates a facility in the vicinity of a nuclear installation as an
Off-Site Center to be used in an emergency. In the case of an emergency, the national
government, the local governments and the licensee establish at the Off-Site Center the"Joint Council for Nuclear Emergency Response", in order to share information and tocoordinate their activities. Off-Site Centers are built on the areas indicated in Fig.
16-1, and have necessary facilities and equipment capable to communicate with thePrime Minister’s Official Residence, the Cabinet Office, the Emergency Response
Center of NISA, the Emergency and Emergency Preparedness Center of MEXT andrelated local governments;−

Each Off-Site Center is equipped with means by which the related organizations
monitor environmental radiation levels and the status of the nuclear installation. The
environmental radiation levels, other than temporary data measured in an emergency,
can be monitored at every moment since the monitoring equipment is connected on linewith the monitoring posts located in the vicinity of the nuclear installation. The on-line
status of the nuclear installation that is sent from the licensee in an emergency can bedisplayed on the monitor panels. The results of estimation are also displayed by meansof an Emergency Response Support System (ERSS), which forecasts progress of anabnormal condition of the nuclear installation using plant information.

−

The national government establishes arrangements to secure quick and coordinatedactivities in an emergency; and,−The national government conducts the comprehensive emergency drill based on the
program established by the competent minister.

2) Responsibilities of local governments
Each local government shall develop and revise the regional emergency preparedness plan inaccordance with Article 40 of the Basic Law on Emergency Preparedness, and shall consult
beforehand with the Prime Minister for its development or revision.
3) Responsibilities of licensees:−

Each licensee shall develop his Nuclear Licensee Emergency Action Plan after
consulting with relevant local governments, and submit it to the Minister of METI
before the commissioning of the reactor;−

Each licensee shall establish an on-site organization for nuclear emergency
preparedness, and designate a Nuclear Emergency Preparedness Manager who
administers the organization; and
16-5

−

The Nuclear Emergency Preparedness Manager shall notify specific initial events to thecompetent authorities.

(2) Measures for On-site and Off-site Nuclear Emergency Preparedness of Nuclear Installations
In order to prepare the “Nuclear Emergency Preparedness” described in paragraph (3), relatedorganizations keep themselves ready to collect and send information and also to start a quick response
against an emergency, conduct emergency drills, disseminate knowledge and promote research on
emergency preparedness. Outline of roles and responsibilities of related organizations are as follows.
1) On-Site Emergency Preparedness of Nuclear Installations
When the licensee detects abnormal release of radioactive material or an abnormal level ofradiation at a nuclear installation, he takes necessary measures to prevent progression of the event
into an emergency.
The licensee, to cope with the emergency properly, prepares the Nuclear Licensee Emergency
Action Plan after consulting with related local governments, which provides for the prevention of,
emergency measures against, and post-emergency restoration of a nuclear emergency, including
on-site and off-site cooperation with other organizations. Especially, quick and accurate
notification of occurrence of specific initial events to related organizations is a very important
obligation of the licensee.
The licensee is required to take part in comprehensive drills with related organizations, and keep
close contact with them.
2) Emergency Preparedness in the Vicinity of Nuclear Installations
Roles and responsibilities of the national government and local governments in emergency
preparedness in the vicinity of nuclear installations are defined in the Special Law for Nuclear
Emergency and the Basic Plan for Emergency Preparedness. Each local government develops its
own regional emergency preparedness plan. They carry out emergency environmental radiation
monitoring, and implement evacuation or sheltering of residents receiving advice or direction
from the Prime Minister based on the report of the Minister of METI. The taking of stable iodine
tablets for prevention, as well as sheltering or evacuation, are defined as some of the protective
measures.
(3) Nuclear Emergency Preparedness concerning Nuclear Installations (Fig. 16-2)
Quick initial response and closely coordinated cooperation among relevant organizations areimportant in a nuclear emergency:−

The Special Law for Nuclear Emergency defines specific initial events in a nuclear
installation (see Table 16-1), the occurrence of which the licensee shall immediately
notify the Minister of METI and the heads of related local governments;−

The Minister of METI, receiving the notification, triggers activities according to theprocedure stipulated by law. Staff with expertise in emergency measures will be sent tolocal governments on request. The Senior Specialist for Nuclear Emergency collects
16-6information and coordinates activities preventing expansion of the events;−

When the Minister of METI recognizes that the specific initial event exceeds thepredetermined level and has developed into an emergency, the Minister immediately
reports it to the Prime Minister;−

The Prime Minister has powerful authority to declare a “Nuclear Emergency”, and toadvise or direct relevant local governments on necessary measures such as sheltering or
evacuation to be taken by them, as well as to request for dispatch of the Self-Defense
Forces concerning implementation of emergency measures;−

The Prime Minister establishes the "Nuclear Emergency Response Headquarters" inTokyo, which he will head, and the "Local Nuclear Emergency Response
Headquarters" at the site;−

In a nuclear emergency, the NSC convenes with the “Technical Advisory Organizationin an Emergency” that is composed of commissioners and the Advisors for Emergency
Response. The Organization gives technical advice to the Prime Minister;−

Local governments establish their own emergency response headquarters; and,−In order to share information between the national government and relatedorganizations such as local governments, nuclear licensees, etc., and, if necessary, tocoordinate emergency measures to be implemented by the respective organizations, the"Joint Council for Nuclear Emergency Response" is to be established at the Off-SiteCenter.

16.3. Implementation of Nuclear Emergency Drill
The emergency preparedness action plan in accordance with the Basic Law on Emergency
Preparedness, and the Off-Site Center in the vicinity on the nuclear installation provided in theprevious section has been established for each nuclear installation, and a nuclear emergency drill is
implemented to confirm the effectiveness of these measures. The purpose of the nuclear emergency
drill is 1) to enhance understanding of, and adequate actions for, nuclear emergency preparedness by
responsible personnel of the national government, local governments, the licensee, and residents, and2) to verify whether emergency measures function in a predetermined way, and whether information
sharing and cooperation among related organizations are adequate. The national government, local
governments, designated public organizations and the licensee cooperate and participate in the drill,which cover communication, monitoring, decision on emergency measures to be taken, sheltering or
evacuation etc., ranging from a large scale national drill to the licensee’s on-site drill. Drills
implemented in past years are shown below.
(1) Drills Planned by the National Government (Table 16-2 (1))
Nuclear emergency drills used to be planned and conducted by local governments with support andcoordination of the national government before the JCO criticality accident. The Special Law for
Nuclear Emergency stipulated the drills to be planned and conducted by the national government.

16-7

Drills including accident management activities assuming a scenario resulting in core damage havebeen implemented in the national emergency drills.
The drill planned by the national government has been conducted once a year as the comprehensive
nuclear emergency drill in collaboration with the national government, local governments, licensees,
etc.
Drills implemented during the reporting period are as follows:
As for 2004, although a drill for the Kashiwazaki Kariwa Nuclear Power Station, Tokyo ElectricPower Co., Inc. was scheduled on November 1 and 2, it was canceled due to the impact of the
Niigata-ken Chuetsu earthquake which occurred just before conducting the drill.
As for 2005, the drill for Unit 4 of the Kashiwazaki Kariwa Nuclear Power Station (located in
Kashiwazaki City and Kariwa Village, Niigata Prefecture) which had been canceled in 2004 was
conducted on November 9 and 10 in collaboration with the national government, the local
governments of Niigata Prefecture and relevant municipalities, Tokyo Electric Power Co., Inc., andorganizations related to the emergency preparedness. About 2,600 persons including about 400 local
residents participated in the drill.
On October 25 and 26, 2006, a drill for Unit 3 of the Ikata Power Station, Shikoku Electric Power Co.,
Inc. (located in Ikata Town, Ehime Prefecture) was conducted in collaboration with the national
government, local governments of Ehime Prefecture and relevant municipalities, Shikoku ElectricPower Co., Inc., and organizations related to the emergency preparedness. About 3,700 persons
including about 300 local residents participated in the drill. In this instance, while verifying urgent
dispatch and urgent conveyance to a remote place, speeding up of determining evacuation areas wasachieved by verifying actions for emergency measures in an alert stage.
Results of the drills held every year are assessed and reflected to the items and methods of drills to be
implemented in and after the following fiscal year. Three kinds of methods, a participant’s
questionnaire, check by an independent assessment agency, and observation by external experts, areadopted for the assessment.
(2) Drills Planned by the NSC
The NSC is conducting communication drills that aim enhancing the emergency communication
system and keeping up and/or improving its functions. The NSC also is conducting field-training drillsof a Technical Advisory Organization in an Emergency that aim to confirm the emergency response
capability and improve its effectiveness.
(3) Drills Planned by Local Government (Table 16-2 (2))
The regional emergency preparedness plan prescribes the local drills to be planned and conducted byeach local government, which NISA and the NSC support by dispatching expert staff. Drills reflecting
the framework of the Civil Protection Law are also conducted in part of those drills.
(4) Drills Planned by Licensees
Each licensee has implemented an on-site drill once a year including establishment of an emergency
16-8

response headquarters, notification and communication, emergency environmental radiation
monitoring, etc. based on the Nuclear Licensee Emergency Action Plan defined for each place ofbusiness.
Each licensee also has implemented a drill taking into consideration the accident management
activities, if necessary, in order to comprehensively confirm effectiveness of the organization
implementing the accident management.
When the said place of business is subject to the drill conducted by the local government, the on-site
drill has been implemented at the same time with the local drill implemented by the local government
etc.
16.4

International Framework and Relationship with Neighboring Countries

Japan is a contracting party to the Convention on Early Notification of a Nuclear Accident, and to theConventions on Assistance in the Case of a Nuclear Accident or Radiological Emergency. The
following domestic implementation system has been established for the notification of a nuclear
accident to neighboring countries:−

For nuclear installations, the Ministry of Foreign Affairs has been designated as theauthority for notification and as the competent authority for foreign accidents, and
METI has been designated as the competent authority for domestic accidents;−

METI receives a report immediately upon an accident in a nuclear installation on thebasis of legal obligation of the licensee; and,−

When an accident occurrence is confirmed and it is predicted that release of radioactive
materials may affect neighboring countries, the IAEA and the countries that may beaffected by the accident are notified of the accident.

In accordance with an arrangement aiming at enhancement of the safety level of commercial nuclear
installations concluded with the People's Republic of China, on the basis of the bilateral agreement forpeaceful use of nuclear energy, the governments should mutually notify without delay of major
accidents of nuclear installations. A bilateral inter-governmental agreement with the Republic of Korea
calls for cooperation in the establishment and operation of an early notification network for nuclear
safety.
If an accident should occur at a foreign nuclear installation and a request for assistance is made, Japan
will dispatch, on the basis of the Conventions on Assistance in the Case of a Nuclear Accident orRadiological Emergency, specialists in emergency monitoring and emergency exposure medical
treatment, and will provide materials and equipment such as radiation measurement equipment andradiation protection equipment.
In addition, Japan has supported Asian countries to enhance their infrastructures of emergency
countermeasures and response for nuclear installations in an operation of the Asian Nuclear Safety
Network (ANSN), which is a part of the IAEA cooperation businesses to Asia, and has cooperated inworkshops, such as emergency medical treatment and emergency drills.

16-9

Kyowa Town

Fukui Pref.
Higashidori Village(Preparing for Service)
Tsuruga City
Mihama Town
Ohi Town

Takahama
Town

Onagawa Town
Kashiwazaki City
Shika Town

Ohkuma Town
Hitachinaka City

Matsue City

Karatsu City

Omaezaki City▲:Off-Site Center in theVicinity of Nuclear Power
Plants (16 Sites)

Ikata Town

Satsuma-sendai City

Fig. 16.1 Location of Off-Site Centers

16-10

Licensee
Nuclear Emergency
Preparedness
ManagerNational Government
NoticeLocal Government

Competent Minister

Notification- to prepare response organization to theevent reported- to judge occurrence of nuclear
emergency situation(Table 16-1 for judgment criteria)- to collect information and conduct
necessary action by the resident
Senior Specialist for Nuclear
Emergency- Recognition of events
to be reported(Table 16-1 forreport criteria)- Initial response
against emergency byonsite nuclear
emergency response
team

Relatedprefectural
governors
Request ofexperts
Dispatchof experts

Municipal
governors

Prime Minister- to declare “Nuclear Emergency”- to announce affected zone, outline ofaction, etc.
- to establish Nuclear Emergency
Response Headquarters

Advice ordirection ofevacuation &sheltering

Residents

NuclearAdviceSafetyCommission

Nuclear Emergency Response
Headquarters
Director General: Prime Minister
Deputy Director: Minister of METI
Member: Minister of related ministries
Government offices
Directors of related agencies

Direction

Establishments ofEmergency Response
Headquarters of each
local government

Dispatch/Advice
Local Nuclear Emergency Response
Headquarters

Self-Defense
ForcesEstablishment of Joint Council for Nuclear Emergency Response
To be established at Off-Site Center to share informational and coordinate
emergency response activities conducted by each organization
Emergency response
activities by on-site
emergency response
team in licensee

Nuclear Safety
Commission

Main activities of Joint Council- Accurate public information- Advice or directions of evacuation or
sheltering of residents- Measurement of radiation inenvironment- Rescue of sufferers- Restoration of facility

Advice- Special supports: JAEA andNational Institute of Radiological
Sciences- Fire extinguishing, Rescue andFirst-aid: Fire and Disaster
Management Agency- Security guard: National Police Agency

Prime Minister
To declare closing of“Nuclear Emergency”

Abolition of Nuclear Emergency
Response Headquarters

Fig. 16-2

Directions and requests

Measures by national and local
governments and licensees- Contamination investigation,- Health consultation,- Rumor damage control

Nuclear Emergency Preparedness based on the Special Law for Nuclear Emergency

16-11

Table 16 – 1Main Specific Events and the Nuclear Emergency specified in the Special Law for Nuclear Emergency

EventsCriteria for reporting by licensees and “Nuclear Emergency” declaration by national government

Events that licensees should report.

a) Dose of radiation near the siteboundary detected doseb) Detection of the radioactive materials
in usual release points, such as exhaustpipesc) Radiation by fire, explosion, etc. or
detection of radioactive materials (outside
the control zone)d) Individual events of each nuclear
installation (Example for reactor)- Failure of scram- Loss of reactor coolant

16-12- Loss of all AC power supplies

5 micro Sv/h at one point for more than 10 min
5 micro Sv/h at more than 2 points at the same time
Radioactive materials worth more than 5 micro Sv/hconditions of declaration of “Nuclear Emergency”
500 micro Sv/h at one point for more than 10 min,
500 micro Sv/h at more than 2 points at the same time
Radioactive materials worth more than 500 micro Sv/hRadiation dose of more than 50 micro Sv/h
Release of radioactive materials worth more than 5 micro Sv/hRadiation dose of more than 5 mSv/h
Release of radioactive materials worth more than 500 micro Sv/h

When the nuclear reactor shutdown cannot be performed byusual neutron absorbers
Occurrence of leakage of nuclear reactor coolant which needs
operation of the emergency core coolant system (ECCS)
When all AC power supplies stop power supply for more than 5
minutes

When all of reactor shutdown functions are lost
When water cannot be poured to the nuclear reactor by all ECCSs
When all measures for the cooling reactor core are lost with loss of all AC
power supplies.

Response of the national government- The Minister of METI sends staff with expertise on request - The Minister of METI reports the nuclear emergency to the Prime Minister
immediately after confirming the situation.
of local governments.
- The Prime Minister declares “Nuclear Emergency” and takes the- The resident Specialist on Nuclear Emergency Preparedness
following responses;
carries out necessary work.
- to advice or direct related local governments on necessary measures
The following responses are carried out based on the agreement
such as sheltering or evacuation.
of related ministries, not specified in the Special Law for Nuclear- to establish the Nuclear Emergency Response Headquarters in Tokyo
Emergency.
and Local Nuclear Emergency Response Headquarters at Off-Site- Related ministries and agencies organize a joint task group forCenter.
the incidents in Tokyo.
- to establish the Joint Council for Nuclear Emergency Response for- Related local organizations organize a joint local task group in
information exchange among the national government and local
the Off- Site Center.
governments.

Table 16-2 Nuclear Emergency Drills
Conductor
Date
Nuclear Power Station(1) Drills conducted by the National Government (April 2004 - March 2007)
METI
11/09/2005 (Wed.)
Kashiwazaki Kariwa Nuclear Power Station–11/10/2005 (Thu.)(Tokyo Electric Power Co., Inc.)
METI
10/25/2006 (Wed.)
Ikata Power Station (Shikoku Electric Power Co., Inc.)–10/26/2006 (Thu.)(2) Drills conducted by Local Governments (April 2004 - March 2007)
Shizuoka Pref.
06/29/2004 (Tue.)
Hamaoka Nuclear Power Station (Chubu Electric Power
Co., Inc.)
Shimane Pref.
10/08/2004 (Fri.)
Shimane Nuclear Power Station (The Chugoku ElectricPower Co., Inc.)
Miyagi Pref.
10/19/2004 (Tue.)
Onagawa Nuclear Power Station (Tohoku Electric Power
Co., Inc.)
Hokkaido
10/22/2004 (Fri.)Tomari Power Station (Hokkaido Electric Power Co., Inc.)
Ehime Pref.
10/26/2004 (Tue.)
Ikata Power Station (Shikoku Electric Power Co., Inc.)
Aomori Pref.
11/16/2004 (Tue.)
Higashidori Nuclear Power Station (Tohoku ElectricPower Co., Inc.)
Saga Pref.
11/22/2004 (Mon.)Genkai Nuclear Power Station (Kyushu Electric Power
Nagasaki Pref.
Co., Inc.)
Fukushima Pref.
11/24/2004 (Wed.)
Fukushima Daiichi Nuclear Power Station(Tokyo Electric Power Co., Inc.)
Kagoshima
01/30/2005 (Sun.)
Sendai Nuclear Power Station (Kyushu Electric Power
Pref.
Co., Inc.)
Fukui Pref.
03/21/2005 (Mon.) Takahama Power Station (The Kansai Electric Power Co.,
Kyoto Pref.
Inc.)
Ishikawa Pref.
03/24/2005 (Thu.)
Shika Nuclear Power Station (Hokuriku Electric PowerCo.)
Fukui Pref.
08/02/2005 (Tue.)
Mihama Power Station (The Kansai Electric Power Co.,Inc.)
Aomori Pref.
08/10/2005 (Tue.)
Higashidori Nuclear Power Station (Tohoku ElectricPower Co., Inc.)
Hokkaido
10/21/2005 (Fri.)Tomari Power Station (Hokkaido Electric Power Co., Inc.)
Ehime Pref.
10/21/2005 (Fri.)Ikata Power Station (Shikoku Electric Power Co., Inc.)
Miyagi Pref.
10/28/2004 (Fri.)
Onagawa Nuclear Power Station (Tohoku Electric Power
Co., Inc.)
Niigata Pref.
11/09/2005 (Wed.)
Kashiwazaki Kariwa Nuclear Power Station–11/10/2005 (Thu.)(Tokyo Electric Power Co., Inc.)
Fukushima Pref.
11/14/2005 (Mon.)
Fukushima Daini Nuclear Power Station–11/15/2005 (Tue.)(Tokyo Electric Power Co., Inc.)
Ishikawa Pref.
11/17/2005 (Thu.)
Shika Nuclear Power Station (Hokuriku Electric Power
Co.)

16-13

Shimane Pref.

11/17/2005 (Thu.)

Shimane Nuclear Power Station (The Chugoku ElectricPower Co., Inc.)

Kagoshima
Pref.

11/19/2005 (Sat.)

Sendai Nuclear Power Station (Kyushu Electric Power
Co., Inc.)

Saga Pref.

11/21/2005 (Mon.)Genkai Nuclear Power Station (Kyushu Electric Power
Co., Inc.)

Fukui Pref.

11/27/2005 (Sun.)

Mihama Power Station (The Kansai Electric Power Co.,Inc.)

Shizuoka Pref.

02/15/2006 (Wed.)

Hamaoka Nuclear Power Station (Chubu Electric Power
Co., Inc.)

Fukui Pref.

08/01/2006 (Tue.)Ohi Power Station (The Kansai Electric Power Co., Inc.)

Ishikawa Pref.

08/20/2006 (Sun.)

Shika Nuclear Power Station (Hokuriku Electric PowerCo.)

Ibaraki Pref.

09/29/2006 (Fri.)

Tokai No.2 Power Station (The Japan Atomic Power Co.)

Miyagi Pref.

10/23/2006 (Mon.)–10/24/2006 (Tue.)

Onagawa Nuclear Power Station (Tohoku Electric Power
Co., Inc.)

Hokkaido

10/30/2006 (Mon.)

Tomari Power Station (Hokkaido Electric Power Co., Inc.)

Niigata Pref.

11/10/2006 (Fri.)

Kashiwazaki Kariwa Nuclear Power Station(Tokyo Electric Power Co., Inc.)

Aomori Pref.

11/14/2006 (Tue.)

Higashidori Nuclear Power Station (Tohoku ElectricPower Co., Inc.)

Kagoshima
Pref.

11/17/2006 (Fri.)

Sendai Nuclear Power Station (Kyushu Electric Power
Co., Inc.)

Fukui Pref.

11/19/2006 (Sun.)Ohi Power Station (The Kansai Electric Power Co., Inc.)

Saga Pref.

11/26/2006 (Sun.)

Genkai Nuclear Power Station (Kyushu Electric Power
Co., Inc.)

Shimane Pref.

01/30/2007 (Tue.)

Shimane Nuclear Power Station (The Chugoku ElectricPower Co., Inc.)

Shizuoka Pref.

02/01/2007 (Thu.)

Hamaoka Nuclear Power Station (Chubu Electric Power
Co., Inc.)

Fukushima Pref.

02/06/2007 (Tue.)–02/07/2007 (Wed.)

Fukushima Daiichi Nuclear Power Station(Tokyo Electric Power Co., Inc.)

16-14

D. Safety of Installations

This page intentionally left blankArticle 17 Siting
Each Contracting Party shall take the appropriate steps to ensure that appropriate
procedures are established and implemented:(i) for evaluating all relevant site-related factors likely to affect the safety of a nuclear
installation for its projected lifetime;(ii) for evaluating the likely safety impact of a proposed nuclear installation on
individuals, society and the environment;(iii) for re-evaluating as necessary all relevant factors referred to in sub-paragraphs (i)and (ii) so as to ensure the continued safety acceptability of the nuclear installation;(iv) for consulting Contracting Parties in the vicinity of a proposed nuclear installation,
insofar as they are likely to be affected by that installation and, upon request
providing the necessary information to such Contracting Parties, in order to enable
them to evaluate and make their own assessment of the likely safety impact on theirown territory of the nuclear installation.

In Japan, in order to judge the suitability of the site for a nuclear installation, it is deemednecessary to conduct the safety impact assessment of natural phenomena and human induced
external events to the nuclear installation, the safety impact assessment on the postulated
accident of the nuclear installation to the general public in the vicinity, and the impact
assessments on the siting of the nuclear installation to the environment other than the safety,
therefore, required legislations and regulations for implementing relevant assessments has
been provided, and the assessments are being conducted.
17.1 Basic Concept on the Siting of Nuclear Installations

The following assessments must be taken into consideration when deciding upon the siting of
nuclear installations, and are incorporated in the relevant legislation, etc.
ySafety impact assessment on a nuclear installation by natural phenomena andpostulated human induced external eventsy

Safety impact assessment of the radioactive impact to the environment by a nuclear
installation should reactor accidents occuryAssessment on environmental impact due to the siting of a nuclear installation

17.2 Principal Assessment System Concerning the Siting of Commercial Power Reactors

The Reactor Regulation Law requires that location of a Commercial Power Reactor must be
selected and its structure and equipment must be designed so that the radiological hazards canbe prevented. The adequacy of siting is examined in accordance with the Regulatory Guide for
Reviewing Nuclear Reactor Siting Evaluation and Application Criteria (hereinafter called, theRegulatory Guide for Reviewing Nuclear Reactor Siting Evaluation and application Criteria)
etc. as part of safety examination of licensing for establishment.

17-1

The Regulatory Guide for Reviewing Nuclear Reactor Siting Evaluation and application
Criteria requires that no such event that might induce serious accidents has occurred in thepast or could be expected to occur in the future at the proposed site and furthermore, there
should not be events that may aggravate accidents, the reactors are located at a sufficientdistance away from the public in terms of safety and protection facilities, and the site and thevicinity are in suitable circumstances to take, if needed, measures protecting the public.
When deciding a site, an adequate attention in design shall be paid to the events caused byexternal factor specific to the site, in addition to the site conditions stipulated by theRegulatory Guide for Reviewing Nuclear Reactor Siting Evaluation and application Criteria.
In this respect, the Regulatory Guide for Reviewing Safety Design of Light Water Nuclear
Power Reactor Facilities (hereinafter called as the “Regulatory Guide for Reviewing Safety
Design”) prescribes that structures, systems and components with safety functions shall be
designed to sufficiently withstand appropriate design basis earthquake forces. As well, theyshall be so designed that the safety of the Commercial Power Reactor will not be impaired byother possible natural phenomena than earthquake and also by postulated human induced
external events.
The Regulatory Guide for Reviewing Nuclear Reactor Siting Evaluation and application
Criteria also prescribes that the dose to the public shall meet with the application criteria in
consideration of the engineered safety features by establishing an non-residential area and low
population zone and ensuring sufficient distance from high population zones, when assessing
radiation impact to the public in the vicinity imposed by the postulated accidents in
Commercial Power Reactor. Meanwhile, the Regulatory Guide for Reviewing Safety
Assessment of Light Water Nuclear Power Reactor Facilities provides events to be evaluatedin siting, acceptance criteria and specific conditions, etc. to be used in the analysis.
Environmental impact assessment of all of the power stations including commercial nuclear
installation is performed in accordance with the Environmental Impact Assessment Law
enforced in June 1999, before when the departmental council decision of MITI (present METI)
dated in July 1977 was applied. This subject is described in Section 17.5.
17.3 Evaluation on Events Caused by External Factor

The Regulatory Guide for Reviewing Safety Design of Light Water Nuclear Power Reactor
Facilities prescribes that the earthquakes, natural phenomena other than earthquake andhuman induced external events shall be addressed in the design, being in accordance withthe prescription in the Regulatory Guide for Reviewing Nuclear Reactor Siting Evaluation
and application Criteria, stating “no such event that might cause serious accidents has
occurred in the past nor could be expected to occur in the future at the proposed site andfurthermore, there should not be events that may aggravate accidents”.
On the seismic design, it is required that the structures, systems and components (SSCs)
with safety functions shall be designed in accordance with seismic classification, and shallbe designed to maintain safety functions.

17-2

The NSC established the Sub-committee on Investigation of the Regulatory Guide for
Reviewing Seismic Design of Nuclear Power Reactor Facilities in July 2001, for the purposeof making the guide more appropriate with incorporation of the latest knowledge in theseismic safety guide used for a safety review and assessment, and promoted the amendment
study of the Guide. The Nuclear Safety Commission revised the Regulatory Guide and related
guides in September, 2006. Description on the new “Regulatory Guide for Reviewing Seismic
Design of Nuclear Power Reactor Facilities” is provided in Section 18.1.
For the assumed natural phenomena other than earthquake (floods, tsunami, breeze, freezing,snowfall, landslides, etc.), the SSCs with safety functions are required to be designed so thatthe safety of the nuclear reactor facility will not be failed by any of these natural phenomena.
Those SSCs with safety functions of particularly high importance shall be designed towithstand against the most severe conditions of natural phenomena or to withstand against
combination of such natural forces and loads induced by an accident.
Moreover, the SSCs with safety functions are required to be so designed that theCommercial Power Reactor should not be impaired by postulated human induced external
event (airplane crashes, dam collapse, explosions etc.).
On the consideration on airplane crash accident, a guide is provided to judge whether or not
it is necessary to take it into design consideration as “an assumed human induced external
event”, as well as the standard evaluation method, in “Evaluation Standards of theProbability of Airplane Falling to a Commercial Power Reactor Facility (NISA Regulatory
Guide)”, which NISA published as a Regulatory Guide in July 2002. Besides, for airplanes,
the flight over nuclear installations is prohibited in principle,
Commercial Power Reactors are required to be provided with appropriate measures toprevent illegal access to the reactor by third persons in Japan.
17.4 Evaluation for the Impacts to the Public of Accidents

In order to ensure safety of the public even in the case of the worst accident, the Regulatory
Guide for Reviewing Nuclear Reactor Siting Evaluation and application Criteria prescribes, asa fundamental siting condition, that a Commercial Power Reactor be located with a sufficientdistance from the public taking into account the engineered safety features. The conditions forfulfilling this requirement are as follows:A) The area within a specified distance from a Commercial Power Reactor shall be thenon-residential area, and no radiation hazard is imposed on the public in the vicinity
outside the non-residential area, even postulating the occurrence of the major accident.
The major accident is defined in the above Guide to be an accident, occurrence of whichis conceivable as a worst scenario from a technical point of view with considering suchfactors as the conditions at the site vicinity, the characteristics of the reactor and theengineered safety features.

17-3B) The area within a specified distance beyond the non-residential area shall be the low
population zone, and no substantial radiation hazard is imposed on the public in thevicinity of the low population zone, even postulating the occurrence of the hypotheticalaccident.
The hypothetical accident is defined in the above Guideline to be an accident, which
exceeds a major accident, and the occurrence of that is not conceivable from a technicalpoint of view. The Guide, for example, hypothesizes that some of engineered safety
features in the reactor, which are assumed to be effective in postulating a major accident,
do not function and corresponding release of radioactive materials occurs.

C) A site of a nuclear reactor shall be located at a specified distance from high population
zones.
The specified distance means a distance where cumulative value of whole-body dose incase of a hypothetical accident shall be small enough to be deemed acceptable based on
the collective dose of view.
The application criteria on dose rate are specified in the attachments of the Regulatory Guide
for Reviewing Nuclear Reactor Siting Evaluation and application Criteria. The meteorologicalobservation methods, the statistical processing methods of the observed data and the methods
for the analysis of the atmospheric diffusion of the released radioactive materials, to be usedin the dose assessments, are prescribed in the Regulatory Guide for MeteorologicalObservation for Safety Analysis of Nuclear Power Reactor Facilities.
17.5 Environmental Impact Assessment

The Environmental Impact Assessment Law was established to ensure business operators,
that are undertaking large-scale projects that could have a serious impact on the environment,
to conduct an environmental impact assessment properly and reflect the results of theassessment in implementing the project in term of protecting the environment, and also set
forth the procedures in conducting the environmental impact assessment. The assessment forcommercial power stations including a nuclear installation must be performed in accordancewith the provisions of the Environmental Impact Assessment Law and the Electricity Utility
Industry Law. All of nuclear installations are subject to assessment regardless of their scale.
Figure 17-1 shows an outline of procedures for environmental impact assessment concerning
establishment of a commercial nuclear installation.
Business operator, prior to application for reactor establishment, must prepare a Scoping
Document presenting information concerning the contents of the project, items to beconsidered in an environmental impact assessment, method of survey, prediction, andassessment method to be utilized, and must submit it to NISA, as well as to the local
governments having jurisdiction over the area deemed likely to be environmentally impacted
by the project. NISA examines the Scoping Document taking into consideration the comments
submitted by the related prefecture governor(s), as well as the comments of the residents andthe views of the business operators regarding such comments, and gives recommendations on

17-4

the contents of Scoping Document to the business operator, if needed.
Then business operator shall prepare a draft environmental impact statement (draft EIS) after
conducting survey, prediction and assessment in consideration with the recommendations
received from NISA and establishing the measures for protecting the environment. The draft
EIS must be submitted to NISA, as well as to the related local governments. NISA, after
examining the draft EIS, taking into account the opinion of the Minister of Environment andthe related Governors as well as the comments of the residents and the views of the business
operators regarding such comments, and receiving the view of advisers on the environment
protection, gives recommendations on the environmental assessment to the business operator
if needed. Meanwhile, concerning the items other than those with little environmental impact,business operators shall check and provide the necessary measures for protecting theenvironment so that the environmental impact by the project would be reduced as low aspractical, considering the project plan and the state of the area environmentally impacted by
the project.
Finally, business operator shall prepare the environmental impact statement (EIS), taking into
account the recommendation on the draft EIS, and submit it to NISA. NISA, after examining
the EIS, orders alteration of the EIS if needed, otherwise notices acceptance of the EIS tobusiness operator. The accepted EIS is distributed to the Ministry of Environment and related
local governments.
At the stage of examining construction plan, NISA does not approve it in case that the plan
does not conform to the accepted EIS.
17.6 Re-evaluation of Site Related Factors

All the factors related to site selection must be re-evaluated at the time of alteration of anestablishment license, such as additional new nuclear installation construction at theexisting site, so as to ensure the continuous safe operation of the nuclear installation.
Adequacy of the safety design is re-evaluated referring to new findings and new experiences
having impact on the design.
17.7 Arrangements with Neighboring Countries on Safety Impact of Nuclear Installations

Nuclear installation in Japan is so located at the place where there are no events liable toinduce serious accidents and so designed to secure the safety against postulated initiating
events including natural phenomena. It also implements the measures for the accident
management. Furthermore, because of the fact that Japan is an archipelago country andseparated from neighboring countries by a considerable distance, adverse impact of JapaneseCommercial Power Reactor over neighboring countries is deemed to be extremely small.
Accordingly, no consultation has been made so far with neighboring countries on the siting of
nuclear installations.

17-5

Residents

Ministry of
Economy, Trade andIndustry

Licensees

Ministry oftheEnvironment

Local
Governments(NISA)
Submittal of SD
Preparation ofSD

Examination ofSD (based on
Comments ofGovernor(s) and
Residents)(Advised by theAdvisory
Committee
on Environmental
ImpactSubmittal of SD

Announcement
andNotification
Outline ofComments withOperator’s
Position

Submittal of
Resident’s
Comments
Review of
Resident Comments

Prefectural
Governor(s)Inquiry
Comments
Governor’s
Comments

Mayors of theRelated
Cities, Towns
and Villages

Recommendation with
Governor’s Comments

Conduction ofEnvironmental Impact
Assessment
Submittal of Draft EIS
Preparation of
Draft EIS

Examination of
Draft EIS
(based on Governor(s)
and Residents
Opinion)(Advised by theAdvisory
Committee on
Environmental
ImpactSubmittal of
Draft EIS
Announcement
and Notification

Explanatory Meeting
Outline ofcomments(with
Operator’s
Position)

Submittal of
Resident’s
Comments
Review of
Resident Comments

Prefectural
Governor (s)InquirGovernor’s
OpinionInquirOpinioOpinionMayors of theRelated Cities,
Towns andVillages

Minister of theEnvironment

Recommendation with
Governor’s Opinion
Preparation of EIS

Submittalofaccepted EIS

Submittal of EIS
Notice of Acceptance(or Order to Change)

Examinationof EIS

Minister of theEnvironment

Submittal of EIS

Finalization of EIS

Prefectural
Governor(s)Notification
to Resident

Public Announcement

Mayors of
the RelatedCities,Towns andVillages

Proper measures forEnvironmental Preservation

Application or
Notification ofConstruction Plan

Application or
Notification

Examination ofConstruction Plan
(Conformance to EIS, etc.)

Approvalor Order to Change

NOTE：EIS: Environment Impact Statement, SD: Scoping Document

Fig. 17-1 Outline of the Environmental Impact Assessment on Nuclear Power Plant

17-6

Article 18 Design and Construction
Each Contracting Party shall take the appropriate steps to ensure that:the design and construction of a nuclear installation provides for several reliable levels
and methods of protection (defense in depth) against the release of radioactive materials,
with a view to preventing the occurrence of accidents and to mitigating their radiologicalconsequences should they occur;
the technologies incorporated in the design and construction of a nuclear installation areproven by experience or qualified by testing or analysis;
the design of a nuclear installation allows for reliable, stable and easily manageable
operation, with specific consideration of human factors and the man-machine interface.

The nuclear installations in Japan (light water reactors in commercial stage and fast breeder
reactors in demonstration stage) were designed, constructed and operated based on the safety
design concept, which is common to most Western countries, adopting fundamentally the same
defense in depth system as prescribed in the Nuclear Safety Standards "NUSS" of the IAEA.
Moreover, the knowledge obtained from operating experiences and various kinds of
examination, analyses, research and development are utilized, on a continuous basis, to realize
safer and easier facilities to carry out the maintenance management. Furthermore, this new
knowledge is reflected appropriately and successively on planning and revising of guides etc.
in order to improve the safety and reliability of nuclear reactors.
18.1

Review at the Design and Construction Stage of Nuclear Installations

Licensing process and relevant law and regulatory requirements applied at the design andconstruction stage for nuclear installation in Japan are described in the report of Article 7.
Regulatory Guide for Reviewing Safety Design of Light Water Nuclear Power Reactor
Facilities, (hereinafter called as the “Regulatory Guide for Reviewing Safety Design”),
Regulatory Guide for Reviewing Safety Assessment of Light Water Nuclear Power Reactor
Facilities (hereinafter referred to as the “Regulatory Guide for Reviewing Safety
Assessment”) and related guides are used for the review of the design concerning safety of a
nuclear installation.
The basic policies of the safety design of nuclear installations for electricity generation areprovided in the Regulatory Guide for Reviewing Safety Design. The Regulatory Guide
requires that each system, structure and component constituting nuclear installations achieve
the assigned functions under environmental and loading conditions during their in-service
period (not only in the normal operating conditions but also in abnormal conditions to be
postulated).
The Regulatory Guide for Reviewing Safety Assessment is used to confirm in the safetyassessment that the reactor facilities consisting of such systems, structures and components,
should be sufficiently safe ones as a whole.. The Regulatory Guides provides the postulated
events, criteria and items that should be taken into consideration.

18-1

When design alteration is required on licensed commercial nuclear installation, the licensee
must undergo the verification on the safety impacts due to the alteration as well as theinspection on the altered segment, including the safety analysis influenced by the design
alteration, in the same procedure as the licensing process as licensing a new installation.
18.2 Realization of Defense in Depth and Confinement of Radioactive Materials at the
Design and Construction Stage
Commercial nuclear installation (light water reactors (BWRs and PWRs)) in Japan are
designed, constructed and operated based on the safety design principals, which are common
among most Western countries and fundamentally the same concept of "defense in depth" asprescribed in the Nuclear Safety Standards (NUSS) of the IAEA. In this section the first 3
levels of defense in depth concepts, which are closely related to design and construction of
nuclear installations, are discussed. Forth and fifth levels of defense in depth concept, which
are severe accident management and emergency preparedness, are discussed in sections 18.6
and report on Article 16. Original design of light water reactors in Japan was introduced fromthe United States. But, the later design of reactors has been improved so that the facilities
have become safer and easier in maintenance management through a series of Improvement
and Standardization Program led by METI (then MITI), reflecting the operating experiencesof those who have obtained license for reactor establishment and knowledge obtained inresearch and development program of nuclear power industries.
(1) Implementation of the Defense in Depth Concept
The principle of "defense in depth" is as follows:yPrevention of deviating from normal operation conditions by means of conservative
design, manufacturing and construction of the nuclear plant in accordance with therelevant quality level and engineering practices.

y

Detection of the occurrence of an abnormal event at an early stage and taking
preventive measures against its progression into an accident. AndyControl of the progression of accident and mitigation of its consequences on theassumption that progression to an accident might not be prevented at the preceding
stage.

In order to apply these fundamentals to design of commercial nuclear installation, theRegulatory Guide for Reviewing Safety Design (see Table 18-1) that was established by the
NSC, stipulates the following items. The first defense is preventive measures for theoccurrence of an abnormal event. More specifically, as stated in the requirements in guidelines
1 to 10 (overall nuclear reactor facility) of the Regulatory Guide for Reviewing Safety Design,
the first defense implies such measures of designing with a safety margin, implementing strict
quality control in fabrication, inspecting the facilities and component to be fabricated as
required by the design and preventing degradation of performance through monitoring, checkand maintenance during the operation. Each component, equipment and system of nuclear

18-2

reactor facility is to be designed considering the importance of its safety function. The
Regulatory Guide for Classification of the Importance of Safety Functions for Light Water
Nuclear Power Reactor Facilities requires that the quality control during design andmanufacturing be conducted corresponding to the importance of safety function.
The second defense is to prevent expansion of abnormalities. More specifically, as stated in
the requirements in guidelines 15 to 18 (Reactor Shutdown System), and 34 to 40 (Safety
Protection System) of the Regulatory Guide for Reviewing Safety Design, the second defense
implies the early detection of the abnormal condition, its correction or taking measures inadvance to prevent the progression into an accident.
The third defense is to mitigate the consequence of an accident. More specifically, as stated in
the requirements in guides 25 (Emergency Core Cooling System) and 28 to 33 (Reactor
Containment) of the Regulatory Guide for Reviewing Safety Design, the third defense implies
taking measures to secure the safety of the public in the vicinity by controlling the progressionof the accident and mitigating its consequence.
The safety of nuclear facilities is ensured through rigorous safety measures on the basis of the
defense in depth concept, which includes 1) preventing the occurrence of an abnormal event,2) detection of the abnormal event and the preventing progression into an accident, and 3)
mitigating an accident consequence. Consequently in Japan, through these measures, it ispossible to reduce the potential for the occurrence of a severe accident to the extent that its
actual occurrence would be technologically inconceivable, and to maintain the risk of thenuclear installation at a sufficiently low level. Based on such a status, preparation of theaccident management can be regarded as a measure to reduce this low risk furthermore beyond
these protection levels. In addition, preparation of the accident management and theemergency measures, which has been carried on in Japan, are described in section 18.4, and in
Article 16, respectively.
(2) Confinement of Radioactive Materials (or Three Barriers of Radiation Protection Walls)
Nuclear facilities shall be designed, constructed and operated, in such a way as to confine
radioactive materials within a series of physical barriers. These physical barriers are the fuelpellet, the fuel cladding, the reactor coolant pressure boundary and the reactor containment.
The requirements for these physical barriers in the Regulatory Guide for Reviewing Safety
Design etc. and the outcome of the design improvements in them are as follows:
1) Fuel (Including Claddings)
The fuel assembly shall be so designed that a) the integrity will be retained under thevarious conditions that could occur in the nuclear reactor in service; b) the safety
protection system will actuate the reactor shutdown system, etc. so that the allowable
design limit of the fuel shall not be exceeded at an abnormal transient during operation; c)the reactor core cooling will not be impaired by a reactivity insertion accident and, more
specifically, the maximum fuel enthalpy by analysis will not exceed the specified value;
and the emergency core cooling system will be capable of preventing major damage to the18-3

fuel in a loss of coolant accident, and the fuel cladding metal water reaction will belimited to sufficiently small amount.
Regarding item a), the design requirements are stipulated in guidelines 11 and 12 of theRegulatory Guide for Reviewing Safety Design. Regarding item b), the design
requirements are stipulated in guidelines 34 to 40 (Safety Protection System). Regarding
item c), the design requirements are stipulated in guidelines 12, 14 and 25. The
requirements for safety assessment are also stipulated in the Regulatory Guide for
Evaluating Reactivity Insertion Events of Light Water Nuclear Power Reactor Facilities
and the Regulatory Guide for Evaluating Emergency Core Cooling System Performanceof Light Water Nuclear Power Reactors.
2) Reactor Coolant Pressure Boundary
The reactor coolant pressure boundary shall be so designed that the integrity will bemaintained during normal and abnormal operating conditions; that the boundary will not
exhibit brittle behavior or develop rapid brittle fracture during normal operation,
maintenance, testing, or abnormal conditions; that the leakage will be detected
immediately and surely; that tests and inspections will verify its integrity throughout theservice life of the nuclear reactor, which are required in guidelines 19 to 22 of theRegulatory Guide for Reviewing Safety Design. Pressure on reactor coolant pressure
boundary will not exceed the specified value during reactivity insertion events, which is
required in guideline 14 of the Regulatory Guide for Reviewing Safety Design.
3) Reactor Containment
The reactor containment shall be so designed that it will withstand the loads of designbasis accident and the appropriate design basis earthquake; that it will prevent leakage
exceeding the predetermined leakage rate; that it will allow periodic testing on theleakage rate; that its boundary will not exhibit brittle behavior or develop rapid brittle
fracture during normal operation, maintenance, testing and in abnormal conditions; andthat isolation valves should be placed in the pipelines that penetrate its walls, which arerequired in guidelines 28 and 29 of the Regulatory Guide for Reviewing Safety Design.
18.3 Systems for preventing the occurrence of abnormalities and for mitigating theimpact of abnormalities (Regulatory Guide for Reviewing Classification of Importance of
Safety Functions for Light Water Nuclear Power Reactor Facilities)
In Japan, the Regulatory Guide for Reviewing Classification of Importance of Safety
Functions for Light Water Nuclear Power Reactor Facilities (hereinafter called as the“Regulatory Guide for Classification of Importance") prescribes the system for preventing theoccurrence of abnormalities and the system for mitigating the impact of abnormalities. That is,since the Regulatory Guide for Reviewing Safety Design used at the safety review andassessment must be appropriately applied according to the safety importance of the subject
structures, systems and components, safety functions and the classification of importance ofthe structures, systems and components are defined in this "Regulatory Guide for Reviewing

18-4

Classification of Importance".
(1) The Concept of the Classification of Importance for the Safety Design
The importance of safety functions of the structures, systems and components are classified
into the following two classes and shown in the Regulatory Guide for Reviewing
Classification of Importance.

1) Those of which loss of the function could result to cause an abnormality of the nuclear
reactor facility, which causes excessive radiation exposure on general public or the working
personnel (the system for preventing the occurrence of abnormalities, hereinafter called as“PS”).
2) Those that have the function to prevent the propagation of abnormality or terminate it
quickly in an abnormal situation of a nuclear reactor facility, and to protect general public or
the working personnel from possible excessive radiation exposure (the system for mitigating
the impact of abnormalities, hereinafter called as “MS”). The structures, systems andcomponents, which belong to these PS and MS respectively, are classified into three classes inaccordance with the importance of their safety function. It stipulates, from the standpoint toensure the safety function, that the basic objective for each class shall meet the following
requirements according to the technologies of design, construction and tests, and operation
management.
Class 1: Secure and maintain as high as reasonably achievable level of reliability.
Class 2: Secure and maintain a high level of reliability.
Class 3: Secure and maintain a level of reliability equal to or higher than that for generalindustry.
Moreover, it is required that functional isolation and physical separation among two or more
systems, structures and components, which have safety functions, are taken into consideration
appropriately. When connecting systems, structures and components having different degreeof importance, it is required that the design requirements equivalent to those on systems,
structures, and components with higher degree of importance should be applied to the othersystems, structures and components with lower importance or that the appropriate functional
isolation should be taken into consideration.
The classifications of the structures, systems and components and their safety functions arelisted in Table 18-2.
(2) Installation of PS and MS
The PS and MS installed in the light water reactors in Japan are as follows. After grouping alllight water reactors, currently installed in Japan, based on the reactor type and thecontainment type, the essential system for PS and MS, which are installed in each nuclear

18-5

reactor facilities are shown in Table 18-3 and Table 18-4 for BWR and PWR respectively.
These tables summarize the system configuration and their classification of reactor shutdown
system, emergency core cooling system and heat removal system, the number of diesel
generators and the containment shape, as essential systems for preventing the occurrence ofabnormalities and system for mitigating the impact of abnormalities.
18.4

Safety Design Assessment

In the safety design assessment, postulated event groups are defined for "abnormal transientsduring operation" and "accidents", respectively, based on the Regulatory Guide for Reviewing
Safety Assessment, as mentioned later, then the safety is evaluated by conducting safety
analysis. These event groups conform to the classification defined in the Nuclear Safety
Standards (NUSS) of IAEA almost.
The person who intends to install a nuclear reactor conducts the safety analysis for these
postulated event groups, compares the analysis results with each criterion, and confirms thatthe safety design is appropriate.
On the other hand, the Nuclear and Industrial Safety Agency examines the safety analysis of
the person who intends to install a nuclear reactor, and confirms its validity, getting an
independent analysis report performed by the incorporated administrative agency, Japan
Nuclear Energy Safety Organization, if necessary. The postulated events for the safetyassessment are selected and evaluated in the following manners;
Malfunctions and erroneous actions of the systems or components, which are applied in thebasic design, are analyzed, and the event which results in the severest case is selected among
similar events in the propagation process as the postulated event group for the safetyassessment. Depending on the possibility of occurrence and the degree of its impact at thetime of occurrence, these postulated events are classified into "abnormal transients during
operation" or "accident" as provided in the Regulatory Guide for Reviewing Safety
Assessment, and the safety of those postulated events are also evaluated based on the criteria
defined to each classification.
a. "Abnormal transients during operation""Abnormal transients during operation" are defined as events that result in abnormal
conditions caused by an single equipment failure, erroneous action or single disoperation
assumable in the lifetime of the nuclear installation, and the external disturbance assumableto occur with similar frequency of the single equipment failure, etc. during the operation ofcommercial power reactors, and fourteen events and twelve events are selected forpressurized water reactors (PWR) and boiling water reactors (BWR), respectively. The
safety analysis is conducted for these events based on the criteria of the Regulatory Guide
for Reviewing Safety Assessment, the integrity of core and reactor coolant pressure
boundary is confirmed, and the adequacy of the safety design of important safety related
equipment, such as the safety protection system and the reactor shut-down system islogically clarified.

18-6

b. "Accident"
The "accident" is an abnormal condition exceeding the "abnormal transients during
operation", which is assumed from a necessity to evaluate a release of radioactive materials
from a commercial power reactor when it occurs, although the frequency of occurrence issmall. Ten events and nine events for PWR and BWR, respectively, are selected. The safety
analysis is conducted for these events based on the criteria of the Regulatory Guide for
Reviewing Safety Assessment, and it is confirmed that the core does not result in asignificant damage and a reactor containment boundary is sound. Moreover, no risk ofexcess radiation exposure to the general public in the vicinity is confirmed. And it islogically confirmed that the safety design of engineered safety features is appropriate.
In addition, the analysis of the "accidents" event are verified and evaluated for the loss of
coolant accidents based on the Regulatory Guide for Reviewing Safety Assessment of Light
Water Nuclear Power Reactor Facilities and the "Regulatory Guide for Evaluating
Emergency Core Cooling System Performance of Light Water Nuclear Power Reactor" andfor the reactivity insertion events based on the Regulatory Guide for Reviewing Safety
Assessment and the "Evaluation Guide for Reactivity Insertion Events of Light Water
Nuclear Power Reactor Facility", etc.
18.5 New Regulatory Guide for Reviewing Seismic Design of Nuclear Power Reactor
Facilities“The Regulatory Guide for Reviewing Seismic Design of Nuclear Power Reactor Facilities”,
which is a specific seismic guide related to the general reviewing guide “the Regulatory
Guide for Reviewing Safety Design”, was revised in September 2006. The old Regulatory
Guide for Reviewing Seismic Design was issued in 1978 by the Atomic Energy Commission.
The provisions about static seismic force was revised in 1981 by the Nuclear Safety
Commission, and the terminology was revised based on the recommendations of theInternational Committee on Radiation Protection (ICRP) in 2001. Since then, lots of new
knowledge about seismology and earthquake engineering has been obtained and the design
and technology concerning seismic safety have been remarkably improved, and in particular,
after the South Hyogo Earthquake which occurred in 1995, knowledge and information havebeen obtained through the researches and studies which have been carried out relating to thatearthquake.
In order to make the Regulatory Guide for Reviewing Seismic Design etc. more appropriate,taking in the latest knowledge etc., the discussions and reviews were promoted by the Nuclear
Safety Commission in 2001. After open discussions and reviews and hearing of opinion fromthe public, the new Regulatory Guide for Reviewing Seismic Design was issued in September
2006.
Summary of the new Regulatory Guide for Reviewing Seismic Design is as follow;
1)

Advanced methods to evaluate /to determine design basis earthquake ground motions(geological survey etc.)

18-7

(i) Extension of geologic age of active-fault evaluation
The geologic age investigated for surveying the active-faults (traces of pastearthquakes), which should be taken into consideration on the seismic design, hasbeen traced back to the late Pleistocene age, if activities cannot be denied. Formerly
it was the age until 50,000 years ago.

(ii)

Careful active-fault survey
The active-fault survey required when deciding the design basis earthquake groundmotion to be used in a seismic design should be carried out more in detail andcarefully, depending on the distance from the site, integrating various methods oftectonic geomorphologic examination, surface-of-the-earth geological survey andgeophysical survey, so that all possible measures can be applied in the evaluation ofactive faults that should be taken into consideration on the seismic design

2) Method to evaluate / to determine earthquake ground motions (Determination of thedesign basis earthquake ground motion)(i)

Unification of design basis earthquake ground motions
The design basis earthquake ground motion is changed from conventional set of two
types (the design basis earthquake ground motion S 1 based on the design basis maximum
earthquake and the design basis earthquake ground motion S2 based on a design basis
extreme earthquake and/or a near field earthquake) to set of one type (S S ) and ofwhich setting conditions are more strict than that of S 2 . The basic requirements for
S S are that the safety function of facilities important to seismic safety should be kept.

(ii) Advanced evaluation method of the earthquake ground motion, which is determinedby identifying the source for each site
In addition to the old experiential evaluation method (the method using a response
spectrum), the "fault model", which is the newest evaluation method, was introduced
newly in full scope, so that the evaluation method of earthquake ground motions areenhanced by taking advantages of both methods.
(iii) Introduction of evaluation method of an earthquake ground motion, which isdetermined without identifying its source
In the case of an earthquake in the inland earth’s crust, it sometimes occurs difficult
to identify the relationship between the source and the active fault. Therefore, it was
decided to determine an earthquake ground motion by setting up a response
spectrum based on the observation records of such earthquakes. This has realized
the preparation for the evaluation on an earthquake, for which the evaluation would
not be possible even if a careful active-fault survey is carried out. As a result, the18-8

provision of the near field earthquake with magnitude 6.5 i n t h e f o r m e r g u i d e w a sabolished.
(iv) Individual evaluation on the vertical earthquake ground motion
In the new Regulatory Guide for Reviewing Seismic Design, it was decided todetermine the vertical dynamic earthquake ground motion also with the horizontal
one in determining S S earthquake ground motion..
(v) Consideration on "a residual risk"
As it is impossible to completely deny the occurrence of an earthquake with theground motion exceeding S S , a "residual risk" was decided to be taken into
consideration. Furthermore, it is required to take into consideration factors and themagnitudes of the "uncertainties" to the size, position, propagation etc. with anappropriate method, and to refer to the probability of exceeding S S earthquake as
the reference information at the safety review and assessment of the plant.
3) Reexamination on the importance classification with regards to seismic safety(i)Expansion of the scope of the facilities important for the safety
For the seismic safety design, the scope of the most important facilities, such as thereactor containment (previous As class), was expanded to include the emergency
core cooling system etc. (previous A class)(ii) Requirement to take into consideration the accompanying events of earthquake
It was described clearly to take consideration of the accompanying events ofearthquake (collapse of the inclined planes around facilities, tsunami etc.).
(iii) Improvement of requirements for a rock-bed support
In consideration of the progress of seismic isolating technology etc. ,the "rock-bed
support" requirement for the building and structure has been changed to a
performance based requirement prescribing, "construct on a soil with adequate
support performance".
4)

Effort to use the probabilistic-safety-assessment methodologies(i)

It was decided that all licensees should make efforts to make the "residual risk" as
low as reasonably achievable, and that the effort towards extensive introduction of
the probabilistic safety assessment methods should be made in the future.
The above-mentioned new Regulatory Guide for Reviewing Seismic Design isapplied to the nuclear installations on which the application of the establishment
approval will be made newly from now on, and for existing nuclear installations, it

18-9

is requested that all licensees evaluate the seismic safety based on the revised
contents. Actions to the existing nuclear installations are provided in Section 14.5.
18.6

Preparation of Accident Management Measures

Since the TMI-2 accident, the researches on phenomena of severe accidents and PSA havebeen conducted extensively worldwide. The NSC decided “Accident Management of Severe
Accidents at Power Generating Light Water Reactor Facilities” in 1992, and revised it in 1997.
Licensees in Japan also have voluntarily implemented their own measures for preventing
severe accidents and for mitigating their consequences at the request of the MITI (then) basedon the NSC’s decision. Typical facility modifications for the accident management to prevent
an occurrence and to mitigate the consequence of a severe-accident are as shown in thefollowing;
PWR: Alternative recirculation (installation of alternative sump-pumps, or core flooding
using the containment spray system by installing the tie-line between the containment spraysystem and the residual heat removal system), containment natural convection cooling(utilization of the common containment cooling system), alternative component cooling(utilization of the HVAC chilled-water system etc.), water injection into a reactor containment(utilization of the fire protection system), common usage of power supply among units (usageof power from the neighboring nuclear installation), hydrogen-concentration control (only forice condenser type PWR)
BWR: Alternative reactivity control (recirculation pump trip and automatic alternative control
rod insertion), alternative cooling water injection(utilization of the fire protection system),automatic reactor depressurization(automatic depressurization at the low water level of RPV ),heat-removal from a reactor containment( pressure venting for preventing vessel rupture andutilization of drywell cooler ), power supply system(common usage of power supply among theneighboring nuclear installation)
For implementing accident management at operating commercial nuclear installations, thelicensees have been developing the accident management measures progressively,
substantiating the facilities as mentioned above during the outage of the periodic inspection as
well as establishing operational measures such as implementing system, procedures, educationof personnel, etc.
The accident management measures that were prepared by licensees were reported to NISA in
May, 2002, together with the PSA results of internal events for representative reactor types forthe purpose of quantitatively verifying the effectiveness of enhancement of the safety. Whilelicensees were developing accident management measures, NISA requested NUPEC (then) toevaluate the effectiveness of the accident management measures, and established the"Accident Management Workgroup" under the Nuclear and Industrial Safety Subcommittee toobtain the opinion of specialists, and evaluation report was compiled and issued in October
2002. The report was submitted to the NSC by NISA in the same month. The PSA results of
the internal events for all commercial power reactor facilities under operation (excluding

18-10

representative reactor types) were reported to NISA by the licensees in March 2004.
The development programs of the accident management for commercial nuclear installation
under construction (three units) were reported to NISA by the licensees in July 2003, and theevaluation results etc. were reported to the NSC by NISA in September 2003. The NSC
evaluated the report and concluded it was reasonable in December 2003. In addition, theaccident management measures for the reactor facilities concerned are being prepared by thelicensees.
18.7

Measures to Ensure the Technical Reliability by Experience, Test and Analysis

In Japan, such actions as feedback of the operating experience and utilization of the technicalknowledge obtained through testing and analysis have been taken, so that the safety andreliability of commercial nuclear installation has been enhanced .The brake-downs aredescribed below. The new knowledge obtained through these actions has been timely
incorporated in existing guidelines and used to develop new guidelines.
(1) Feedback of Operating Experiences from Commercial Nuclear InstallationsyGood practices and non-compliance examples identified during periodic inspections,as well as the experiences in design, construction and operation of domestic andforeign commercial nuclear installation, were analyzed, and the results areincorporated in design modification, improvement of construction methods, etc.,
when they can be recognized to be effective, during the course of licensing forestablishment, approval of construction plans and pre-service inspection.

y

For accidents or failures occurred in the domestic commercial nuclear installations aswell as in foreign reactors, the corrective measures are implemented after identifying
the cause of failures.

yFrom the standpoint of the comprehensive preventive maintenance of nuclear reactor
facilities, periodic safety review is performed for each commercial power reactor with
the interval of approximately ten years. And its safety and reliability are confirmed,
reflecting the results of comprehensive evaluation on operating experiences and thelatest technical knowledge. The situation of periodic safety review is described in
Article 14.

ySince the Three Mile Island accident, the habitability of a central control room has
been reexamined. Meanwhile in Japan, there was an event in which the steam invaded
the central control room at the time of the pipe break accident due to the wall
thinning of the second system piping of the Mihama Unit No. 3 in 2004. It was foundthat the air-tightness of the central control room was insufficient. Although filling
was provided as a temporary measure to keep the air-tightness, a fundamental
measure has to be taken. On the occasion when the technical standard was amended
to be based on performance requirements, the air-tightness requirement was addedon the central control room, and it will be a matter to be evaluated at the safety

18-11

examination of a new nuclear installation. And the tests of leak-tightness of thecentral control room have been conducted on 3 BWRs and one PWR so far and there
are the plans to conduct leak tightness test on some additional control rooms further.
Based on the results of these testing, the private sectors’ standards, relating to leaktightness testing will be developed..
yAs many fire events have occurred repeatedly in and outside Japan and as the past
OSART review h a s made recommendations/suggestions. About the fire-protection
management, the standards and the guidelines on design and management for fire
protection of nuclear installations have been re-examined and improved in Japan. On
an occasion when the technical standards was amended to more performance-based
one, the requirements of the fire protection in the technical standard were
reexamined, and the requirements were clarified for each stage of "fire prevention","early stage detection and extinguishing of fire" and "fire consequence mitigation".
After these activities, the fire-protection standards for the design and development ofstandards of academic societies and industrial associations for operation management
have been re-examined. Furthermore, safety researches started, including the study
for development of a fire PSA method and the participation to the international
project for various fire experiments..

y

The electric cables currently used in the nuclear installation, while the aging
advances gradually by oxidation etc., sudden performance degradation may occur inthe environment of high temperature steam and high radiation at the time of a
postulated design base accident. The studies are carried out to evaluate such aging
and performance degradation and to confirm the integrity of cables during in-service
operation. The studies to obtain the heat deterioration data and the heat andradiation deterioration data, etc. have been conducted using test samples of thesafety-related cable currently used in the nuclear installation, aiming atcomprehensive assessment of the cable aging characteristics as well as at correct
assumption of the environmental design conditions and at establishment of integrity
judging methods, adding the new knowledge obtained in recent years to the study
results. Thereby the evaluation methods and evaluation test guideline of aged cables
will be established on the basis of the actual conditions in a nuclear installation.

(2) Feedback of the Knowledge Obtained Through Test and Analysis
Recognizing the importance of assuring safety in development and utilization of nuclear
energy, the researches for advancement of safety standards, guidelines, reference materials for
acceptance decision in safety review and assessment etc., as well as improvement of the safety
itself, are promoted in Japan. The explanation of promoting safety research in Japan isprovided in Section 14.8.
The major subjects related to the research are shown in the following;
1) Study corresponding to advanced light water reactor fuels

18-12

y
y
y
y

Study on the safety of high-burn-up MOX fuelConfirmatory test on the safety margin of high burn-up fuels
Reliability demonstration test of 9X9 type fuelReliability demonstration test of nuclear design methodology for the full MOX
core2) Study on advanced safety assessment technologiesyResearch of advanced nuclear and thermal-hydraulic best estimate methodyImprovement of safety analysis codes for nuclear power reactors
3) Study on severe accidentyStudy on the water hammer by rapid void growth at the time of severe accidentyStudy on maintenance of the containment confinement function at the late stage ofa severe accident
4) Seismic safety study of nuclear installationsyStudy on the design earthquake ground motion with a consideration of thecharacteristics of the postulated earthquakey
Test on the seismic-assessment technologies of nuclear installationsyStudy on the hazard map for seismic design
18.8

Consideration of Human Factors and Man-Machine Interface

It is the safety requirements regarding operating management to make nuclear installation
more reliable, more stable and more easily manageable taking into human factors andman-machine interface. These requirements are implemented in design and operation of thecommercial power reactors in Japan.
Considerations of the design to an operator behavior, requirements on the design of a control
room and concrete design to approach these requirements are described in the report of Article
12.

18-13

Table 18-1 Individual guides established in the NSC Regulatory Guide
for Reviewing Safety Design of Light Water Nuclear Power Reactor Facilities (1/2)(General requirements for nuclear reactor facilities)
Guideline 1. Applied codes and standards
Guideline 2. Design considerations against natural phenomena
Guideline 3. Design considerations against external human initiated events
Guideline 4. Design considerations against internal missile
Guideline 5. Design considerations against fire
Guideline 6. Design considerations against environmental conditions
Guideline 7. Design considerations for share use
Guideline 8. Design considerations against operator actions
Guideline 9. Design considerations for reliability
Guideline 10. Design considerations for testability(Nuclear reactor and reactor shutdown system)
Guideline 11. Reactor Core design
Guideline 12. Fuel design
Guideline 13. Reactor characteristics
Guideline 14. Reactivity control system
Guideline 15. Independence and testability of reactor shutdown system
Guideline 16. Reactor shutdown margin by control rods
Guideline 17. Shutdown capability of reactor shutdown system
Guideline 18. Reactor shutdown system capability at the accident(Reactor cooling system)
Guideline 19. Integrity of reactor coolant pressure boundary
Guideline 20. Prevention of reactor coolant pressure boundary failure
Guideline 21. Detection of the reactor coolant pressure boundary leaks
Guideline 22. In-service test and inspection of reactor coolant pressure boundary
Guideline 23. Reactor coolant make-up system
Guideline 24. Systems for removing residual heat
Guideline 25. Emergency core cooling system
Guideline 26. System for transporting heat to ultimate heat sink
Guideline 27. Design considerations against loss of power(Reactor containment)
Guideline 28. Function of reactor containment
Guideline 29. Prevention of reactor containment boundary failure
Guideline 30. Isolation function of reactor containment
Guideline 31. Reactor containment isolation valves
Guideline 32. Reactor containment heat removal system
Guideline 33. System for controlling containment facility atmosphere(Safety protection system)
Guideline 34. Redundancy of safety protection system
Guideline 35. Independence of safety protection system
Guideline 36. Function of safety protection system during transients
Guideline 37. Function of safety protection system in case of the accident
Guideline 38. Function of safety protection system in case of failure
Guideline 39. Separation of safety protection system from instrumentation and control systems
Guideline 40. Testability of safety protection system

18-14

Table 18-1 Individual Guidelines established in the NSC Regulatory Guide
for Reviewing Safety Design of Light Water Nuclear Power Reactor Facilities (2/2 )(Control room and emergency facilities)
Guideline 41. Control room
Guideline 42. Reactor shutdown function from outside of control room
Guideline 43. Design considerations for control room habitability
Guideline 44. On-site emergency station
Guideline 45. Design considerations for communication equipment
Guideline 46. Design considerations for evacuation routes(Instrumentation and control system and electrical system)
Guideline 47. Instrumentation and control system
Guideline 48. Electrical system(Fuel handling system)
Guideline 49. Fuel storage and handling system
Guideline 50. Prevention of fuel criticality
Guideline 51. Monitoring of fuel handling area
(Radioactive waste processing facility)
Guideline 52. Radioactive gaseous waste processing facility
Guideline 53. Radioactive liquid waste processing facility
Guideline 54. Radioactive solid waste processing facility
Guideline 55. Radioactive solid waste storage facility(Radiation management)
Guideline 56. Environmental radiation protection
Guideline 57. Radiation protection for personnel engaged in radiation work
Guideline 58. Radiation management for personnel engaged in radiation work
Guideline 59. Radiation monitoring

18-15

Table 18-2 Definitions and functions with respect to classifications ofimportance of safety function (1/2)
Classification
Class 1
PS-1

MS-1

Definition
Function(i) Reactor coolant pressure boundary
Structures, systems andfunctioncomponents where there is concern(ii) Excessive reactivity insertion
that(a) a conspicuous damage to the core,
prevention function
or(iii)Core shape maintenance(b) significant damaging the corefunctionmay occur, due to an event causedby such damage or malfunction.
(i) Reactor emergency shutdown(i) Structures, systems andfunctioncomponents that implement an(ii) Sub-criticality maintenance function
emergency shutdown of the(iii)
Function to prevent
nuclear reactor, remove theover-pressurization of reactor
residual heat, prevent excess
coolant pressure boundarypressure in the reactor coolant
pressure boundary and prevent (iv) Cooling function after reactor
shutdown
an impact of excessive radiationon the public in the site vicinity, (v) Core cooling function
at the occurrence of an abnormal (vi) Radioactive material confinement
function, shielding of radiation andcondition.
release reduction functions(ii) Other essential safety related
structures, systems andcomponents

Class 2PS-2(i) Generation function of an
actuation signal for the engineered
safety features and to the reactor
shutdown system(ii) Specially important safety related
functions(i) Function the builds in reactor(i) Structures, systems andcoolant (However, this excludes
components for which there issmall diameter piping, such as
no concern of the immediate
instrumentation, etc., excluded from
causing of conspicuous reactor
the reactor coolant pressuredamage or significant fuelboundary and those that are notdamage due to an event thatconnected directly to the boundary.)
occurs due to such a damage or(ii) Components not directly connectedmalfunction, however, forto the reactor coolant pressure
which there is a concern ofboundary, which have theexcessive release of radioactive
radioactive materials storage
materials outside the site
functionvicinity.
(iii) Function for the safe handling offuel(i) Safety valve and relief valve(ii) Structures, systems andre-closing function
components which must be
actuated during normal
operation and upon an abnormaltransient during operation forwhich there is a high potentialthat core cooling will be lost
due to the concerned
malfunction.
Reference: “Regulatory Guide for Reviewing Classification of Importance of Safety Functions forLight Water Nuclear power Reactor Facilities”, decided by the NSC in August 30, 1990 .

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Table 18-2 Definitions and functions with respect to classifications ofimportance of safety function (2/2)
Classification
MS-2

Class 3PS-3Definition

Function(i) Structures, systems and(i) Fuel pool water supply function(ii) Function to prevent the discharge ofcomponents for adequately
radioactive materials
reducing the impact of radiationon the general public in thevicinity of the site, due todamage or malfunction in the
structures, systems andcomponents of the PS-2.
(i) Function for determining the(ii) Structures, systems andsituation of the plant at the time ofcomponent with an especiallyan accident
important function in the
response of abnormal situations. (ii) Function for mitigation of abnormal
situations(iii)
Function for safe shutdown
from outside the control room(i) Reactor coolant preserving function(i) Structures, systems and(components other than PS-1 andcomponents where initiating
PS-2)
events of abnormal situations
take place, and which are other (ii) Reactor coolant circulation function
Radioactive material storage
than PS-1 and PS-2 components. (iii)function(iv) Power supply (excluding
emergencies)(v) Plant instrumentation and control
function (excluding safety
protection function)(vi) Plant operation supporting functions(ii) Structures, systems and(i) Function for preventing thecomponents which hold the
diffusion of fission products into theconcentration of the radioactive
reactor coolantmaterials in the reactor coolant (ii) Reactor coolant purification
function
to a level low enough not to
impair normal operation

MS-3(i) Structures, systems andcomponents which mitigate
events in conjunction with theMS-1 and MS-2, even when
there is an abnormality during
operation(ii) Structures, systems andcomponents required for the
response of abnormal situations

18-17(i) Function for mitigation of reactor
pressure increase(ii) Function to control the powerincreases(iii)
Reactor coolant make-up
function(i) Important for emergency response
and function for recognizing
abnormal situations

18-18

Table 18-3 Establishment situation of prevention and mitigation system (BWR nuclear installation) (1/2)
Plant type
BWR 2 & 3
BWR 4
BWR 5
ABWR
Containment type
MARK-I
MARK-I
Improved MK-I, MK-II and Improved MK-II
RCCVUnit 6 of Kashiwazaki Kariwa NPS
Name of power Unit 1 of Tsuruga PS (BWR 2) Unit 1 of Onagawa NPS
Unit 1 of Shika NPS (Improved MK-I)
Unit 7 of Kashiwazaki Kariwa NPS
Unit 1 of Shimane NPS
Unit 2 of Shimane NPS (-ditto-)
Unit. 1 of Fukushima Daiichi,
station
Unit 2 of Shika NPS
Unit 2 of Fukushima Daiichi NPS Unit 2 of Onagawa NPS (-ditto-)(BWR 3)
Unit 5 of Hamaoka NPS
Unit 3 of Fukushima Daiichi NPS Unit 3 of Onagawa NPS (-ditto-)
Unit 4 of Fukushima Daiichi NPS Unit 3 of Hamaoka NPS (-ditto-)
Unit 5 of Fukushima Daiichi NPS Unit 4 of Hamaoka NPS (-ditto-)
Unit 2 of Hamaoka NPS
Tokai No.2 (MK-II)
Unit 6 of Fukushima Daiichi NPS, (-ditto-)
Unit 1 of Fukushima Daini NPS (-ditto-)
Unit 1 of Kashiwazaki Kariwa NPS (-ditto-)
Unit 2 of Fukushima Daini NPS (Improved MK-II)
Unit 3 of Fukushima Daini NPS (-ditto-)
Unit 4 of Fukushima Daini NPS (-ditto-)
Unit 2 of Kashiwazaki Kariwa NPS (-ditto-)
Unit 3 of Kashiwazaki Kariwa NPS (-ditto-)
Unit 4 of Kashiwazaki Kariwa NPS (-ditto-)
Unit 5 of Kashiwazaki Kariwa NPS (-ditto-),OthersReactorSCRAM system
SCRAM system
SCRAM system
SCRAM system
Stand-by Liquid Control System
Stand-by Liquid Control System
shutdown system Stand-by Liquid Control System Stand-by Liquid Control System
Containment shapeTable 18-3 Establishment situation of prevention and mitigation system (BWR nuclear installation) (2/2)
Plant type

BWR２＆３

BWR４

BWR５

ABWR

Containment type

MARK-I type

MARK-I type

Improved MK-I, MK-II and
Improved MK-I

RCCV type

Systemconfiguration of
ECCS and heatremoval system

HPCI
HPCS

HPCF
LPFL/RHR

IC

IC

CS

CS

CS

CS

18-19

CCS

CCS

CCS

CCS

SHC

RCIC
CS

HPCI
CS

LPCI

LPCI

LPCI

LPCI

RHR

RHR

RCIC
LPCS
LPCI/RHR

LPCI
LPCI/RHR

RCIC
LPFL/RHR

HPCF
LPFL/RHR

SHC
ADS

ADS

ADS

ADS

3 partitions

3 partitions

Divisions ofsystemconfiguration

2 partitions

2 partitions

Number of D/G2

2

IC: Isolation Condenser.
CS: Core Spray Sys.
CCS: Containment Cooling Sys.
RHR: Residual Heat Removal Sys.

SHC: Shutdown Cooling Sys.
HPCI: High Pressure Core Injection Sys.
LPCI: Low Pressure Coolant Injection Sys.
HPCF: High Pressure Core Flooder

3

3

RCIC: Reactor Core Isolation Cooling Sys.
ADS: Automatic Depressurization Sys.
HPCS: High Pressure Core Spray Sys.
LPFL: Low Pressure Core Flooder

Table 18-4 Establishment situation of prevention and mitigation systems (PWR nuclear installation) (1/2)

18-20

Plant Type

4 LoopContainment type

PCCV type

Name ofpower station

Unit 3 of Ohi PS
Unit 4 of Ohi PS
Unit 2 of Tsuruga PS
Unit 3 of Genkai NPS
Unit 4 of Genkai NPS

Dry type
Unit 1 of Ikata PS
Unit 2 of Ikata PS
Unit 1 of Mihama PS
Unit 2 of Mihama PS
Unit 1 of Genkai NPS
Unit 2 of Genkai NPS
Unit 1 of Tomari PS
Unit 2 of Tomari PS

Reactor shutdown Scram system
Boric acid injection systemsystemScram system
Boric acid injection system

Containmentshape

PCCV type2 Loop3 Loop4 LoopDry type

Ice condenser type

Unit 1 of Takahama PS
Unit 2 of Takahama PS
Unit 3 of Takahama PS
Unit 4 of Takahama PS
Unit 3 of Mihama PS
Unit 1 of Sendai NPS
Unit 2 of Sendai NPS
Unit 3 of Ikata PS
Scram system
Boric acid injection system

Dry type

Freestanding steel type (no top dome)

Unit 1 of Ohi PS
Unit 2 of Ohi PS

Scram system
Boric acid injection system
Ice condenser typeFreestanding steel type (with top dome)

Table 18-4 Establishment situation of prevention and mitigation systems (PWR nuclear installation) (2/2)
Plant Type

4 Loop2 Loop3 Loop4 LoopContainment type
Systemconfiguration of
ECCS and heatremoval system

PCCV type

Dry type

Dry type

Ice condenser type

ACC 4 units

ACC 2 units

ACC 3 units

ACC 4 units

HPIS (/CHP)

18-21

HPIS
LPIS
(/RHR)
AFWS

HPIS
LPIS
(/RHR)
AFWS

HPIS
LPIS
(/RHR)
AFWS

HPIS
LPIS
(/RHR)
AFWS(motordriven)(motordriven)(motordriven)(motordriven)AFWS(turbine driven)

Divisions ofsystemconfiguration

2 systems
HPIS boosting unnecessary

Number of D/Gs

HPIS(/CHP)LPIS(/RHR)
AFWS(motordriven)HPIS(/CHP)LPIS(/RHR)
AFWS(motordriven)HPIS
HPIS
C/CHP)LPIS(/RHR)
AFWS(motordriven)HPIS
HPIS
C/CHP)LPIS(/RHR)
AFWS(motordriven)AFWS(turbine driven)

AFWS(turbine driven)

AFWS(turbine driven)

2 systems

2 systems

2 systems

HPIS boosting necessary

HPIS boosting necessary

HPIS boosting necessary2222
ACC: Accumulator, AFWS: Auxiliary Feed Water Sys, LPIS: Low Pressure Coolant Injection, PIS: High Pressure Injection Sys.
RHR: Residual Heat Removal Sys., CHP: Charging Pump

This page intentionally left blankArticle 19 Operation
Each Contracting Party shall take the appropriate steps to ensure that:(i)
the initial authorization to operate a nuclear installation is based upon an appropriate
safety analysis and a commissioning programme demonstrating that the installation, as
constructed, is consistent with design and safety requirements;(ii) operational limits and conditions derived from the safety analysis, tests and operational
experience are defined and revised as necessary for identifying safe boundaries for
operation;(iii) operation, maintenance, inspection and testing of a nuclear installation are conducted inaccordance with approved procedures;(iv) procedures are established for responding to anticipated operational occurrences and to
accidents;(v) necessary engineering and technical support in all safety-related fields is available
throughout the lifetime of a nuclear installation;(vi) incidents significant to safety are reported in a timely manner by the holder of therelevant license to the regulatory body;(vii) programmes to collect and analyse operating experience are established, the results
obtained and the conclusions drawn are acted upon and that existing mechanisms areused to share important experience with international bodies and with other operatingorganizations and regulatory bodies;(viii) the generation of radioactive waste resulting from the operation of a nuclear installation is
kept to the minimum practicable for the process concerned, both in activity and involume, and any necessary treatment and storage of spent fuel and waste directly related
to the operation and on the same site as that of the nuclear installation take into
consideration conditioning and disposal.
Licensees are allowed to commence operation after the licensing conditions specified by the Reactor
Regulation Law etc. are complied with in the stage from licensing for establishment to theconstruction of commercial nuclear installations.
As legal regulations for licensees to operate reactors safely, it is obligated to provide the Operational
Safety Programs approved by the Minister of METI before commencement, and to observe necessary
measures for operation and maintenance of the commercial nuclear installation like the Periodic
Inspection etc., and all through its operating life.
By amendment of the Electricity Utilities Industry Law in 2003, the Periodic Inspection and the
Periodic Licensee’s Check were clearly established, and the scheme of the Periodic Safety
Management Review was newly established. Moreover, quality assurance activities, maintenancemanagement activities, the periodic safety review, etc. were decided to be included into the
Operational Safety Program. The Task Force on the Inspection System including the use of risk
assessments and performance-based evaluations is underway aiming to start at 2008.

19-1

19.1 Initial License
In Japan Licensees are required by the Reactor Regulation Law to take necessary measures for the safe
operation of nuclear installations and protection of specific nuclear fuel materials.
(1) “Reactor Establishment”
Throughout each stage of detailed design, construction and operation of a nuclear installation, thebasic design (items defined in the main text of application documents for Reactor Establishment)
approved by the Minister of METI must be followed(2) “Construction Plan approval”
Licensees are also required to observe the conditions of Reactor Establishment within the Construction
Plan approval (or Design Approval for fuel assembly), in which the detailed design for each facility ofa nuclear installation is reviewed.
(3) “Pre-service Inspection”
Before the commissioning of a nuclear installation, NISA conducts the pre-service inspection to ensurecompliance with the licensing conditions.
19.2 Limiting Conditions for Operation
Operation and maintenance of nuclear installations are carried out in accordance with the Operational
Safety Program approved by the Minister of METI in Japan (items included in the Operational Safety
Program are shown in “Operational Safety Program” in Section 19-3).
Limiting Conditions for Operation (hereinafter referred to as LCO) of a nuclear installation include
shutdown margin, reactor thermal limits, etc. Table 19-1 shows the details.
If the LCO is not complied with, the Minister of METI can order the licensee to suspend the operationof the nuclear installation, etc. in accordance with the Reactor Regulation Law.
Since December 2001, the rated thermal power operation that limits only thermal power was allowed.
Power stations shown in Table 19-3 have been subject to this operating mode to date.
19.3 Regulations for Operation, Maintenance, Inspection and Testing
The overviews of (1) present regulatory framework (2) inspection systems and (3) the on-going Task
Force on the Inspection System in line with the Reactor Regulation Law are as follows:(1) Present Regulatory Framework
Elements that consist of the regulation in Japan after commissioning are as follows;•“Application for alteration of the Reactor Establishment and application of Construction Planapproval”

19-2

In the case of reconstruction or repair after commissioning, an application for alteration of theReactor Establishment or an application for Construction Plan approval may be needed,“Operational Safety Program”•

Licensees are obliged to describe the following items related to the operation in the Operational Safety
Program in accordance with the rules of the Reactor Regulation Law,
Duties of personnel engaged in the operation and management of the nuclear reactor facility,1)

and the organization2)

Items with respect to safe operation education for personnel engaged in the operation andmanagement of the nuclear reactor facility

3)

Operation of the nuclear reactor facility

4)

Safety reviews on the operation of the nuclear reactor facility

5)

Designation of radiation controlled areas, access controlled areas and environment monitoringareas, and restriction of access to these areas

6)

Ventilation and drainage monitoring equipment

7)

Monitoring of the dose, the dose equivalent, the concentration of radioactive materials and thesurface contamination density of radioactive materials of objects contaminated by radioactive
materials, and the decontamination

8)

Management of radiation measuring instruments

9)

Patrols and checks of the nuclear reactor facility and their associated measures10) Receipt, delivery, transport, storage and other handling of nuclear fuel materials
11) Disposal of radioactive waste12) Measures to be taken in an emergency13) Records on safe operation of the nuclear reactor facility (including compliance of the

Operational Safety Program)
14) Maintenance management of the nuclear reactor facility (except those contained in the next

item)15) Periodic assessment of the nuclear reactor facility16) Quality assurance of the nuclear reactor facility17) Other necessary items for safe operation of the nuclear reactor facility
The “Operational Safety Program” is also subject to change as appropriate, during the usual
operational period.

The LCO of nuclear installations, such as the shutdown margin and reactor thermal limits, etc., areincluded in the Operational Safety Program. Table 19-1 shows the items of the LCO of nuclear
installations in Japan.
Items 14, 15, and 16 are modified as new items after October 2003;19-3

• Maintenance management of the nuclear reactor facility (except those contained in the nextitem): Licensees must establish and implement matters related to policies and objectives ofmaintenance management, a plan for implementation of the maintenance management,
evaluations of the results, corrective actions, records etc,
• Periodic assessment of the nuclear reactor facility: Licensees must perform the Periodic Safety
Review of reactor facilities after the commissioning of a reactor facility every ten years, and,• Quality assurance of the nuclear reactor facility: Licensees must establish the organization in
charge of the quality assurance and the system where planning, implementation, evaluation andimprovement are continuously conducted.

Licensees prepare various kinds of operation manuals and test procedures so as to determine moredetailed operation procedures on the basis of Operational Safety Programs. Licensees, by
establishing committees, assess important matters related to the alternation of Operational Safety
Programs or procedures and safe operation of nuclear reactors in advance of implementation
thereof.
NISA gave an order to change the Operational Safety Programs as a result of the Comprehensive
Check referred in Article 6. The contents of the changes here vary from one licensee to another; theprincipal changes were as follows;• Stronger commitment of the licensees’ management into safety enhancement activities• Ensure independency of Chief Engineers of Reactors and provide sufficient responsibility andauthority in order that they can competently achieve their duties of supervision of safe operationand,• Ensure thorough consistency that actual work shall be conducted in compliance with the official
work procedures etc. for operation management of reactor facilities among the each organizationof the licensee and contractors in charge of maintenance and repair work.
•“Periodic Inspection”
The Periodic Inspection of nuclear power generation facilities (excluding those underdecommissioning) is conducted periodically in order to prevent accidents and failures orpropagation thereof for electric facilities that are provided for power generation, such as nuclear
reactors and associated facilities and steam turbine facilities. It is an obligation to implement theinspection in accordance with the rules defined in Article 54 of the Electricity Utilities Industry Law.
As stated in the law, the Periodic Inspection is conducted in shutdown condition at the interval not toexceed one year plus 13 months for steam turbines and 13 months for other electric facilities (i.e.,
nuclear reactors and associated facilities). During the Periodic Inspection it shall be verified that thesefacilities are maintained and operated in conformance to the application for Construction Plan approval
and the Technical Standards defined by the METI ordinance.

19-4

At the Periodic Inspection, more than 60 items are the subjects for the inspection currently. The
inspection is conducted aiming at the following points in order to ensure the integrity of eachfacility;
1) Conformance to the content of the Construction Plan approved in accordance with Article 47 of
the Electricity Utilities Industry Law and the content of the Construction Plan submitted inaccordance with Article 48 of the said law,2) Compliance to the Technical Standards relating to nuclear equipment,
3) Actions to prevent recurrence of trouble which occurred in the past and,4) Conformance to the content of the Reactor Establishment in accordance with the Reactor
Regulation Law.
•

“Periodic Licensee’s Check and Audit Of Licensee’s Periodic Check System”
Although licensees have voluntarily verified conformance to Technical Standards of the nuclear
power generation facilities heretofore, by the amendment of the Electricity Utilities Industry Law in2003, this action was defined as the “Periodic Licensee’s Check” of licensees which are subject tothe regulatory body’s periodical and mandatory check since then. Specifically, JNES examines theimplementing system of this Periodic Licensee’s Check by reviewing documents and witnessing fromthe standpoint of the inspecting organization, inspection methods, schedule control, recordkeeping,
control of contractors, and appropriateness in education and training. This is called “Audit ofLicensee’s Periodic Check System.” The national government establishes the “Evaluation
Committee on Audit of Licensee’s Periodic Check System for Nuclear Power Stations (hereinafterreferred to as the “Evaluation Committee”)” in NISA. The Evaluation Committee performs acomprehensive evaluation of the implementing system of the licensee related to Periodic Licensee’s
Check referring to the results of the Audit of Licensee’s Periodic Check System performed by NISA,
and officially announces the evaluation results so that the licensee is encouraged to establish theimplementing system to conduct the Periodic Licensee’s Check appropriately, making continuous
improvement.
The “Periodic Inspection” is the inspection especially important to the facilities and for theequipment to ensure the safety of nuclear power generation facilities that the national government
has performed heretofore. Specifically, it is an inspection of the equipment as an important safetyfunction among the “Periodic Licensee’s Check” performed by the licensee voluntarily. It verifies
appropriateness of inspection procedures, inspectors and judgment of results utilizing quality
assurance standards while paying attention to the process of the Periodic Licensee’s Check, by thenational government and JNES attending to or verifying the record.

•“Operational Safety Inspection”
It is required to establish, implement, and continuously assess and improve the quality assurance
program for licensees’ operational safety activities, and to define such quality assurance in the
Operational Safety Program.

19-5

Thereby, the following two matters are positioned comprehensively as implementation of the quality
assurance;
1)

Maintenance management that defines inspection methods etc. of facilities and equipment,
and operation management that defines limitations during operation etc., and,2)

Operational safety activities that licensees perform for ensuring the safety of reactor facilities,
such as fuel management, radioactive waste management, radiation management, andemergency measures.

The national government verifies the appropriateness of the quality assurance program, thesituation of implementation, evaluation, and corrective actions through the Operational Safety
Inspection that inspects such an observance status of Operational Safety Program. The Operational
Safety Inspection was introduced in 2000 and it is conducted 4 times per year for about three weeks
of duration.
•“Periodic Safety Review”
Licensees are also obliged to perform the so-called “Periodic Safety Review” every ten years,
which reviews the situation of the operational safety activities and the situation of reflection of thenewest technical knowledge to the operational safety activities.
At the same time, as measures for aging management, licensees are obliged to perform technicalevaluation of aged deteriorations no later than 30 years after commissioning and to establish aten-year (in length) term maintenance plan based on the evaluation.
The national government verifies through the Operational Safety Inspection whether the operationalsafety activities are appropriately performed reflecting the result of the Periodic Safety Review.
Moreover, the implementation of the long-term maintenance plan is verified at the Operational
Safety Inspection, the Periodic Inspection, and the Audit of Licensee's Periodic Check System (seeSection 14.3 for details).
Concerning the aging management, establishment of a technical assessment and long-term
maintenance plan for reactor facilities had both been required as part of the licensee’s voluntary
efforts. By amendment of the related ministerial orders in October 2003, this has been shifted to the
mandatory duty of licensees and since January 2006 reporting of the results to NISA has beenrequired, while NISA verifies the adequacy of the above assessment.

•“Resident inspectors and the on-site inspection”
Nuclear Safety Inspectors reside in each power station permanently and they make inspection tours
to observe the status of the Operational Safety Program every day, as well as perform theOperational Safety Inspections 4 times per year in accordance with the Reactor Regulation Law.
Moreover, as the result of the above-mentioned “Comprehensive Check,” it is more encouraged thateach Nuclear Safety Inspector’s Office is fully informed of and shared the findings by the Nuclear
Safety Inspector checking the safety of reactor facilities through the free access within the site.

19-6

In addition, an on-the-spot inspection by the personnel of NISA can be performed at any time when
the Minister of METI deems necessary. During the on-the-spot inspection, NISA inspector caninspect documents, records and other objects and question relevant persons through the on-the spot
inspection at the licensee's offices.
•“Chief Reactor Engineer and Responsible Operator”
The Chief Reactor Engineers allocated to each nuclear reactor by the licensee need to have their
qualification authorized by the national examination and the appointment or dismissal of themneeds to be notified to NISA. The Chief Reactor Engineers can offer their opinion to thesuperintendents of the plants when they recognize it is necessary for the safe operation, can give
advice or recommendation to respective duty positions, and can take part in establishing plans forsafe operation.
Responsible Operator is designated by the licensee and allocated for each nuclear reactor. The
mission of the person responsible for operation is to perform the monitoring of the overall operationand the command and supervision of operators. He understands the situation of operations and thepresent condition of the safe operation by periodical patrol to the premises.
The documentation developed and kept by the licensee should include the records relating to fuelassemblies, inspection of the nuclear reactor, operation, radiation management, maintenance,
abnormalities and accidents, and whether it is in accordance with the Reactor Regulation Law.
Moreover, subjects, methods, results, etc. of the inspection should be recorded as a result of the
Periodic Licensee’s Check in accordance with the Electricity Utilities Industry Law.

•

“Cooperation in the investigation on regulatory activity performed by the NSC”
Licensees are obliged to cooperate with the NSC when it conducts investigation by the law
concerning with the reports by NISA which are relating to the regulatory approval and inspection
and so on.

19-7

The overall view of the regulatory activities mentioned above is illustrated below.

Overall View of the Regulatory Activities for Safe Operation
Operation (operational period (less than 13 months) + shutdown period)

Regulatory authority
Operational Safety
Inspection（NISA）

Observance status of safety preservation rule is inspected in accordance with
Paragraph 5 of Article 37 of the Reactor Regulation Law etc.
(Once in 4 months).

Periodic Inspection

Inspection performed periodically in accordance with Paragraph 1 of Article 54of the Electricity Utilities Industry Law etc.

（NISA/JNES）

Inspection

Audit of Licensee's
Periodic Check
System (JNES)

Document check and witnessing of implementing system of Periodic Licensee's
Check in accordance with Paragraph 4 of Article 55 of the Electricity Utilities
Industry Law etc.

Licensee

Review

Manpower
systemPeriodic Licensee's
Check

Licensees Control the Conformity to the Technical Standards for nuclear power
generation equipment voluntarily in accordance with Paragraphs 1 and 3 ofArticle 55 of the Electricity Utilities Industry Law, etc.
---JNES confirms the results of the above

Periodic Safety
Review

Situation of reflection of the newest technical knowledge is evaluated by thelicensee every ten years and a technical evaluation for aging is performed after
30 years operation in accordance with Paragraph 3 of Article 15 of theRegulations for Establishment, Operation etc. of Commercial Power Reactors

Resident Nuclear Safety Inspectors, witnessing inspections (NISA), On-site Inspection
Chief engineer of reactors“Investigation on regulatory activities” (the NSC)

(2) Pre-History Leading to the Present Inspection System
History leading to the present inspection system is described below.
• The “first meeting of the Task Force on the Inspection System was held” in February 2002.

“Task Force on the Inspection System” was established in the Nuclear and Industrial Safety
Subcommittee of the Advisory Committee for Natural Resources and Energy in February 2002, andthe study started. The result was summarized in June 2002 as a report named “On theReexamination of the Inspection System” of the Nuclear and Industrial Safety Subcommittee(hereinafter referred to as the "Nuclear and Industrial Safety Subcommittee Report").
As a fundamental policy for reexamination of the inspection systems, a new philosophy was
introduced to move from the conventional prescriptive approach “to confirm the integrity of the

19-8

facilities through the pre-determined manner which are pre-selected” to more performance-based
approach where “installation methods of facilities and licensee’s safety activities as a whole will bechecked by introducing unannounced inspection”.
• The following actions were recognized as necessary in order to improve the effectiveness of theinspection;(i)to enhance quality assurance activities,(ii)

to introduce an unannounced inspection,(iii) to utilize quantitative risk assessment,(iv)

to apply the performance-base inspection,(v)

to establish criteria and standards as needed,(vi)

to be flexible in applying legal measures and,(vii) to utilize lessons learned from minor troubles.
(viii) “Action taken after an inspection data falsification scandal by TEPCO.” in August 2002
The inspection data falsification by TEPCO was uncovered and study was started to reform thecurrent inspection system at the Nuclear and Industrial Safety Subcommittee, NISA, etc.
(ix)

“Revision of legislations” in October 2003

As the results of the study mentioned above, the Reactor Regulation Law was revised and thepresent framework of the inspection system was introduced. The inspection system is assuming that
operational safety activities including maintenance management by licensees are defined as
activities to implement the quality assurance, while the Operational Safety Inspection, the Periodic
Inspection, and the Audit of Licensee’s Periodic Check System are all instruments of the regulatory
authorities of the national government etc. to check how this system is properly functioning.
In addition, the system for fitness-for-service assessment for cracks was introduced into theregulations at that time. Initially, there had been no definite maintenance standard for cracks inJapan. Therefore, any crack etc. discovered had to be repaired. Introduction of the system forfitness-for-service assessment enabled in-service operation without repairing such cracks etc. withina certain period. Thereby, licensees could make an appropriate and rational maintenance plan.
Reduction of the collective dose by optimizing the number of work projects resulting from theintroduction of fitness-for-service will be attempted.
(3) Reexamination of the Present Inspection System
The inspection system mentioned above has become much more established nowadays as a result of
the accumulation of experience obtained in the past two years or so. However, measures for aging

19-9management are in need of further enhancement, taking into consideration;1) the present situation of conducting the inspection system after two years passage2) the causal analysis of the secondary system pipe rupture accident at the Mihama Unit 3 which
occurred in August 2004,
3) projected increase of number of nuclear power stations under long term operation.
Thereby, the “Task Force on the Inspection System” was resumed in November 2005 and a report of
the Task Force on the Inspection of Nuclear Safety (“Improvement in the Inspection System for
Nuclear Power Generation Facilities”) was issued in September 2006. It pointed out the following
issues, and led to the understanding that these issues required immediate improvement;a. Enhancement of maintenance activities for individual nuclear installation
As for the measures for aging management, licensees are obliged to assess each nuclear
installation and the national government verifies the adequacy of their assessments. In the
Periodic Licensee’s Check or the Periodic Inspection, refined inspections reflecting the individualfeatures of each nuclear installation are not easy, since only the routine type inspection system
has been available. Since conditions of equipment and operation, such as plant operating year,
design of nuclear installation, operating history such as incidents and troubles, etc., andlicensees’ management system including contractors’ vary from one nuclear installation toanother, it is considered more desirable, in order to take thorough measures for maintenanceactivities for a nuclear installation, to ask each licensee to understand the operational status ofeach nuclear installation more individually and perform the maintenance activities based on thisunderstanding even before aging measures have to be introduced.
In order to enhance the measures for aging management, it is necessary to control deterioration
due to aging in a proper manner. It would be effective to introduce new monitoring andevaluation technologies (including on-power monitoring) in order to;-understand the progress of deterioration based on scientific knowledge and conduct lifetime
prediction and,-inspect and repair at an appropriate time by monitoring conditions of equipment and systems
based on the condition-based maintenance. For this purpose, it is necessary to monitor thesituation of deterioration of the equipment and systems of every nuclear installation continuouslyor regularly according to a fixed time period, and to evaluate the trend of deterioration.
Therefore, introduction of an inspection system that asks each licensee to daily enhance themaintenance activities based on the features of each nuclear installation is a new attempt.
b. Further thoroughness to ensure safety in the operational safety activities
The operational safety activities for nuclear installations include those performed during
operation time, such as routine plant patrol, condition monitoring of equipment and surveillance
test of important safety related equipment. Therefore, the regulator should conduct inspections
exactly when these activities are being conducted for verifying appropriateness of the operationalsafety activities.

19-10

Items related to the operational management occupies seventy percent of all the serious
non-conformances in relation to the operational safety management (conditions of notconforming to the requirements of quality assurance) that were revealed at the Operational Safety
Inspection and were ordered to be remedied immediately after the introduction of new inspection
system in October 2003. Therefore operational management represents the essential operationalsafety activities along with maintenance management.
The Nuclear Safety Inspector conducts the safe-operation survey when the inspector is notconducting the Operational Safety Inspection. There are many cases where comparatively minor
non-conformances are identified at the safe-operation survey and remedial actions by the licensee
are to be confirmed at the subsequent Operational Safety Inspection. Taking such cases into
consideration, it is necessary to conduct inspections during the operation time as well in orderthat non-conformances are remedied at an early stage and measures to be taken avoiding an
accident or trouble by means of performing witness inspections etc. and confirming theappropriateness of the operational safety activities of licensees. Introduction of an inspection
system during operation and shutdown, which further ensures licensees’ safety, is also a new
attempt.
It is also necessary to check if the Operational Safety Inspection is efficiently and effectivelyfunctioning and to eliminate duplication in practices by utilizing the results of the Audit ofLicensee's Periodic Check System for the quality assurance of the Periodic Licensee's Check.
c. Thoroughness to remedy non-conformances by licensees
Although the safety level of nuclear installations in Japan is satisfactory in general, the ratio ofaccidents and troubles resulting from human error to the total events does not present downward
tendency. Some human errors have organizational factors or show deterioration of safety culture
and/or organization culture as their background.
Accidents and troubles that occurred due to organizational factors or the deterioration of safety
culture and/or organization culture have been increasing after the latter half of the 1990s. It waspointed out that deterioration of the safety culture of the licensee served as a background to thesecondary system pipe rupture accident of the Mihama Unit 3.
In order to raise the safety level of nuclear installations further, thorough challenge should bemade to eliminate non-conformances due to those factors.
(4) Future Direction for Improving the Inspection System
In order to pursue the above-mentioned three attempts, namely, to enhance individual maintenanceactivities for each nuclear installation, to ensure further safety in operational safety activities, and toeliminate non-conformances thoroughly by licensees, it is necessary to improve the inspection system
as follows and this study by the various organizations is underway with an aim to start the new system
in April 2008;(i)

Introduction of the inspection system as the integrated part of the operational safety activities

19-11

based on the “overall maintenance plan”
a. Further comprehensive “overall maintenance plan”
This “overall maintenance plan” is defined as part of the quality assurance program andcovers the maintenance activities as a whole performed by licensees. It is essential to define
and establish the organization to implement the maintenance activities, scope of thecomponents and structure subject to the maintenance activities and implementing programb. Inspection for the purpose of verifying the situation of the “overall maintenance plan”
This inspection is required to verify that the licensee is actually performing the maintenanceactivities appropriately in accordance with the “overall maintenance plan”. The “overall
maintenance plan” includes maintenance activities during plant operation as well as thoseduring shutdown. For this reason, the inspection performed by the national government
mentioned above also needs to include the inspection during plant operation in addition to thatof during shutdown. Moreover, a system is needed that enables the national government to ask
licensees for necessary alternation or improvement of the “overall maintenance plan” when
improper acts etc. are identified as a result of inspecting the licensee’s implementing situationof the “overall maintenance plan.”(ii) Introduction of an inspection system focused on the important operational actions to ensure
safetya. Inspection system focused on the important operational actions to ensure safety
For introduction of an inspection system to thoroughly ensure safety in the licensees’
operational safety activities, it is necessary to identify the important operational actions toensure safety whether during operation or during shutdown.
b. More efficient and effective inspection system
As a result of performing the inspection also during operation, which has been periodicallyconducted as the Operational Safety Inspection until now, introduction of the new inspection
system focused on the important operational actions to ensure safety during operation andshutdown will enable to avoid the duplication of periods for Periodic Inspection andOperational Safety Inspection,(iii) Development of a guideline for root cause analysis (refer to Section 10.2)(5) Actual examples of the records of the Inspection System etc.
Actual examples of the past records of the Operational Safety Inspection, the Audit of Licensee’s
Periodic Check System and the fitness-for-service assessment are shown in the following:(i) Past records of the “Operational Safety Inspection”

19-12

Two examples of the Operational Safety Inspection conducted in the 2006 fiscal year are shown in
the following. At the early stage of the Operational Safety Inspection, it was likely to detect moreformalistic deficiencies related to the quality management system but now more quality assurance
related aspects are emphasized.
(Example 1) A domestic BWR nuclear power station
June 5 (Mon.) to June 23 (Fri.)• Situation of management review and internal audit (inspection at the head office)• Situation of maintenance management• Situation of quality assurance activities including procurement controls of reactor feedwaterflowmeter etc. (including inspection at the head office)• Surveillance test (manual start-up test of standby gas treatment system); an unannounced
inspection• Situation of improvement measures relating to the past violation• Situation of management review and internal audit (inspection at the head office)• Situation of maintenance management• Situation of quality assurance activities including procurement controls for reactor feedwaterflowmeter (including inspection at the head office)In this Operational Safety Inspection, the “situation of management review and internal audit(inspection at the head office),” “situation of maintenance management,” “situation of quality
assurance activities including procurement controls f of reactor feedwater flowmeter etc.
(including inspection at the head office),” etc. were inspected.
Since the inspection found out items to be observed concerning the plan for piping thickness
control and competence of maintenance management personnel for the “situation of maintenancemanagement” and the action plan for the reactor feedwater flowmeter for the “situation of quality
assurance activities including procurement controls for reactor feedwater flowmeter (including
inspection at the head office),” in the future the improvements should be checked everyday by
the patrol, the Operational Safety Inspection, etc.
In addition, concerning the procurement control for reactor facilities including the feedwaterflowmeter, it was found out that;- effort to avoid the recurrence was in place but still at planning stage and specific approachshould be still needed.
- there are still some room for improvement in terms of avoiding the reoccurrences and control
the procurement methods.
So the monitoring of the implementation situation to avoid recurrence will be continued.
Check of the results of past remedial actions (i.e., “duties of personnel in charge of civilengineering and the building of a power station” and “corrective actions concerning waterchemistry control”), showed that appropriate improvements had been achieved, respectively.
Concerning the situation of routine operations management during the period of the Operational
Safety Inspection, by interviewing the licensee for the situation of operational management offacilities, check of the record of operations, patrol of reactor facilities, witnessing of thesurveillance test (manual start-up test of standby gas treatment system), etc. presented no special
problems.
From the above results, summarizing this Operational Safety Inspection, it is judged that theoperational safety activities concerning the selected inspection items are acceptable in general.

19-13

(Example 2) A domestic PWR nuclear power station
May 29 (Mon.) to June 16 (Fri.)• Situation of fuel management of Unit 2• Situation of plant patrol• Situation of management review• Situation of work management• Situation of education on operational safety: an unannounced inspection• Witnessing surveillance tests (turbine driven auxiliary feed pump start-up test, etc. of Unit 1
and 2)• Situation of fuel management of Unit 2• Situation of plant patrol

In this Operational Safety Inspection, the “situation of fuel management of Unit 2,” “situation ofplant patrol,” “situation of management review,” “situation of work management,” etc. were
inspected.
As the results of the inspection items for remedy were found in the following;- unclear descriptions in the documentations- insufficient control of the component parts in stock
in relation to the procedures of “situation of plant patrol,” “situation of work management,” and“situation of education on operational safety,” Check should be done on the corrective action
through routine patrol, the Operational Safety Inspections, etc.
Concerning the situation of routine operations management during the period of the Operational
Safety Inspection, by interviewing the licensee for the situation of operational management offacilities, check of the record of operations, patrol of reactor facilities, witnessing of thesurveillance test (turbine driven auxiliary feed pump start-up test, etc. of Unit 1 and 2), etc.
presented no special problems.
From the above results, summarizing this Operational Safety Inspection, it is judged that theoperational safety activities concerning the selected inspection items are acceptable in general.

(ii) Past records of the “Audit of Licensee’s Periodic Check System”
The actual application of the Audit of Licensee’s Periodic Check System is shown in thefollowing:
JNES selects items of the actual audit from the Periodic Licensee’s Check and performs the audit
by document check and witness. The basic idea of the sampling is as follows;(a)Number of items subject to the audit shall be around ten percent of that of the Periodic
Licensee’s Check items, and shall be dependent on the preceding evaluation result,(b)

Items for the audit shall be selected without any prejudice taking into consideration thetypes of inspection, equipment and commercial contract within the scope of the Periodic
Licensee’s Check .

(c)

The preceding audit result at the place of business shall be reflected. Especially, if there wasany indicated item or finding in the preceding audit, its improvement should be checked and,

19-14

(d)

If the preceding audit identified the items for further monitoring, those items should bechecked with priority.

The national government evaluates the results of the Audit of Licensee’s Periodic Check Systembeing informed from JNES, indicates the rating according to the following three categories andinforms the licensee of the evaluation results;
A: The implementing organization of the licensee is capable of meeting the requirement for the
Periodic Licensee’s Check voluntarily and appropriately,
B: The implementing organization of the licensee is capable of meeting the requirement for the
Periodic Licensee’s Check voluntarily and appropriately despite room for improvement and
C: The implementing organization of the licensee is capable of meeting the requirement for the
Periodic Licensee’s Check voluntarily despite considerable room for further improvement.
Incentives based on evaluation results
An incentive type control where selection of items for next audit is dependent of the rating of
the audit of the preceding year has been conducted in order to encourages licensee’s effort forsafety and ensure transparency and reliability of the Periodic Licensee’s Check.
Past records of the audit after this audit system was started are shown below.

Audit Result
Fiscal Year

Number of Audit
A

B

C

2003

12

2

4

5

2004

37

1

28

7

2005

35

4

26

2

2006

24

–

2

–

Total

108

7

60

14

19-15

An example result of the Audit of Licensee’s Periodic Check System of a specific power station in2006 fiscal year is shown below.
ExamplePresident, A domestic Electric Power Co., Inc.

Applicant of Audit ofLicensee’s Periodic
Check System(Application date: x/x/2005 Application Number: xxxxx No. xxx)

Items subject to audit

Periodic licensee’s check in xx Periodic Inspection for xxxx Nuclear
Power Station, Unit x

Audit of Licensee's
Periodic Check System

1. Audit period(NISA)

x/xx/2005 to x/x/2005
2. Notification date of results of Audit of Licensee’s Periodic Check
Systemx/x/2005 (Notification number: 05xxxx-xxxx)
3. Summary of audit results
According to JNES’s notification and description of the results of the
Audit of Licensee's Periodic Check System, it identified no significant
non-conformity, but identified three items determined necessary to be
improved. Corrective actions for two of the three items wereconfirmed during the audit period. Any corrective action for anotherone could not be confirmed during the audit period, and it shall befollowed up in the next Audit of Licensee’s Periodic Check System forthe said power station.
Concerning follow-up of the four items that had been determined to be
improved in the Audit of Licensee’s Periodic Check System for Unit x,
Unit x, Unit x and Unit x which had been subject to the precedent
audit, corrective actions for two items were confirmed. Effort for thecorrective actions for the other two items effort was visible but could
not be confirmed because there was no inspection applied this time forthe unit. Those shall be followed up in the next Audit of Licensee’s
Periodic Check System.
Consequently JNES judges that the quality management system is in
general functioning well and the Periodic Licensee’ Check is
conducted with the voluntary and appropriate system. Since thecorrective actions for the three items which were notified for
improvement during the audit of this unit and the units subjected to theprecedent audit could not be confirmed this time, further investigationand effort for the improvement are encouraged steadily and addressed
continuously.
4. Audit item
Thirteen items of document audit and actual spot audit (inspection of(1) monitoring system, (2) steam turbine facility, (3) core-internal,
etc.)

Evaluation

1. Audit result: B(NISA)

2. Notification date of the evaluation
xx/x/2006 (Notification number xxxxxx Nuclear No.x

19-16

3. Reason for the audit result (result and grounds)
NISA carefully examined the audit results of the notification
concerned and the explanation by JNES, and determines that three
items, which were determined necessary to be improved but could not
be verified for adequacy of the corrective action, needs to bemonitored continuously.
Consequently, NISA determines that the implementing organization of
the licensee is capable of meeting the requirement for the Periodic
Licensee’s Check voluntarily and appropriately despite room for
improvement.
4. Situation of holding of the evaluation committee
Hearing of the explanations concerning the audit result, Q&A,
x/xx/2006
Study on the evaluation result, x/xx/2006
5. Special notes for rating
N/A
Others(iii) Historical experiences of fitness-for-service assessment
Since the assessment system for fitness-for-service was introduced, cracks were found on the
shroud etc. of total of 15 nuclear power generation units, and the fitness-for-service assessments
have been performed. Situation of crack propagation is monitored and whether these meet thesafety level or not is verified at the time of the Periodic Inspection performed about every 13
months. (As of October 2006)
Licensee
Chubu Electric Power
Co., Inc.
Tohoku ElectricPower Co., Inc.
The Chugoku ElectricPower Co., Inc.
Shikoku ElectricPower Co., Inc.
Tokyo Electric Power
Co., Inc.
The Chugoku ElectricPower Co., Inc.
Tohoku ElectricPower Co., Inc.
Chubu Electric Power
Co., Inc.

Power
Station

Unit

Notification
Date

Hamaoka

Unit 4

12/21/2004

Onagawa

Unit 1

01/06/2005

ShimaneUnit 2

02/09/2005

Ikata

Unit 103/01/2005

Kashiwazaki Kariwa

Unit 3

04/13/2005

ShimaneUnit 2

04/13/2005

Onagawa

Unit 2

05/27/2005

Hamaoka

Unit 305/24/2005

19-17

Part Subject to Evaluation(1) Crack on weld of the shroud support
ring(2) Crack on the shroud lower ring andlower shell
Crack on the shroud middle ring andlower ring
Crack on the PLR piping
Micro-crack on the surface of the reactor
vessel inlet nozzle
Crack near outside weld-line of the
shroud middle shell and lower ring
Crack on inside weld-line of the shroud
middle shell
Crack inside the shroud support ring
(1) Crack on the shroud support ring(2) Crack on the shroud support cylinder
and weld on the support leg

The Japan Atomic
Power Company Co.
The Chugoku ElectricPower Co., Inc.
Tokyo Electric Power
Co., Inc.

ShimaneTokyo Electric Power
Co., Inc.
Shikoku ElectricPower Co., Inc.
Tokyo Electric Power
Co., Inc.
Tokyo Electric Power
Co., Inc.

07/13/2005

Crack on the vertical weld-line on the
shroud support cylinder

Unit 1

07/20/2005

Crack on the PLR piping

Fukushima
Daini

Unit 308/18/2005
04/19/2006

Crack on the PLR piping

Kashiwazaki Kariwa

Unit 2

11/04/2005

Ikata

Unit 2

02/26/2006

Unit 1

04/19/2006

Crack on the PLR piping

Unit 304/19/2006

Crack on the PLR piping

Tokai

Kashiwazaki Kariwa
Kashiwazaki Kariwa

Crack on the weld-line between the
shroud middle shell and the shroud
support ring
Micro-crack on the surface of the reactor
vessel inlet nozzle(PLR: Primary Loop Recirculation)

19.4 Response to Accidents and Anticipated Operational Occurrences
Licensees are required to include the items related to “operation of nuclear reactor facility” in the
Operational Safety Program.
Those include the operational procedures for accidents and anticipated operational occurrences(abnormal events) as well as normal operation are described so as to cope with any incidents orabnormal events smoothly.
In the procedures of “measures for any abnormal events,” the following matters are included:• Recognition of the situation;• Elimination of the cause;• Emergency measures;• Measure after reactor automatic scram and,• Manual startup of an emergency AC power supply and a gas treatment system.

Moreover, licensees are required to prepare “emergency measures” in the operation of a nuclear
reactor facility stipulated by the Reactor Regulation Law.
As “emergency measures,” the licensee defines the following in the Operational Safety Program:• Establishment of the nuclear emergency preparedness organization in accordance with therequirements of the Special Law of Emergency Preparedness for Nuclear Disaster;• Preparation of resources necessary for nuclear emergency preparedness;• Maintenance of the communication system among the related parties;

19-18

• Implementation of nuclear emergency exercises;• Official announcement of the emergency system and,• Clearance of the emergency preparedness organization, etc.

The details of emergency preparedness are described in Article 16.
19.5 Engineering and Technical Support
In Japan, the reliability verification test and safety research programs on major components andequipment have been carried out to enhance the safety of nuclear installations. Promotion of thesetests and research is shown in Section 14.8.
Main subjects of related research are shown below:(1) Study on aging of light water reactors;• Survey of technology related to measures for aging management (accumulation of technicalknowledge, aging evaluation technology for cable, etc.),• Demonstration of the nondestructive inspection technology of component material for nuclear
reactors as technical measures for aging management, and,• SCC crack-growth evaluation technology etc. as measures for aging management (international
information sharing in OECD/NEA).

(2) Study on probabilistic safety assessment for nuclear installations etc. and,• Study on utilization method of risk information and,• Study on Nuclear Emergency Preparedness.

(3) Study on safety, such as seismic safety of nuclear installations.
• Clarification study on rupture process of wall-thinned pipe under seismic load

19.6 Reporting of Incidents
Reactor Regulation Law and the Electricity Utilities Industry Law require licensees to report thesituation and measures taken to the incidents or failures occurred in nuclear installations to NISA. The
reporting criteria prescribed in these laws are shown in Table 19-2.
Licensees are making efforts to perform feedback of the lessons learned from the situation andmeasures taken to these incidents, and the corrective actions which were derived by root cause
analysis to other nuclear installations.
As the result of the “Comprehensive Check” referred in Article 6, incidents related to the move of the

19-19control rod when it is not allowed to be operated is now the subject for the reporting to the national
government.
The frequency of unplanned reactor shutdowns per year of nuclear installations in Japan is around 0.2
times per reactor-year in recent years and is well below the world average.
The International Nuclear Event Scale (INES) was introduced in August 1992 to assess events that
occurred in commercial power reactor facilities. INES results in Japan are shown in Annex 2.5.
Of the past cases identified by the “Comprehensive Check,” those that NISA rated as 1 or over to theelectric utilities of Japan are as follows:

Cases identified by the Comprehensive Check relating to power generation facilities
Event Date
December 24,
1996

June 11, 1998

June 18, 1998

Facility Name
The Japan Atomic Power
Company
Tsuruga Power Station
Unit 2
Tohoku Electric Power Co., Inc.
Onagawa Nuclear Power Station
Unit 1
Hokuriku Electric Power Co.
Shika Nuclear Power Station
Unit 1

Subject

INES

Leak from piping of the chemical
volume control system etc.

1

Reactor automatic shutdown due tohigh neutron flux during reactor
shutdown operation

1

Criticality accident

2

19.7 Collection, Utilization and Sharing of Operating Experience Information
NISA makes a public news release and reports to the NSC incidents or failures upon immediate receiptof the information from the licensees. NISA also makes a public news release and reports to the NSC
the causes and recurrence-preventive measures when the investigation is completed. The NSC points
out issues on the content of the report when necessary. The NSC has established the Special
Committee on Analysis and Evaluation of Nuclear Accidents and Failures and investigated andreviewed in and outside Japan. In March 2007 this committee has summarized the guidance to use theincident and failure information of nuclear installations.
NISA assesses each incident or failure in detail to abstract the lessons learned with respect to the safety,
being advised by subcommittee members of the Advisory Committee on Nuclear and Industrial Safety,
who are experts on operation management, inspection and radiation control.
JNES has the system to collect and analyze safety information in and outside Japan. JNES provides
collected safety information and the results of analysis for NISA for quick sharing with NISA. JNES
and NISA have jointly established the “Safety Information Review Meeting” to evaluate and to takeadequate regulatory measures. The “Safety Information Review Meeting” is held periodically.

19-20

In regard to the international information exchange by NISA and JNES, a mechanism has been
established to share the information on nuclear incidents and failures with the IAEA and OECD/NEA
as well as in the bilateral corporations with China, France, Korea, Sweden and the United States.
Meanwhile, the licensees established the specialized organization named, “Japan Nuclear Technology
Institute” (JANTI), in March 15, 2005 in order to develop the technical infrastructure and promote thevoluntary operational safety activity for the purpose of contributing to the activation of nuclear powerindustry. JANTI unified, reorganized and inherited the activities conducted by the Nuclear Information
Center (NIC) of CRIEPI and the activities of the Nuclear Safety Network (NS Net) conducted by the
research organizations and nuclear industry. JANTI collects and analyses domestic and overseas
operational experience information. Concerning the safety information of domestic nuclear
installations, they developed a system named “NUCIA” that is a nuclear information publication
library, with which information including minor events can be shared all over Japan, and it is postedon the Internet site of JANTI. In May 2007, JANTI reexamined the criteria to register the incidents andevents to NUCIA.
Overview of the “NUCIA” is shown in Fig. 19-1.
Moreover, the licensees and JANTI perform overseas information exchange through the Institute Of
Nuclear Power Operation (INPO) and the World Association of Nuclear Operators (WANO) Tokyo
Center. Furthermore, each licensee utilizes individual agreements on information exchange withutilities and manufacturers in France, Germany and the United States.
There are many feedback of operating experiences by licensees, which are reflected in preventive
maintenance and planned repair and replacement of parts. Examples for BWR are replacements of thecore shroud and the in-core monitoring housing etc. An example for PWR is replacement of the upper
head of reactor vessel.
At manufacturer side as well, for the purpose of information, two groups are formed, firstly “JapanBWR Owners Group” consisting of the electric power companies of Tohoku, Tokyo, Chubu, Hokuriku,
Chugoku, JAPC and J-POWER and Toshiba/Hitachi and secondly “Japan PWR Owners Group”
consisting of the electric power companies of Hokkaido, the Kansai, Shikoku, Kyushu and JAPC and
Mitsubishi Heavy Industry and Mitsubishi Electric.
19.8 Management of Spent Fuel and Radioactive Waste
Details are described in the Second National Report of Japan (October 2005) for the "Joint Convention
on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management".

19-21

Table 19-1 (Part 1) Items of Limiting Conditions for Operation (BWR)
SystemItem of limiting conditions for operation

Reactivity control
systemShutdown margin, reactivity monitoring, control rod motion monitoring,
control rod scram time, control rod operation, boron water injection system

Power distribution

Reactor thermal limit, reactor thermal power and core flow

Control &Instrumentation

Instrument and control equipment

Reactor coolant
systemReactor re-circulation pump, jet pump, main steam relief and safety valve,reactor coolant leak rate, system pressure monitoring of the emergency corecooling system and reactor isolation cooling system, concentration of iodine
131 in reactor coolant, reactor shutdown cooling system, limit of temperature& temperature change rate limit of primary coolant, reactor pressure, reactor
water level when moving a fuel or control rod

Emergency corecooling system

Emergency core cooling system, reactor core isolation cooling system

Reactor containment
vessel system

Main steam isolation valve, reactor containment vessel & isolation valve,vacuum break valve from suppression chamber to drywell, averagetemperature of suppression pool, flammability control system, oxygen
concentration in containment vessel, reactor building, reactor building heating
and ventilation isolation valve, standby gas treatment system

Plant system

Cooling system and cooling sea water system for residual heat removal
system, emergency diesel generator cooling system, cooling system andcooling sea water system for diesel generator of high pressure core spraysystem, water level & temperature of spent fuel pool, central control roomheating and ventilation system

Emergency power
supply system

Offsite power supply system, emergency diesel generator, emergency diesel fuel,dc power supply, station power system

OthersWithdrawal of single control rod during reactor shutdown, removal of single
control rod drive mechanism, inspection with withdrawal of multiple control rods,
in-service leak-rate or hydrostatic test, inspection with switching of reactor
mode, inspection causing the temperature increase in the reactor

19-22

Table 19-1 (Part 2) Items of Limiting Conditions for Operation (PWR)
SystemItem of limiting conditions for operation

Reactivity control
systemShutdown margin, critical boron concentration, moderator temperature
coefficient, control rod motion function, control rod insertion limits, control
rod position indication, physics tests, chemical and volume control system(function of boron concentration)

Power distribution

Reactor thermal power limit, heat flux hot channel factor, nuclear enthalpy rise
hot channel factor, axial neutron flux difference, quadrant power tilt ratio

Control &Instrumentation

Instrument and control equipment

Primary coolant
systemDNB ratio, temperature & pressure and temperature change rate of primarycoolant, primary coolant system, pressurizer, pressurizer safety valve,pressurizer relief valve, low temperature over-pressurization protection,
primary coolant leak rate, steam generator tube leak monitoring, leakmonitoring to residual heat removal system, iodine 131 concentration inprimary coolant

Emergency corecooling system

Accumulator tank, emergency core cooling system, refueling water storagetank, boron injection tank

Reactor containment

Reactor containment vessel, reactor containment vessel vacuum relief valve,reactor containment vessel spray system, annulus air cleanup system, annulus

Plant system

Main steam safety valve, main steam isolation valve, main feedwater isolation
valve, main feedwater control valve, main feedwater bypass control valve,main steam relief valve, auxiliary feedwater system, condensate water tank,component cooling water system, sea water system for component coolingwater system, emergency circulation system of control room, air cleanup
system of safety auxiliary equipment room, air cleanup system of fuelhandling building

Emergency power
supply system

Offsite power supply, diesel generator, emergency diesel fuel & lubricating oil andstarting air for emergency diesel generator, emergency dc power supply, station
bus bar for emergency

OthersBoron concentration in primary coolant, water level of reactor cavity, reactor
containment penetrations, water level & temperature of spent fuel pit,pressure-proof and leak inspections of reactor coolant system, leak rate
inspection of safety injection system check valve

19-23

Table 19-2 Reporting Criteria of Incidents and Failures Provided in Legislations
Provision of the Reactor Regulation Law
1. When nuclear fuel material was stolen or lost.
2. When a reactor was shut down by failure of a reactor facility or when it became necessary to
shut down a reactor during operation, or when reactor power output changes more than 5%, orwhen reactor power output change of more than 5% was required. Except when it was one of
the following and the licensee announced officially about the situation of the concerned
failure.
i) When it occurs in the term of the Periodic Inspection provided in Article 54-1 of theElectricity Utilities Industry Law (Law No. 170, 1964) i.e. the failure in the equipmentsthat the functional and operational conditions of the failed equipment cannot be checked
under the reactor shutdown condition.
ii) When the failure did not cause any deviation from the limit of operation (it is arequirement defined in the Operational Safety Program for operation of the nuclear
reactor facility, and when it cause any deviation from the concerned conditions themeasure that the licensee should take is also defined in the Operational Safety Program,the same in this paragraph), and there is no change observed related with the concerned
failure, and when the licensee performs inspection of the failed equipment concerned.
iii) When the reactor output change is required to follow the limit of operation.
3. When a licensee checked the equipment and structure important to the safety of the nuclear
reactor facility provided by the Minister of METI (hereinafter called as "equipment etc.
important to safety" in this paragraph as), and when concerned equipment etc. important tosafety was considered that it does not satisfy the standard described in Article 9 or in Article
9-2 of the Ministerial Ordinance of Establishing Technical Standards for Nuclear Power
Generation Equipment (Ordinance No. 62 of MITI, 1965), or when it was considered that it
does not have function to ensure safety of the nuclear reactor facility.
4. When there was a failure of equipment etc. important to safety by the fire. Except theconcerned failure was associated to the measure of fire extinguishing or prevention of the
spread of fire.
5. Except for the preceding three items, when deviation from the limit of operation by the failureof a the nuclear reactor facility (except those minor troubles whose impact on operation of
nuclear reactor is insignificant) was caused, or when the measure for the concerned deviation
defined in the Operational Safety Program was not implemented at the time of deviation fromthe limit of operation.
6. When the failure of a reactor facility or occurrence of other undetermined situation had caused
any trouble to discharge gaseous radioactive wastes through the ventilation facility or todischarge liquid radioactive wastes through the drainage facility.
7. When the concentration of radioactive materials in the air outside the environment monitoringarea exceeds the allowable limit in the case of discharge of gaseous radioactive wastes throughthe ventilation facility.
8. When the concentration of radioactive materials in the water outside the environment
monitoring area exceeds the allowable limit in the case of the discharge of liquid radioactive
wastes through the drainage facility.
9. When nuclear fuel materials or materials contaminated with nuclear fuel materials (hereinafterreferred to as "nuclear fuel materials etc.") leaked out of the controlled area.

19-24

10. When nuclear fuel materials etc. leaked within the controlled area associated to failure of a
nuclear reactor facility or occurrence of other undetermined situation. Exceptions are thefollowings (except the case when new measures such as access control into the leakage-related
place and key control have been taken or when the leaked substances have spread outside thecontrolled area):i) When revealed liquid nuclear fuel materials etc. did not spread out of the floodgate that is
installed in the circumference of the equipment of the concerned leakage for prevention ofleakage spread.
ii) When the ventilation facility of the concerned area of the leakage was working properlyat the time when gaseous nuclear fuel materials etc. leaked.
iii) When the amount of radioactivity of the leaked nuclear fuel materials etc. is very little
and when the degree of the leakage is minor.
11. When the person who enters into the controlled area suffered radiation exposure due to thefailure of a nuclear reactor facility or occurrence of other undetermined situation, and whenthe effective dosage of concerned exposure exceeds or could exceeds five mSv for a personnel
engaged in radiation work or 0.5 mSv for a person other than the personnel engaged inradiation work.
12. When the dosage of personnel engaged in radiation work exceeds or could exceed theallowable dose limit.
13. Other than those above items, when persons are injured or could be injured in the nuclear
reactor facility (except when the injury was other than radiation hazard and was not necessary
for hospitalization).
Provision of the Electricity Utilities Industry Law
1. Fatal and injury accidents associated to electric shock, or breakdown of the electric structureof a nuclear power generation, or miss-operation, or omission of the necessary operation of theelectric structure of a nuclear power generation (limited to the case of death or being
hospitalized at hospital or clinic for treatment).
2. Electric fire accident (limited to the case of more than the partial destruction by fire. Except
matters referred to in the previous item and the next to the fifth item).
3. Causing damage to public property due to failure of the electric structure of a nuclear power
generation, or miss-operation, or omission of necessary operation of the electric structure,
accident by which usage of road, park, school and other institution or structure were madeimpossible or accident which did influence socially (except referred to previous two items).
4. Breakdown accident of main electric structures (except referred to previous three items and thenext item).
5. Incident that influenced other electric utility to force suspension of electric power supply ofthan 7,000 kW and less than 70,000 kW for more than one hour, or suspension of electric
power supply more than 70,000 kW for more than ten minutes, associated to the breakdown of
the electric structure of a nuclear power generation, or miss-operation, or omission ofoperation necessary for the electric structure of a nuclear power generation.

19-25

Table 19-3 Power station performing “rated thermal power operation”
Electric power company
Hokkaido Electric Power Co., Inc.

Tohoku Electric Power Co., Inc.

Tokyo Electric Power Co., Inc.

Nuclear installation

Date of Introduction

Tomari Unit 1

06/18/2003

Tomari Unit 2

02/21/2003

Higashidori Unit 1–

Onagawa Unit 1

08/11/2003

Onagawa Unit 2

12/08/2003

Onagawa Unit 3

05/06/2003

Fukushima Daiichi Unit 1–

Fukushima Daiichi Unit 2–

Fukushima Daiichi Unit 3

12/25/2003

Fukushima Daiichi Unit 4–

Fukushima Daiichi Unit 5

12/25/2003

Fukushima Daiichi Unit 6

02/22/2005

Fukushima Daini Unit 1

09/03/2003

Fukushima Daini Unit 2

07/09/2002

Fukushima Daini Unit 3
Fukushima Daini Unit 4

Chubu Electric Power Co., Inc.

Hokuriku Electric Power Co.
The Kansai Electric Power Co., Inc.

05/27/2002

Kashiwazaki Kariwa Unit 1

04/15/2004

Kashiwazaki Kariwa Unit 2

05/07/2002

Kashiwazaki Kariwa Unit 3

04/27/2004

Kashiwazaki Kariwa Unit 4

07/31/2003

Kashiwazaki Kariwa Unit 5

05/07/2002

Kashiwazaki Kariwa Unit 6

07/04/2003

Kashiwazaki Kariwa Unit 7

08/06/2002

Hamaoka Unit 1–

Hamaoka Unit 2

08/21/2003

Hamaoka Unit 3

12/10/2003

Hamaoka Unit 4

11/28/2003

Hamaoka Unit 5

12/20/2004

Shika Unit 1

04/01/2003

Shika Unit 2–

Mihama Unit 1

11/02/2002

Mihama Unit 2

07/17/2002

Mihama Unit 3

06/19/2003

Takahama Unit 1

02/15/2003

Takahama Unit 2

06/06/2002

Takahama Unit 3

11/06/2002

Takahama Unit 4

06/17/2003

Ohi Unit 1

06/04/2003

Ohi Unit 2

12/18/2002

19-26

The Chugoku Electric Power Co., Inc.

Shikoku Electric Power Co., Inc.

Ohi Unit 3

02/25/2003

Ohi Unit 4

04/16/2002

Shimane Unit 1

01/05/2004

Shimane Unit 2

11/14/2003

Ikata Unit 1

05/10/2002

Ikata Unit 2
Ikata Unit 3

Kyushu Electric Power Co., Inc.

The Japan Atomic Power Company

04/16/2002

Genkai Unit 1

03/20/2002

Genkai Unit 2

03/22/2002

Genkai Unit 3

03/07/2003

Genkai Unit 4

11/12/2002

Sendai Unit 1

03/20/2002

Sendai Unit 2

06/28/2002

Tokai No.2

12/20/2002

Tsuruga Unit 1

03/14/2003

Tsuruga Unit 2

07/15/2002

19-27

Fig. 19 -1 Use of Operation and Maintenance Information by License Holders

D Company

Operationand maintenance information
自社プラントの運転・保守情報

19-28

Patrol and check
duringnormal operation
運転中の巡視・点検

Abnormal events during
運転中に異常事象を検知
operation(power surge,（出力変動、蒸気漏れ、水漏
steam
leak, water leak,
れ等）
etc.)

Maintenance
通常の保全プログラム
based on the normalに基づく保守maintenanceprogram・ Faults discovered
during the periodic
・定検中に不具合を発見
inspection

Measures
他社情報のbased水平展開に
on the feedback
Of information
from
基づく対策

（機器の損傷、ひび割れ等）
・
Measures forとして・あるいは予防保全
preventive
対応措置を決定

maintenance

国へ報告
必要な対Report
応措置to
をthe
実施
Safety法律
Government
Significant
対象トラブル(Regulatory Req.)（安全上重要なもの）
Safety Related(Quality
Information)
保全品質情報（安全に関連するもの）

Othersその他情報(Nonsafety)（安全に関連しないもの）

other companies

Each company officially
各社にてthrough、プレス発表、ある
announcespresses or
いはhomepages.
ホームページ上で公表

Evaluation
事象の評価
Implement necessary measures

C Company
B Company
Ｂ社
AＡ社
Company

Official
announcement
公表
Official
announcement

公表

Each company
各社が公表範囲を決定
determines（地元の安心等を考慮）
the scope of official
announcement(with considerations
for the local public)A
Ａ
A
Measures required

Analysis and分析・評価
Evaluation

Obtaininformation of other companies
他社情報の入手（電気事業者で共有）(to share among license holders)
国内他社情報・法律対象トラブル
Domestic
Information
・保全品質情報

Foreign Information

海外他社情報（ WANO、ＩＮＰＯ、 NRC 等から入手）(WANO, INPO, NRC, Venders etc.)
海外メーカー情報他

WANO: World Association of Nuclear Operators
INPO: Institute of Nuclear Power Operations
NRC: Nuclear Regulatory Commission, USA

ニューシア
データベースNUCIA database
Sharing and public presentation of

法律対象トラブル、および保全品
registered information on
troublessubject to the regulatory
情報を登録して共有化・公開
and quality information

Activities for Improvement of Safety

Safety Improvement-1

Activities for Improvement of Safety
Activities for improvement of safety are described in each article, but the future activities arecompiled in this report.
(1) Assessment of the Seismic Safety of Existing Nuclear installations Following the
Revision of the Regulatory Guide for Reviewing Seismic Design of Nuclear Power
Reactor Facilities (Section 14.5)
The Nuclear Safety Commission revised the Regulatory Guide for Reviewing Seismic Designof Nuclear Power Reactor Facilities, etc. in September 2006
Following the revision, NISA required licensees to assess the seismic safety for existing
nuclear installations in accordance with the revised Regulatory Guide for Reviewing Seismic
Design of Nuclear Power Reactor Facilities, and to report the results.
Each licensee submitted its implementation plan for the seismic safety evaluation to NISA in
October 2006.
Each licensee submitted the process to implement the safety assessment within two to three years
respectively, and the implementation plan describes that the assessment will be implemented in theorder of geological and earthquake survey, determination of the reference earthquake groundmotion, and seismic safety evaluation.
NISA decided to confirm the adequacy of contents of the licensees’ reports for the seismic safetyassessment in the order of submittal.

(2) Training of Human Resources in Nuclear Fields (Section 11.4)
In order to ensure safety of nuclear power generation, highly capable human resources shouldbe trained and ensured. In Japan, it has been an issue to keep future human resources, under
the severe conditions of low birthrate and increase of elderly people, decrease in population,
retirement of skillful engineers, and decrease of construction opportunities of nuclear
installation.
The Ministry of Education, Culture, Sports, Science and Technology and the Ministry of
Economy, Trade and Industry will implement the nuclear human resource training program
focusing on the following items from 2007 fiscal year; i) support of educational activities,
such as enhancement of education and research in nuclear field, of internship for nuclear
engineering students, and of preparation of core curriculum for the nuclear engineering, ii)
support of research activities focusing on fundamentals and infrastructure of nuclear
technology which might contribute to bring up researchers in nuclear fields.

Safety Improvement-2

(3) Ensuring Safety of Existing Nuclear installation (Section 14.3 (2))
NISA imposes on licensees the duty for the periodic safety review conducted every ten years
interval, and assesses the feedback of the operating experiences and latest technicalknowledge and probabilistic safety assessment. NISA also continues the assessment as a partof the periodic safety review from the viewpoint of improvement in aging management andprevention of degradation of licensee's organizational climate.
As measures for aging management of existing nuclear installation, NISA enhances theperiodic safety review, the Periodic Inspection, makes efforts for preparation of the technicalstandards, and promotes technological developments.
(4) Nuclear Emergency Drill (Section 16.3)
In executing emergency exercises, it is important for relevant persons on emergency
preparedness of the national government, local governments, licensees and residents tounderstand the measures for nuclear emergency preparedness and to take actions
appropriately in an emergency. The emergency exercises will be continued, including
participation to international drills.
(5) Promotion of Nuclear Safety Research (Section 14.6)
The NSC proposed nuclear safety researches (prioritized nuclear safety researches) thatshould be performed selectively for about five years from 2005 fiscal year. In the proposal, thefollowing important research areas were proposed as safety researches for nuclear
installations; 1) the regulatory system area (example: use of risk information, assessment ofroot-cause of failures and accidents), 2) light water reactor area (example: safety analysis,material degradation and aging management, seismic safety technologies) and nuclear reactor
emergency prevention technologies. The advancement of these studies is checked at theSpecial Committee on Nuclear Safety Research of the NSC .

Safety Improvement-3

This page intentionally left blank

Safety Improvement-4

ANNEXES

This page intentionally left blankTABLE OF CONTENTS
Annex 1 List of Nuclear Installations ........................................................................................... A1-1
Annex 2 Data on Nuclear Installations......................................................................................... A2-1
2.1 Capacity of Electricity of Commercial Nuclear Power Reactors ............................................. A2-1
2.2 Capacity Factor of Commercial Power Reactors ....................................................................... A2-1
2.3 Frequency of Unplanned Shutdown at Commercial Nuclear Power Reactors ......................... A2-2
2.4 Reported Events (by Laws & Notifications) of Commercial Nuclear Power Reactors ........... A2-2
2.5 Assessment of Events by INES for Commercial Nuclear Power Reactors .............................. A2-3
2.6 Human Error Induced Events Reported ..................................................................................... A2-3
2.7 Dose per Persons at Commercial Nuclear Power Reactors........................................................ A2-4
2.8 Averaged Dose at Commercial Nuclear Power Reactors ......................................................... A2-4
2.9 Radioactive Gaseous Waste (I-131) Released from Commercial Nuclear Power Reactors ..... A2-5
2.10 Radioactive Liquid Waste (except H-3) Released from Commercial Nuclear Power ReactorsA2-5
2.11 Radioactive Solid Waste Generation per Electricity Generation of Commercial Nuclear Power
Reactors .......................................................................................................................................... A2-6
Annex 3 Legislation and Guidelines ............................................................................................. A3-1
(1) The Rule for the Installation, Operation, etc. of Commercial Nuclear Power Reactors(Excerpt)........................................................................................................................................... A3-1
(2) Ordinance of the Ministry for Establishing Technical Standards for Nuclear Power Generation
Systems ............................................................................................................................................ A3-2
(3) The Law for the Independent Administrative Agency Japan Atomic Energy Agency(Excerpt)......................................................................................................................................... A3-16(4) Regulatory Guide for Reviewing Seismic Design of Nuclear Power Reactor Facilities.......... A3-16(5) Regulatory Guide for Reviewing Technical Competence of Nuclear Operators ...................... A3-23(6) Emergency Preparedness Guide "Emergency Preparedness for Nuclear Facilities"(Excerpt)......................................................................................................................................... A3-25
Annex 4 The NSC views on, and future actions to take for, the impacts due to the Niigata-ken
Chuetsu-oki Earthquake in 2007 ..................................................................................................... A4
Annex 5 Answers to the Synopsis of the Relevant IAEA Safety Requirements ........................... A5i

This page intentionally left blankii

Annex 1

List of Nuclear Installations(As of August, 2007)(1)

Commercial Power Reactors
License Holder
The Japan Atomic
Power Co.
Hokkaido ElectricPower Co., Inc.

Tohoku ElectricPower Co., Inc.

Tokyo ElectricPower Co., Inc.

In
Operation

Chubu ElectricPower Co., Inc.

Hokuriku ElectricPower Co.

Power Station & Unit
Tokai II Power Station
Tsuruga Power Station,
Unit 1
Unit 2
Tomari Power Station,
Unit 1
Unit 2
Onagawa NPS,
Unit 1
Unit 2
Unit 3
Higashidori NPS, Unit 1
Fukushima Daiichi NPS,
Unit 1
Unit 2
Unit 3
Unit 4
Unit 5
Unit 6
Fukushima Daini NPS,
Unit 1
Unit 2
Unit 3
Unit 4
Kashiwazaki Kariwa NPS,
Unit 1
Unit 2
Unit 3
Unit 4
Unit 5
Unit 6
Unit 7
Hamaoka NPS,
Unit 1
Unit 2
Unit 3
Unit 4
Unit 5
Shika NPS,
Unit 1
Unit 2

A1-1

Reactor
Type
BWR

Power
[MWe]

1,100

Commis-si
oning
11/28/78

BWR
PWR

357
1,160

03/14/70
02/17/87

PWR
PWR

579
579

06/22/89
04/12/91

BWR
BWR
BWR
BWR

524
825
825
1,100

06/01/84
07/28/95
01/30/02
12/08/05

BWR
BWR
BWR
BWR
BWR
BWR

460
784
784
784
784
1,100

03/26/71
07/18/74
03/27/76
10/12/78
04/18/78
10/24/79

BWR
BWR
BWR
BWR

1,100
1,100
1,100
1,100

04/20/82
02/03/84
06/21/85
08/25/87

BWR
BWR
BWR
BWR
BWR
ABWR
ABWR

1,100
1,100
1,100
1,100
1,100
1,356
1,356

09/18/85
09/28/90
08/11/93
08/11/94
04/10/90
11/07/96
07/02/97

BWR
BWR
BWR
BWR
ABWR

540
840
1,100
1,137
1,267

03/17/76
11/29/78
08/28/87
09/03/93
01/18/05

BWR
ABWR

540
1,358

07/30/93
03/15/06

The Kansai
Electric Power
Co., Inc.
In
Operation

The Chugoku
Electric Power
Co., Inc.

Mihama Power Station,
Unit 1
Unit 2
Unit 3
Takahama Power Station,
Unit 1
Unit 2
Unit 3
Unit 4
Ohi Power Station,
Unit 1
Unit 2
Unit 3
Unit 4
Shimane NPS,
Unit 1
Unit 2

Ikata Power Station,
Unit 1
Shikoku Electric
Unit 2
Power Co., Inc.
Unit 3Genkai NPS,
Unit 1
Unit 2
Unit 3Kyushu Electric
Unit 4
Power Co., Inc.
Sendai NPS,
Unit 1
Unit 2
Subtotal
Hokkaido Electric Tomari Power Station
Unit3
Power Co., Inc.
Under
Construction

The Chugoku
Electric Power
Co., Inc.

Shimane NPS
Unit3

Subtotal
The Japan Atomic Tsuruga Power Station
Power Co.
Unit3
Unit4

In
Planning

PWR
PWR
PWR

340
500
826

11/28/70
07/25/72
12/01/76

PWR
PWR
PWR
PWR

826
826
870
870

11/14/74
11/14/75
01/17/85
06/05/85

PWR
PWR
PWR
PWR

1,175
1,175
1,180
1,180

03/27/79
12/05/79
12/18/91
02/02/93

BWR
BWR

460
820

03/29/74
02/10/89

PWR
PWR
PWR

566
566
890

09/30/77
03/19/82
12/15/94

PWR
PWR
PWR
PWR

559
559
1,180
1,180

10/15/75
03/30/81
03/18/94
07/25/97

PWR
PWR(55 units)

890
890
49,46707/04/84
11/28/85

PWR

912

ABWR

1,373

2009/12(Planned)
2011/12(Planned)(2 units)

2,285

APWR

1538

APWR

1538

2016/03(Planned)
2017/03(Planned)
2012/03(Planned)

Electric Power
Development Co.
LTD

Ohma NPS, Unit 1

ABWR

1,383

Tokyo ElectricPower Co., Inc.

Higashidori
Unit 1
Unit 2

ABWR
ABWR

1,385
1,385

FY2014FY2016or later

The Chugoku
Electric Power
Co., Inc.

Kaminoseki NPS,
Unit 1
Unit 2
Subtotal

ABWR
ABWR(7 units)

1,373
1,373
9,975

FY2014FY2017

A1-2

The Japan Atomic
Power Co.

Tokai Power Station

GCR

166

07/25/66
Commercial
Operation
discontinued
on
03/31/1998

Under
Decommissioning

NotificationofDismantlingon
10/04/2001(2)

Reactors at the stage of research and development
LicenseHolderIn Preparation of
Decommissioning
Under
Construction

Note:

Japan Atomic
EnergyAgencyPower Station& UnitFugen*

Monju **

Reactor
Type
ATR

FBR

Power
[MWe]

Commis-sioning

165

03/20/79

280

Commercial Operation
discontinued on
03/29/2003
Criticality on 04/05/94

In planning: Projects that were designated as sites for electric power development by the
Minister of Economy, Trade and Industry, and have not obtained construction plan approval.

* : This plant discontinued commercial operation on March 29, 2003, and is in preparation
for decommissioning.

**: These plants reached criticality and correspond to the category “reactor in operation” of
the Convention on Nuclear Safety

A1-3

This page intentionally left blankA1-4

Annex 2

Data on Nuclear Installations
2.1 Capacity of Electricity of Commercial Nuclear Power Reactors

50,00060

45,00040

30,00025,00020,00020

15,00010,000
5,000
0

0
'87

'88

'89

'90

'91

'92

'93

'94

'95

'96 '97 '98
fiscal year

Capacity of Electricity

'99

'00

'01

'02

'03

'04

'05

'06

Number of Units

2.2 Capacity Factor of Commercial Nuclear Power Reactors
100

80

60

40

20

0
'87 '88 '89 '90 '91 '92 '93 '94 '95 '96 '97 '98 '99 '00 '01 '02 '03 '04 '05 '06
fiscal year

A2-1

Number of Units

35,000Capacity Factor [%]Capacity of Electricity [MW]40,0002.3 Frequency of Unplanned Shutdown at Commercial Nuclear Power Reactors(except during commissioning)
60

3
40

2

20
1

0

0
'87 '88 '89'90 '91 '92 '93 '94 '95 '96 '97 '98 '99 '00
fiscal year

Automatic Shutdown'01 '02 '03 '04 '05 '06

Manual Shutdown

Number of Units

2.4 Reported Events (by Laws & Notifications)of Commercial Nuclear Power Reactors
100

4

Number of Reported Events

3

60

2

40

1
20

0

0
'87

'88

'89

'90

'91

'92

'93

'94

'95

'96

'97

'98

'99

'00

'01

'02

'03

'04

fiscal year
Reported Events by Laws
Number of Reported Events per unit

Reported Events by Notifications

A2-2

'05

'06

Number of Reported Events per unit

80

Number of Units

Frequency of Unplanned Shutdown
per reactor year4

2.5 Assessment of Events by INES for Commercial Nuclear Power Reactors

Number of Reported Events

30

20

10

0

'92

'93

'94

'95

'96

'97

'98

'99

'00

'01

'02

'03

'04

'05

'06

fiscal year
Out of Scale

Level 0Level 1

2.6 H um an Error Induced Events R eported
1.2

1.00

0.75
0.8
0.6

0.50

0.4
0.25
0.2
0.0

0.00
'86 '87 '88 '89 '90 '91 '92 '93 '94 '95 '96 '97 '98 '99 '00 '01 '02 '03 '04
fiscal year
N um ber of H um an Error Induced Events per unit
R atio of H um an Error Induced Events to the T otal N um ber of R eported Events

A2-3

Ratio of Human Error Failures

Number of Human Error Failures
per unit

1.0

2 .7 D o s e p e r P e r s o n s a t C o m m e r c ia l N u c le a r P o w e r R e a c to r s
5

80

4

40
2

20
1

0
'8 7

'8 8

'8 9

'9 0

'9 1

'9 2

'9 3

'9 4

'9 5

'9 6

'9 7

'9 8

'9 9

'0 0

'0 1

'0 2

'0 3

'0 4

'0 5

f is c a l y e a r
N u m b e r o f P e r s o n s E n g a g e d in R a d ia tin W o r k

D o se p e r P e rso n s

2.8 Averaged Dose at Commercial Nucear Power Reactors
8

6

4

2

0
'87 '88 '89 '90 '91 '92 '93 '94 '95 '96 '97 '98 '99 '00 '01 '02 '03 '04 '05 '06
fiscal year

A2-4

'0 6

Dose per Persons [mSv/per]

3

Dose per Unit [Person Sv/Unit]

Number of Persons [x 1,000]60

2 .9 R ad io ac tiv e G a se o u s W aste (I-1 3 1 ) R e le ased
fro m C o m m e rcial N u c ea r P o w e r R ea cto rs(N u m b er o f u n its is su m m ed fro m th eir in itia l c ritic ality .)1 .0 E + 1 2

60

1 .0 E + 1 1

1 .0 E + 0 940
Number of Units

Released Radioactive Iodine [Bq]

1 .0 E + 1 0

1 .0 E + 0 8

1 .0 E + 0 7

1 .0 E + 0 6

20

1 .0 E + 0 5

1 .0 E + 0 4

1 .0 E + 0 3

0
'8 7

'8 8

'8 9

'9 0

'9 1

'9 2

'9 3

'9 4

'9 5

'9 6

'9 7

'9 8

'9 9

'0 0

'0 1

'0 2

'0 3

'0 4

'0 5

'0 6

fisca l y ear
R elea se d R ad io ac tiv e Io d in e

N u m b er o f U n its

R ele a sed R a d io a c tiv e Io d in e p er u n it

2.10 Radioactive Liquid Waste (except H-3) Released
from Commercial Nuclear Power Reactors
60

1.0E+11
1.0E+101.0E+0940

1.0E+08
1.0E+07
1.0E+0620

1.0E+05

Number of Units

Released Radioactive Liquid Waste(per Unit) [Bq]1.0E+12

1.0E+04
1.0E+03

ND

1.0E+020'87 '88 '89 '90 '91 '92 '93 '94 '95 '96 '97 '98 '99 '00 '01 '02 '03 '04 '05 '06
fiscal year

Released Radioactive Liquid Waste

Number of Units

A2-5

Released Radioactive Liquid Waste per Unit

2 .1 1 R ad io activ e S o lid W aste G e n e ratio n p e r E le ctricity G e n e ratio n
o f C o m m e rcial N u ce ar P o w e r R e acto rs( T o tal q u an tity o f rad io activ e s o lid w aste is co n v e rte dto th e d ru m o f 2 0 0 -lite r cap acity.)
80

10

40
4

20
2

0

0
'8 7

'8 8

'8 9

'9 0

'9 1

'9 2

'9 3

'9 4

'9 5

'9 6

'9 7

'9 8

'9 9

'0 0

'0 1

'0 2

fis c a l y e a r
N u m b er o f D ru m s

N u m b er o f D ru m s p er E le c tric ity G e n e ra tio n

A2-6

'0 3

'0 4

'0 5

'0 6

[drum/MW year]

Number of Drums [x 1,000

6Number of Drums per Electricity Generation

860

Annex 3 Legislation and Guidelines(1) The Rule for the Installation, Operation, etc. of Commercial Nuclear Power Reactors (Excerpt)(Ordinance No.77 of the Ministry of International Trade and Industry, dated December 28, 1978)
Latest Revision: Ordinance of the Ministry of Economy, Trade and Industry dated June 19, 2007
In accordance with the Law on the Regulation of Nuclear Source Material, Nuclear Fuel Material and Reactors (Law No. 166,
1957) and provisions for installation, operation etc. of commercial nuclear power reactors of the Ordinance for the
Enforcement of the Law on the Regulations of Nuclear Source Material, Nuclear Fuel Material and Reactors (Ordinance No.
324, 1957), and in order to enforce the said provisions, the Rules for Installation, Operation etc. of Commercial Power
Reactors is established as follows.
(Periodic Evaluation of Nuclear Facility)
Article 15-2. The reactor establisher shall take actions provided for in the followings for every reactor and every period notexceeding 10 years in accordance with Article 35, Paragraph 1 of the Law:(i)
Evaluation of the implementing situation of fitness-for-safety activities at the nuclear facilities; and(ii)
Evaluation of the situation of reflection of state of the art technical information on fitness-for-safety activities at thenuclear facilities.
2. The licensee shall perform a technical assessment for deteriorations due to aging on essential components and structures inorder to ensure safety of the reactor facility that have been defined by the Minister of Economy, Trade and Industry(hereinafter called as “important safety-related equipment etc.”) and those described in the following by the day elapsed 30
years from the date the reactor operation started, and shall develop a ten years plan concerning actions to be taken formaintenance of the reactor facility based on the said assessment. However, it shall not apply when the condition of thedeterioration accompanying the service of the reactor facility can be adequately comprehended for portions of thecomponents and structures that have active functions:(i)
The components and structures that have functions to generate actuation signals to the engineered safety features andthe reactor shutdown system;(ii)
The components and structures that have functions for monitoring condition of the reactor facility in an accident;(iii) The components and structures that have functions to safely shut down the reactor facility from other places than thecentral control room;(iv) The components and structures that have functions to retain the reactor coolant, which are not the important safety
related equipment etc.;(v)
The components and structures that have functions to circulate the reactor coolant;(vi) The components and structures that have functions to clean-up the reactor coolant;(vii) The components and structures that have functions to store radioactive materials;(viii) The components and structures that have functions to supply electric power, which are not the important safety related
equipment etc.;(ix) The components and structures that have functions to measure and control the reactor facility (excluding thosedescribed in Item (i));(x)
The components and structures that have functions to support operation of reactor facilities;(xi) The components and structures that have functions to prevent release of fission products into the reactor coolant;(xii) The components and structures that have functions to mitigate the reactor pressure increase;(xiii) The components and structures that have functions to suppress the power escalation;(xiv) The components and structures that have functions to make up the reactor coolant; and,(xv) Essential components and structures for taking emergency measures, and the components and structures that havefunctions for monitoring an abnormal situation.
3. The licensee shall review the assessment set forth in the preceding paragraph for every period not exceeding 10 years after
the day on which the assessment and the plan set forth in the preceding paragraph were performed and developed, and shalldevelop a ten years plan concerning actions to be taken for maintenance of reactor facilities on the basis of the said review.
4. Provisions set forth in the preceding three paragraphs shall not apply to the reactor that received the license set forth in
Article 43-3-2, Paragraph 2 of the Law.
(Report on Accidents and Troubles, etc.)
Article 19-17. Pursuant to the provision of the Law Article 62-3, any reactor establisher (including previous reactor
establisher, etc. The same shall apply hereinafter in the immediately following Article and Article 24) shall, when any of thefollowing subparagraphs is applicable, report to the Minister of Economy, Trade and Industry immediately to that effect, andwithin ten (10) days on the situation and the measures against it:(i)When nuclear fuel material is stolen or its whereabouts is unknown;(ii)
When a reactor is shut down by failure of a reactor facility or when it become necessary to shut down a reactor during
operation, or when reactor power output changes more than 5%, or when reactor power output change of more than
5% is required. Except when it is one of the following and the establisher announced officially about the situation of
the concerned failure;
A. When it occurs in the term of the periodic inspection provided in Article 54-1 of the Electricity Utilities Industry
Law (Law No. 170, 1964) (limited to those equipment concerned to the failure, which functional and operational-- Provisional translation for information only --A3-1

situation cannot be checked under the reactor shutdown condition),
B. When the failure does not cause deviation from the limit of operation (it is a requirement defined in the operationalsafety program for operation of the reactor facility, and when it cause any deviation from the concerned
conditions the measure that the establisher should take is also defined in the operational safety program, the samein this paragraph), and there is no change observed related with the concerned failure, and when the establisher
performs inspection of the failed equipment concerned, or
C. When the reactor output is required to follow the limit of operation.
(iii) When a reactor establisher has checked the equipment etc. important to the safety, and when the equipment etc.
important to safety is considered that it does not satisfy the standard described in Article 9 or in Article 9-2 of theOrder of Establishing Technical Standards for Nuclear Power Generation Equipment (Order No. 62 of Ministry ofInternational Trade and Industry, 1965), or when it is considered that it does not have function to secure safety of thenuclear reactor facility;(iv) When there is a failure of equipment etc. important to safety by the fire. Excluding the concerned failure is due to themeasure of fire extinguishing or prevention of the spread of fire;(v)
Except for the preceding three subparagraphs, when deviation from the limit of operation by the failure of a thenuclear reactor facility (except those minor troubles which impact on operation of nuclear reactor is insignificant) iscaused, or when the measure for the concerned deviation defined in the operational safety program is not implemented
at the time of deviation from the limit of operation;(vi) When the failure of a reactor facility or occurrence of other unexpected situation is considered to have caused any
trouble in the situation of discharge of gaseous radioactive wastes through the ventilation facility or in the situation ofdischarge of liquid radioactive wastes through the drainage facility;(vii) When the concentration of radioactive materials in the air outside the environment monitoring area exceeds theallowable limit in the case of discharge of gaseous radioactive wastes through the ventilation facility;(viii) When the concentration of radioactive materials in the water outside the environment monitoring area exceeds theallowable limit in the case of the discharge of liquid radioactive wastes through the drainage facility;(ix) When nuclear fuel materials or materials contaminated with nuclear fuel materials (hereinafter referred to as "nuclear
fuel materials etc.") leak out of the control zone;(x)When nuclear fuel materials etc. leak within the control zone due to failure of a nuclear reactor facility or occurrenceof other unexpected situation. However, this is not the case in any of the followings (except the case when new
measures such as human entry restriction into the leakage-related place and key control have been taken or when theleaked substances have spread outside the control zone);
A. When revealed liquid nuclear fuel materials etc. do not spread out of the floodgate that is installed in thecircumference of the equipment of the concerned leakage for prevention of leakage enlargement,
B. When the function of ventilation facility of the concerned area of the leakage is maintained properly at the time
when gaseous nuclear fuel materials etc. leak, or
C. When the amount of radioactivity of the leaked nuclear fuel materials etc. is very little and when the degree of theleakage is minor.
(xi) When the person who enters into the control zone suffered radiation exposure due to the failure of a nuclear reactor
facility or occurrence of other unexpected situation, and when the effective dosage of concerned exposure exceeds or
could exceeds five mSv for a personnel engaged in radiation work or 0.5 mSv for a person other than the personnel
engaged in radiation work;(xii) When the dosage of personnel engaged in radiation work exceeds or could exceed the allowable dose limit; or(xiii) When a control rod that is not actually operating for insertion into or withdrawal from the original control rod position(it is one of control rod positions that the procedure for operating control rods, which is established by the licensee
based on the Operational Safety Program, defines to set up based on a certain interval in order to control the control
rod, the same shall apply hereinafter) moves to another control rod position or moves passing through the said anothercontrol rod position, or when the control rod that is at the full insertion position (it is a control rod position where thecontrol rod is to be inserted to the maximum extent, the same shall apply hereinafter) which is not actually operating
for insertion or withdrawal moves passing through the full insertion position into the further insertion direction.
However, this is not the case when fuels are not loaded in the core.
(xiv) Other than those above subparagraphs, when persons are injured or could be injured in the nuclear reactor facility(except for what hospitalization medical treatment is not needed for other than radiation hazard).

(2) Ordinance of the Ministry for Establishing Technical Standards for Nuclear Power Generation Systems(Ordinance of the Ministry of International Trade and Industry No. 62, June 15, 1965)
Latest Revision: Ordinance No.121 of the Ministry of Economy, Trade and Industry dated December 22, 2005

Pursuant to the provision of Article 48, paragraph 1 of the Electric Utilities Industry Law (Act No. 170, 1964) (including thecases when Article 74, paragraph 2 is applied mutatis mutandis), the Ordinance of the Ministry for Establishing TechnicalStandards for Nuclear Power Generation Systems was enacted as follows.
Article 1 (Scope of Application)
This Ordinance of the Ministry applies to the electric structures to be established for electricity production by utilizing
nuclear energy as its energy source.
-- Provisional translation for information only --A3-2

Article 2 (Definitions)
In this Ordinance of the Ministry, the meanings of the terms listed in the following items shall be as prescribed respectively inthose items:(i) The term "Radiation" means the radiation specified in Article 3, item 5 of the Atomic Energy Basic Act (Act No. 186,
1955) or electron beams or X-rays with energy less than one mega electron volts (eV) excluding natural radiation;(ii) The term "Reactor Facility" means a reactor and its associated systems;(iii) The term "Primary Coolant" means the fluid of which main function is to take out the heat generated in reactor coredirectly from the reactor;(iv) The term "Secondary Coolant" means the fluid of which main function is to take out the heat of the Primary Coolant
using heat exchangers with the main purpose to drive turbine;(v) The term "Primary Cooling System" means a circuit through which the Primary Coolant circulates;(vi) The term "Abnormal Transient During Operation" means the abnormal condition due to a single failure or malfunction ofequipment, or a single error of operator’s action and external disturbances expected to occur with a similar frequency to theseduring operation of the Reactor Facility;(vii) The term "Engineered Safety Features" means facilities with functions to suppress or prevent the potential release of alarge amount of radioactive materials due to failure of fuel in a reactor, etc. resulted from a failure, damage etc. of theReactor Facility;(viii) The term "Safety Systems" means systems described in the followings, which failure, damage etc. directly or indirectly
could cause possible Radiation hazards to the public;(a) Systems related to the Primary Cooling System and their associated systems,(b) Reactivity control system (system to adjust the reactivity during normal operation; hereinafter the same shall apply),and facilities related to reactor shutdown system (system to bring a reactor to subcritical condition and shut it down
to maintain the non-critical state; hereinafter the same shall apply) and their associated systems,(c) Safety protection device (it means devices to actuate the reactor shutdown system automatically in such cases as anabnormal transient during operation, a difficulty in reactor operation due to an earthquake etc., or a loss of primarycoolant accident, and the Engineered Safety Features automatically in the case of possible release of a large amountof radioactive materials due to failure of fuel in a reactor etc.; hereinafter the same shall apply), the emergency corecooling system (it means systems to remove the heat generated in the reactor pressure vessel in the case of thefunctional loss of the facility to remove the heat generated in the reactor pressure vessel during normal operation;
hereinafter the same shall apply) and other systems required to ensure safety of the reactor in an emergency andtheir associated systems,(d) Reactor containment and their isolating valves, and(e) Emergency power supply systems and their associated systems.
(ix) The term "Controlled Area" means the area within a nuclear power plant where there is a hazard that the dose fromexternal Radiation would exceed the limit specified separately by a Public Notice, and that the concentration of radioactive
material in the air (excluding those included in the natural air or water; hereinafter the same shall apply) would exceed thelevel specified separately by a Public Notice or that the density of radioactive material on the surface of objects contaminatedby radioactive materials would exceed the level specified separately by a Public Notice;(x) The term "Peripheral Monitoring Area" means the area surrounding the Controlled Area, the outside of which there would
be no possibility of the dose exceeding the limit of the dose specified separately by a Public Notice;(xi) The term "Reactor Coolant Pressure Boundary" means the part that becomes a pressure barrier by automatic closure ofvalves actuated by a damage etc. of the facilities related to the Primary Cooling System;(xii) The term "Allowable Fuel Damage Limit" means a limit on the extent of fuel cladding tube damage, which extent is
acceptable by the safety design and enables the reactor to operate safely;(xiii) The term "Reactivity Worth" means a variation in reactivity of a reactor produced by insertion or withdrawal of control
rods, or injection of liquid control materials, etc.;(xiv) The term "Maximum Reactivity Worth of A Control Rod" means the maximum value of the Reactivity Worth produced
in a reactor core by withdrawal of one control rod, when the reactor is in a critical condition (including near-to-criticalcondition);(xv) The term "Reactivity Insertion Rate" means the amount of reactivity added in a core per unit time by withdrawal ofcontrol rods, etc.;(xvi) The terms "Class 1 Vessel", "Class 1 Piping", "Class 1 Pump" or "Class 1 Valve" (hereinafter referred to as "Class 1
Component") mean the components that compose the Reactor Coolant Pressure Boundary,(xvii) The terms "Class 2 Vessel ", "Class 2 Piping", "Class 2 Pump" or "Class 2 Valve" (hereinafter referred to as "Class 2
Component") mean the following components;(a) The components belonging to systems necessary to shutdown the reactor safely or to ensure safety in an emergency,
which failure, damage etc. would indirectly cause Radiation hazards to the public (for ducts belonging to theRadiation management systems, only their parts from the penetrations of the Reactor Containment to the outboard
isolating valves are Class 2 Components),(b) The component belonging to systems of the circuit through which fluids (i.e., steam and feed water) circulates
mainly for the purpose to drive a turbine, located between the Class 1 Component and its nearest stop valve in thesteam system located in the downstream of the Class 1 Component, and located between the Class 1 Component
and its nearest stop valve in the feet water system located in the upstream of the Class 1 Component, and(c) The component other than that listed in the items 1 and 2 located between a penetration of the Reactor Containment
and an inboard isolation valve or outboard isolating valve,-- Provisional translation for information only --A3-3

(xviii) The terms "Class 3 Vessel", "Class 3 Piping", (hereinafter referred to as "Class 3 Component") mean the vessel orpiping other than the Class 1 Component, Class 2 Component, Reactor Containment and the duct belonging to the Radiation
management systems (limited to the piping which contains a fluid of radioactive material concentration of 37 mBq/cm3 (forliquid fluid, 37 kBq/cm3) or more, or the piping where its Maximum Operating Pressure exceeds zero MPa);(xix) The term "Class 4 Piping" means the duct belonging to the Radiation management system, which contains a fluid ofradioactive material concentration of 37 mBq/cm3 or more (excluding the part belonging to the Class 2 Piping);(xx) The term " Reactor Containment" means a vessel provided to prevent a leak of hazardous materials such as radioactive
materials that are released from the mechanical equipment in the vessel;(xxi) The term "Concrete Reactor Containment" means the Reactor Containment that has the steel-plate lined Concrete Part;(xxii) The term "Concrete Part" means the part of reinforced-concrete structure or pre-stressed-concrete structure of theConcrete Reactor Containment;(xxiii) The term " Steel-Plate Lined Part etc." means the steel plate that is lined on the Concrete Part (hereinafter referred toas "Liner Plate"), the steel plate to connect the Liner Plates of the shell and the bottom (hereinafter referred to as "Knuckle"),the penetration sleeve, and the metal fitting to the Concrete Part to prevent a leak of hazardous materials such as radioactive
materials released from the mechanical equipment in the Concrete Reactor Containment;(xxiv) The terms "Class 1 Support Structure", "Class 2 Support Structure" or " Reactor Containment Support Structure" mean
the structures that support the Class 1 Component, Class 2 Component or Reactor Containment, respectively;(xxv) The term "Operating Condition I" means the normal operation condition of a Reactor Facility;(xxvi) The term "Operating Condition II" means the condition other than the Operating Condition I, Operating Condition III,Operating Condition IV, and the Testing Condition;(xxvii) The term "Operating Condition III" means the condition that a shutdown of reactor operation is urgently required due
to a failure or malfunction, etc. of the reactor facility;(xxviii) The term "Operating Condition IV" means the condition that the abnormal condition assumed in the safety design of
the Reactor Facility exists;(xxix) The term "Testing Condition" means the condition that a pressure exceeding the Maximum Operating Pressure isapplied to the Reactor Facility for a pressure test;(xxx) The term "Load Condition I" means the condition where the Concrete Reactor Containment receives the load assumed
in the Operating Condition I (excluding the condition during snow accumulation and storm.);(xxxi) The term "Load Condition II" means the condition where the Concrete Reactor Containment receives the load assumed
in either of the following conditions;(a) The condition during SRV actuation (excluding the condition during snow accumulation and storm)(b) The condition during a pressure test of the Reactor Containment (excluding the condition during snow
accumulation and storm), and(c) The condition during snow accumulation in the Operating Condition I (excluding the condition during storm).
(xxxii) The term "Load Condition III" means the condition of Concrete Reactor Containment other than the condition duringstorm in the Operating Condition I or the Load Condition IV in the Operating Condition IV;(xxxiii) The term "Load Condition IV" means the condition that an abnormal event assumed for the safety of the Reactor
Containment exists in the Concrete Reactor Containment during the Operating Condition IV (including the time during snow
accumulation or storm);(xxxiv) The term "Maximum Operating Pressure" means the pressure equal to or greater than the highest pressure to bereceived by the subject component or core support structure under the operating condition in which their main functions
should be achieved and defined by the design;(xxxv) The term "Maximum Operating Temperature" means the temperature equal to or greater than the highest temperaturearising to the subject component, support structure or core support structure under the operating condition in which theirmain functions should be achieved and defined by the design;(xxxvi) The term "Minimum Operating Temperature" means the temperature equal to or less than the lowest temperaturearising to the subject component, support structure or core support structure under the operating condition or Testing
Condition in which their main functions should be achieved and defined by the design; and(xxxvii) The term "Mechanical Load" means the load by the dead weight, reactive force by piping or support structure, andother added loads, but excluding the seismic load and defined by the design.
Article 3 (Facility by Special Design)
When approved by the Minister of Economy Trade and Industry for a special ground, reactors, steam turbines, and theassociated systems can be established not pursuant to the provisions of this Ordinance of the Ministry.
2 Entities intending to obtain an approval set forth in the preceding paragraph shall apply with the written application
attached with related drawings, which describes the ground and the establishing method.
Article 4 (Protection Measures etc.)
When the Reactor Facility or the steam turbine driven by the Reactor Coolant or Secondary Coolant or its associated systems
would impair the safety of the reactor due to the assumed natural phenomena (land sliding, fault, snow slide, flooding, tidal
wave, high tide, uneven subsidence of the foundation ground, etc., however, excluding an earthquake), protection measures,
improvement of the foundation ground or other appropriate measures shall be taken.
(2) When there is a place of business, railway, road etc. in the region adjacent to the Peripheral Monitoring Area, protection
measures and other appropriate measures shall be taken in order not to impair the reactor safety by fire or explosion accident
at the place of business, or an accident etc. of a vehicle carrying dangerous goods.
(3) When crash of an airplane would impair the reactor safety, protection measures and other appropriate measures shall betaken.
Article 4-2 (Prevention of Damage by Fire)-- Provisional translation for information only --A3-4

The Reactor Facility, steam turbine or its associated systems shall be provided with the measure appropriately combining themeasures listed in the following each item in order not to impair the reactor safety by fire:(i) In order to prevent an outbreak of fire, the following measures shall be taken;(a) Leak prevention and other measures for the system containing incendiary or inflammable materials shall be taken,(b) Noncombustible or flame retarding materials shall be used for the cable, reactor control room and other associatedsystems of the reactor according to the amount of combustible material, etc.,(c) Lightning protection systems etc. to prevent the outbreak of a fire due to lightning strike or other natural
phenomena shall be provided,(d) The hydrogen supply facility etc. shall be installed such that it would not impair the reactor safety even in theevent of hydrogen combustion, and(e) When rapid combustion of hydrogen generated and accumulated due to radiolytic decomposition would impair thereactor safety, preventive measures against buildup of hydrogen shall be taken.
(ii) The following measures for detection and extinguishing of fire shall be taken;(a) Fire detection and extinguishing systems that can extinguish the fire at an early stage shall be provided, and(b) Capabilities of the fire detection and extinguishing systems provided in the above (a) shall not be impaired by thenatural phenomena expected to occur simultaneously with the fire.
(iii) Firewalls and other measures shall be taken to mitigate the consequence of a fire.
(2) Fire detection and extinguishing systems provided in Item 2, (a) of the preceding paragraph shall be those not to impair
the functions of the Safety Systems due to a failure, damage, inadvertent actuation etc.
Article 5 (Earthquake Resistance)
The Reactor Facility or steam turbine driven by the Reactor Coolant or Secondary Coolant or its associated systems shall be
provided not to cause Radiation hazards to the public due to the damage by the seismic force working on these.
(2) The seismic force described in the preceding paragraph shall be derived based on the condition of the foundation ground,and the extent of the earthquake damage and situation of seismic activities, etc. based on the past seismic records in theregion, corresponding to the structure of the Reactor Facility or steam turbine driven by the Reactor Coolant or Secondary
Coolant or its associated systems and the extent of the disaster when these are damaged.
Article 6 (Prevention of Damage by Flow Induced Vibration etc.)
The fuel assembly and reflector, the structures supporting these components, the thermal shield, and the vessel, piping, pumpand valve belonging to the systems related to the Primary Cooling System shall be provided so as not to suffer a damage by
the flow-induced vibration due to circulation, boiling etc. of the Reactor Coolant or Secondary Coolant or the temperaturefluctuation due to mixing of fluids with a temperature difference, etc.
Article 7 (Facilities such as Fences)
The Controlled Area of a nuclear power plant shall be provided with walls, fences, barriers, etc. not to allow free access tothe area, and it shall be indicated that the area is the Controlled Area.
(2) The boundary between the access-controlled areas (an area that especially requires control for maintaining the integrity of
the Reactor Facility other than the Controlled Area; hereinafter the same shall apply) and the Controlled Areas, and otherareas of a nuclear power plant shall be provided with boundaries, fences etc. in order to distinguish them from other areas, or
it shall be indicated that the area is the access-controlled area.
(3) The Peripheral Monitoring Area of a nuclear power plant shall be provided with boundaries, fences etc. in order to restrict
unnecessary access to the area except those who have authorized access for job-related reasons, or it shall be indicated that itis the Peripheral Monitoring Area. However, this shall not apply when it is obvious that there would be no possibility ofaccess to the area concerned.
Article 7-2 (Prevention of Unlawful Entry)
A nuclear power plant shall be provided with appropriate intrusion preventive measures to a facility where the Safety
Systems are provided with in order to prevent unlawful intrusions.
Article 7-3 (Prevention of a Collapse of Steep Sloping Ground)
The electric structures to be constructed at the designated dangerous area for a collapse of steep sloping ground pursuant to
the provision of Article 3, paragraph 1 of the Act for Preventing a Disaster by Collapse of a Steep Sloping Ground (1969, Act
No. 57) shall be installed in such a manner that they would not contribute to or induce a collapse of the steep sloping groundin the area concerned (a sloping ground specified in Article 2, paragraph 1 of the sama Act).
Article 8 (Reactor Facility)
The Reactor Facility shall have the capability to control the chain reaction of the nuclear fission by controlling the reactivityof the reactor as well as the capability to control the reactivity of the reactor safely and stably during its normal operation andthe power suppression characteristics inherent to the reactor even in the event of an Abnormal Transient During Operation.
(2) The Reactor Facility (excluding an auxiliary boiler) shall be provided in such a manner that the maintenance andinspection (including tests and inspections) of necessary parts can be carried out during operation or shutdown of the reactor
in order to confirm its integrity and capability.
(3) The Reactor Facility shall be provided in such a manner that, in the event of significant leakage of the fluid containing
radioactive materials from a vessel, piping, pump, valve, or other mechanical equipment during normal operation, the leakage
could be processed safely by the system to process the liquid radioactive waste.
(4) A system belonging to the Reactor Facility, assumed to be damaged by a missile following a damage of a steam turbine,pump etc. and then impair safety of the Reactor Facility shall be provided with damage preventive measures such as
installation of protection component or others(5) When a system belonging to the Reactor Facility is shared with other Reactor Facility, it shall be provided so that thereactor safety would not be compromised.
Article 8-2 (Safety Systems)-- Provisional translation for information only --A3-5

The Safety Systems listed in Article 2, item 8, (c) and (e) shall be provided such that they have redundancy or diversity, andindependency taking into consideration of the function, structure and principle of operation of their mechanical equipment inorder to enable them to function even when the external power is not available in the event of a single failure (it means that
one mechanical equipment loses its intended safety function due to a single cause; hereinafter the same shall apply) of themechanical equipment composing the Safety Systems concerned.
(2) Safety Systems shall be provided such that those can accomplish their function under all environmental conditions
assumed.
Article 9 (Material and Structure)
The materials and structures of the vessel, piping, pump or valve (hereinafter referred to as "component") that belongs to theReactor Facility (excluding a compressor and an auxiliary boiler,) or their support structures, or core support structures shallbe in accordance with the following each item, and, in this case, the provisions from Items (i) to (vii) and Item (xv) shall beapplied prior to their use:(i) Materials to be used for the Class 1 Component and Class 1 Support Structure shall be pursuant to the followings;(a) The Class 1 Component or Class 1 Support Structure shall have an appropriate mechanical strength and chemicalcomposition (including appropriate corrosion resistance under the in-service stress etc.) for its in-service
conditions such as pressure, temperature, water quality, Radiation and load,(b) The material to be used for the Class 1 Vessel shall be one confirmed by a mechanical test etc. to have anappropriate fracture toughness under the conditions for the vessel concerned to be used, such as pressure,temperature, Radiation and load,(c) The material to be used for the Class 1 Component (excluding the Class 1 Vessel) or Class 1 Support Structure(excluding the support structure for the Class 1 Piping and valve) shall be one confirmed by a mechanical test etc.
to have appropriate fracture toughness under the Minimum Operating Temperature of the component or support
structure concerned, and(d) The material to be used for the Class 1 Component or Class 1 Support Structure (limited to the rod and bolt) shallbe one confirmed by nondestructive examinations to have no harmful flaw.
(ii) Materials to be used for the Class 2 Component and Class 2 Support Structure shall be pursuant to the followings;(a) The Class 2 Component or Class 2 Support Structure shall have appropriate mechanical strength and chemicalcomposition for its in-service conditions such as pressure, temperature and load,(b) The material to be used for the Class 2 Component shall be one confirmed by a mechanical test etc. to haveappropriate fracture toughness under the Minimum Operating Temperature of the component concerned, and(c) The casting belonging to the Class 2 Component shall be one confirmed by nondestructive examinations to have
no harmful flaw.
(iii) Materials to be used for the Class 3 Component shall be pursuant to the followings;(a) The Class 3 Component shall have an appropriate mechanical strength and chemical composition for its in-service
conditions such as pressure, temperature and load, and(b) The material to be used for the Class 4 component belonging to the Engineered Safety Features shall be one
confirmed by a mechanical test etc. to have appropriate fracture toughness under the Minimum Operating
Temperature of the component concerned.
(iv) Materials to be used for the Class 4 Piping shall have an appropriate mechanical strength and chemical composition for
its in-service conditions such as pressure, temperature and load;(v) Materials to be used for the Reactor Containment (excluding the Concrete Reactor Containment; hereinafter the same
shall apply in this item) and the Reactor Containment Support Structure shall be pursuant to the followings;(a) The Reactor Containment or Reactor Containment support structure shall have an appropriate mechanical strengthand chemical composition for its in-service conditions such as pressure, temperature, humidity and load, and(b) The material confirmed by a mechanical test etc. to have appropriate fracture toughness under the Minimum
Operating Temperature for the Reactor Containment or Reactor Containment support structure.
(vi) Materials to be used for the concrete and steel lining parts, etc. of the Concrete Reactor Containment shall be pursuant to
the followings;(a) The concrete shall have an appropriate compressive strength under the conditions such as pressure, temperatureand load under which the Reactor Containment concerned is to be used,(b) The concrete shall have long-term durability not to cause the harmful expansion and the corrosion of thereinforcement rebar,(c) The reinforcement rebar, tendon and fixing device (hereinafter referred to as "reinforcement rebar etc.") to be
used as a structural member for strength of the Concrete Part shall have an appropriate mechanical strength,chemical composition and geometry/dimension for its in-service conditions such as pressure, temperature andload of the Reactor Containment concerned, and(d) The provisions of (a) and (b) of the preceding item shall apply mutatis mutandis to the materials to be used for the
Steel Lining Part etc.
(vii) The provisions of item 1, (a) and (c) shall apply mutatis mutandis to the materials to be used for a core support structure;(viii) The structure and strength of Class 1 Component and Class 1 Support Structure shall be pursuant to the followings;(a) The overall strain of the Class 1 Component shall be kept within the elastic range under the conditions that the
Maximum Operating Temperature, Maximum Operating Pressure and mechanical load are applied (hereinafterreferred to as "conditions defined by the design"),(b) The overall strain of the Class 1 Support Structure shall be kept within the elastic range under the Operating
Condition I and the Operating Condition II,-- Provisional translation for information only --A3-6

(c) The Class 1 Vessel (excluding the omega seal etc.), Class 1 Piping, Class 1 Valve and Class 1 Support Structure
shall not produce an overall plastic strain under the Operating Condition III. However, this clause shall not apply
to a local plastic strain at structural discontinuities,(d) The Class 1 Vessel (excluding the omega seal etc.), Class 1 Piping and Class 1 Support Structure shall not produce
a plastic strain that results in a ductile fracture under the Operating Condition IV,(e) The Class 1 Vessel (excluding the bolt, omega seal etc.) shall not produce an overall plastic strain under the
Testing Condition. However, this clause shall not apply to a local plastic strain at structural discontinuities,(f) The Class 1 Vessel (excluding the bolt etc.), Class 1 Piping, Class 1 Valve (limited to the valve body) and Class 1
Support Structure shall not produce a progressive strain under the Operating Condition I and Operating Condition
II,(g) The Class 1 Vessel, Class 1 Piping, Class 1 Valve (limited to the valve body) and Class 1 Support Structure shallnot produce a fatigue fracture under the Operating Condition I and Operating Condition II,(h) The Class 1 Vessel (limited to the shell, end plate etc.) shall not buckle under the Operating Condition I, Operating
Condition II, Operating Condition III and Operating Condition IV and Testing Condition,(i) The Class 1 Piping shall not buckle under the conditions defined by the design,(j) The Class 1 Support Structure shall not buckle under the Operating Condition I, Operating Condition II, Operating
Condition III and Operating Condition IV, and(k) Notwithstanding the provisions of the preceding (b), (c), (d), (f), (g) and (k), the provisions for the Class 1 Vessel
shall apply mutatis mutandis to the Class 1 Support Structure welded to a Class 1 Vessel, which failure would
damage the Class 1 Vessel,(ix) The structure and strength of Class 2 Component and Class 2 Support Structure shall be pursuant to the followings;(a) The overall strain of the Class 2 Component shall be kept within the elastic range under the conditions defined by
the design,(b) The expansion joint belonging to the Class 2 Component shall not produce a fatigue fracture under repeated stressadded under the conditions defined by the design,(c) The Class 2 Piping (excluding the expansion joint) shall not produce a fatigue fracture under the Operating
Condition I and Operating Condition II,(d) The Class 2 Vessel and Class 2 Piping shall not buckle under the conditions defined by the design, and(e) The Class 2 Support Structure welded to the Class 2 Component, which failure would damage the Class 2
Component, shall not produce a ductile fracture or buckle under the Operating Condition I and Operating
Condition II.
(x) The structure and strength of Class 3 Component shall be pursuant to the followings;(a) The overall strain shall be kept within the elastic range under the conditions defined by the design,(b) The expansion joint belonging to the Class 3 Component shall not produce a fatigue fracture under repeated stressadded under the conditions defined by the design, and(c) A buckling shall not occur under the conditions defined by the design.
(xi) The structure and strength of the Class 4 Piping shall not produce a plastic strain that results in a ductile fracture under
the conditions defined by the design;(xii) The structure and strength of the Reactor Containment (excluding the Concrete Reactor Containment) and Reactor
Containment Support Structure shall be pursuant to the followings;(a) The overall strain of the Reactor Containment (excluding the parts described in (b)) shall be kept within the elastic
range under the conditions defined by the design,(b) The provisions of item 8, (a), (c), (d) and (e) for the Class 1 Vessel shall apply mutatis mutandis to the extremelystressed and specially formed parts of the Reactor Containment,(c) The provisions of item 8, (b), (c) and (d) for the Class 1 Support Structure hall apply mutatis mutandis to the
support structure of the Reactor Containment,(d) The extremely stressed and specially formed parts and the support structure of the Reactor Containment shall not
break due to progressive strain under the Operating Condition I and Operating Condition II,(e) The expansion joint of the Reactor Containment shall not produce a fatigue fracture under repeated stress added
under the conditions defined by the design case,(f) The extremely stressed and specially formed parts and the support structure of the Reactor Containment shall not
produce a fatigue fracture under the Operating Condition I and Operating Condition II,(g) The Reactor Containment shall not buckle under the conditions defined by the design and the Operating Condition
III and Operating Condition IV, and(h) The support structure of the Reactor Containment shall not buckle under the conditions defined by the design andthe Operating Condition I, Operating Condition II, Operating Condition III and Operating Condition IV.
(xiii) The structure and strength of the Concrete Reactor Containment shall be pursuant to the followings;(a) The concrete shall not suffer a compression fracture under the Load Condition I, Load Condition II and Load
Condition III, and, the Concrete Reactor Containment shall not suffer a compression fracture leading to a large
plastic strain under the Load Condition IV,(b) The reinforcement rebar etc. shall not yield under the Load Condition I, Load Condition II, and Load Condition
III and produce a strain leading to a rapture under the Load Condition IV,(c) The Concrete Part shall not suffer a shear fracture under the Load Condition I, Load Condition II and Load
Condition III, and the Concrete Reactor Containment shall not suffer a shear fracture leading to a large plasticstrain under the Load Condition IV,-- Provisional translation for information only --

A3-7

(d)

The Liner Plate (excluding the part to which a penetration sleeve is installed) shall not produce an extreme
residual strain under the Load Condition I and Load Condition II, and shall not result in a rapture under the Load
Condition III and Load Condition IV,(e) The provision of item 12 for the Reactor Containment in addition to the provision of (d) shall apply mutatis
mutandis to the Liner Plate (excluding the part to which a penetration sleeve is installed),(f) The provisions of item 12, (c), (d), (f) and (h) for the Reactor Containment Support Structure shall apply mutatis
mutandis to the Liner Plate (limited to the part to which a penetration sleeve is installed), penetration sleeve andfixing metal fittings (excluding fixing metal fittings to be attached to a Liner Plate, which overall strain can be
kept within the elastic range under any load conditions). In this case, the terms "Operating Condition I andOperating Condition II" of item 12 shall be deemed to be replaced with the "Load Condition I and Load
Condition II", and the terms "Operating Condition I, Operating Condition II, Operating Condition III andOperating Condition IV" with the "Load Condition I, Load Condition II, Load Condition III, and Load Condition
IV", and(g) The provisions of item 12, (b), (d) and (f) for the extremely stressed and specially formed parts of the Reactor
Containment shall apply mutatis mutandis to the Knuckle.
(xiv) The structure and strength of a core support structure shall be pursuant to the followings;(a) An overall strain shall be kept within the elastic range under the conditions defined by the design,(b) An overall plastic strain shall not be produced under the Operating Condition III. However, this clause does not
apply to a local plastic strain at structural discontinuities,(c) The plastic strain leading to a ductility fracture shall not be produced under the Operating Condition IV,(d) The core support structure shall not be destroyed due to progressive strain under the Operating Condition I andOperating Condition II,(e) A fatigue fracture shall not be produced under the Operating Condition I and Operating Condition II, and(f) A buckling shall not occur under the Operating Condition I, Operating Condition II, Operating Condition III andOperating Condition IV.
(xv) The weld (weld metal and heat affected zone) of major pressure parts of the Class 1 Vessel, Class 1 Piping, Class 2
Vessel, Class 2 Piping, Class 3 Vessel, Class 3 Piping, Class 4 Piping and Reactor Containment shall be pursuant to thefollowings;(a) It shall not be a discontinuous and unusual shape,(b) The weld would not produce a crack due to welding, and shall be one confirmed by nondestructive examinations
not to have incomplete penetrations and other defects harmful for ensuring a sound weld,(c) It shall have an appropriate strength, and(d) It shall be welded with the welding procedure etc. confirmed to be appropriate one beforehand by mechanicaltests etc.
Article 9-2 (Prevention of Destruction during Service due to a Crack etc.)
The Class 1 Component, Class 1 Support Structure, Class 2 Component, Class 2 Support Structure, Class 4 component, Class
4 Piping, Reactor Containment Support Structure of the Reactor Containment and core support structure in use shall not haveany cracks and other defects that could cause their destruction.
(2) The pressure part of the Class 1 Component in use shall not have any cracks and other defects that penetrate the pressure
part.
Article 10 (Safety Valve etc.)
The Reactor Facility shall be provided with the safety valve or relief valve (hereinafter referred to as "safety valve etc.";
hereinafter the same shall apply in this Article) pursuant to the following each item:(i) The safety valve etc. shall have a mechanism that functions without fail;(ii) The valve stem of safety valve etc. shall have a structure that can prevent the leak from the valve seat surface
appropriately;(iii) Materials of the safety valve etc. shall be pursuant to the followings;(a) The provision of Article 9, item 1 shall apply mutatis mutandis to the materials of safety valve etc. attached to theClass 1 Vessel and Class 1 Piping, and(b) The provisions of Article 9, item 2 shall apply mutatis mutandis to the materials of safety valve etc. attached to theClass 2 Vessel and Class 2 Piping.
(iv) The safety valve etc. with an auxiliary actuation device shall have a structure that can have a required blow-down
capacity even in the event of a failure of the auxiliary actuation device concerned;(v)
The reactor pressure vessel (pressurizer when it is provided; hereinafter the same shall apply in this item) shall be
pursuant to the followings;(a) Two or more safety valves with bellows (referred to as "safety valve with bellows" in item 7) shall be provided at
appropriate locations to prevent malfunction of its actuation due to the effect of backpressure, and(b) The total of safety valve capacity shall be equal to or greater than the capacity required to prevent overpressure of
the reactor pressure vessel concerned by appropriate combination of the blow-off pressure and the installednumber of the safety valve concerned. However, for the reactor pressure vessel provided with the device to have
overpressure preventive effects other than the safety valve, the value equivalent to the overpressure preventive
capability of the device concerned could be subtracted.
(vi) The steam generator shall be pursuant to the followings;(a) Two or more safety valves shall be provided at appropriate locations,-- Provisional translation for information only --A3-8

(b) The total capacity of safety valves shall be equal to or greater than the capacity required to prevent overpressure of
the steam generator concerned by appropriate combination of the blow-off pressure and the installed number ofthe safety valve concerned, and(c) The safety valve shall cease to blow-off promptly after the pressure goes down below its blow-off pressure.
(vii) The piping provided with the pressure reducing valve, of which low pressure portion or the component connected to it
is not designed so as to withstand the pressure of the high pressure portion, shall be pursuant to the followings;(a) For the Class I piping, two or more safety valve with bellows shall be provided closely to the low-pressure side of
the pressure-reducing valve,(b) For piping other than that described in (a), one or more safety valve etc. shall be provided closely to the
low-pressure side of the pressure-reducing valve,(c) The total capacity of safety valves etc. shall be equal to or greater than the capacity required to prevent
overpressure of the low pressure side of the piping or the components connected to it when the pressure reducing
valve is fully open by appropriate combination of the blow-off pressure and the installed number of the safetyvalve concerned etc., and(d) The safety valve shall cease to blow-off promptly after the pressure goes down below its blow-off pressure.
(viii) The vessel (excluding those described in item 5, item 6 and paragraph 3, an auxiliary boiler and the Reactor
Containment) or piping (excluding those described in the preceding item) belonging to the Reactor Facility that would beover-pressurized inside shall be provided with safety valves etc. at appropriate locations in the same manner as the provisionsof item 6, (b) and (a), (b) and (d) of the preceding item;(2) In the case of the preceding paragraph, when a rapture disk is provided at the inlet-side or outlet-side of the safety valveetc., it shall be provided pursuant to the following each item:(i) When provided at the inlet-side of the safety valve etc., it shall be provided pursuant to the followings;(a) The blow-off pressure of rapture disk shall not be more than the Maximum Operating Pressure of the vessel
concerned, and(b) Destruction of the rapture disk shall not compromise the function of the safety valve etc.
(ii) When provided at the outlet-side of the safety valve etc., it shall be provided pursuant to the followings;(a) The rapture disk shall break at a low pressure not to prevent actuation of the safety valve etc.,(b) The pressure that is the sum of the blow-off pressure of a rapture disk and that of the safety valve etc. shall belower than a blow-off pressure required to prevent overpressure,(c) The support mechanism of a rapture disk shall be such that the passage area for fluid discharge is equal to or morethan the discharge area of the safety valve etc., and(d) Destruction of a rapture disk shall not compromise the function of the discharge piping.
(3) The vessel which belongs to the Reactor Facility and contains materials such as liquid carbon dioxide gas etc. that would
disable actuation of safety valve etc., shall be provided with rapture disks pursuant to the following each item;(i) One or more rapture disk shall be installed at appropriate locations to obtain the capacity required for overpressure
prevention of the vessel concerned by appropriately combining the blow-off pressure and the number of the installed rapture
disk, and(ii) The cross section of the connecting piping of the vessel and the rapture disk shall be equal to or more than the crosssection of the rapture disk.
(4) If stop valve is to be provided at the inlet or outlet side of the safety valve etc. or rapture disk for the case of paragraph 1or the preceding paragraph, a device capable to confirm that the stop valve is fully open during startup and operation of thereactor shall be provided.
(5) The vessel or piping belonging to the Reactor Facility, which would receive a pressure exceeding the pressure defined by
the design on the external surface due to the internal pressure getting less than the atmospheric pressure, shall be provided
with a vacuum breaker in order to ensure the capacity equal to or more than one required for overpressure prevention
pursuant to the following each item:(j) The material of vacuum breaker shall be pursuant to the followings;(a) The provision of Article 9, item 1 shall apply mutatis mutandis to the material of vacuum breaker to be provided
to the Class 1 Vessel and Class 1 Piping, and(b) The provision of Article 9, item 2 shall apply mutatis mutandis to the material of vacuum breaker to be provided
to the Reactor Containment, Class 2 Vessel and Class 2 Piping.
(ii) The Reactor Containment shall be provided with two or more vacuum breakers at appropriate locations; and(iii) The vessel or piping other than one described in the preceding item shall be provided with one vacuum breaker atappropriate location.
(6) In the case that the fluid released from safety valves, relief valves, rapture disks or vacuum breakers contains radioactive
materials, the Reactor Facility shall be provided such that they can be safely processed.
Article 11 (Pressure Test etc.)
The Class 1 Component, Class 2 Component, Class 3 Component, Class 4 Piping and Reactor Containment shall withstandthe pressure proof test under the pressure of the following each item, and have no significant leakage. However, when it has
been confirmed to withstand the pressure concerned in the case that the test is performed using air pressure, it can be
confirmed that there is no significant leakage at the pressure concerned reduced to the Maximum Operating Pressure (0.9
times of the Maximum Operating Pressure for the Reactor Containment);(i) The pressure of pressure test for a component that receives internal pressure shall exceed the Maximum Operating
Pressure of the component and be one at which the overall strain produced is within the elastic range. However, the test
pressure for the Class 1 Component, Class 2 Piping or Class 3 Piping which pressure test is conducted together with the-- Provisional translation for information only --A3-9reactor pressure vessel as a unit, can be the pressure exceeding its operating pressure during normal operation after the test
has been conducted prior to the fuel loading, and(ii) The test pressure of the pressure test of the component that receives external atmospheric pressure as the inside gets
below atmospheric pressure shall be the pressure exceeding the largest difference between atmospheric pressure and theinternal pressure. In this case, the test pressure of the pressure test can be applied from the inside of the component.
(2) The Class 1 Component, Class 2 Component, Class 3 Component and Class 4 Piping shall have no significant leakage
when the leak test is conducted at its normal operating pressure.
(3) The Reactor Containment shall have no significant leakage when the air tightness test is conducted at an air pressure
equal to 0.9 times of the Maximum Operating Pressure.
Article 12 (Surveillance Test Piece)
The vessel belonging to the Reactor Facility, which material would significantly deteriorate when irradiated by neutron with
one mega electron-volt or more, shall be provided in its inside with the surveillance test piece defined in the following each
item for verification of the effects of the irradiation so that the vessel concerned would not have a brittle fracture under its
expected operating conditions;(i) The material from which the surveillance test piece is made shall have the production history equivalent to that of thevessel material subject to the neutron irradiation,(ii) The number of surveillance test pieces shall be appropriate to confirm the change in the mechanical strength and fracture
toughness of the vessel material by taking out and testing them after the vessel is placed in service, and(iii) The surveillance test pieces shall be located such that they would be under the conditions equivalent to those of neutron
spectrum, amount of neutron irradiation and temperature history that the vessel material is subject to.
Article 13 (Reactor Core etc.)
The material of fuel, moderator, reflector and structures supporting these shall maintain the required physical and chemical
properties under the severest conditions that are produced with the pressure, temperature and Radiation during normal
operation.
(2) The material of fuel, moderator, reflector and structures supporting these shall withstand the Maximum Operating
Pressure, dead weight, load to be applied etc.
Article 14 (Thermal Shield)
The reactor pressure vessel of which material would significantly deteriorate due to Radiation shall be provided with thethermal shield to prevent it.
(2) The thermal shield set forth in the preceding paragraph shall be provided in a manner not to compromise operation of thereactor due to its deformation caused by thermal stress.
Article 15 (Primary Coolant)
The Primary Coolant shall maintain the required physical and chemical properties under the severest conditions that are
produced with the pressure, temperature and Radiation during normal operation.
Article 16 (Circulation Systems etc.)
The nuclear power plant shall be provided with the system described in the following each item;(i) The system to circulate the Primary Coolant with the capacity that can transfer the heat generated in the reactor pressure
vessel,(ii) The system to automatically control the pressure fluctuation in the reactor pressure vessel due to a load change etc.,(iii) The system to automatically make up the loss of the Primary Coolant during normal operation or during small leakage of
the Primary Coolant, etc.,(vi) The system to maintain concentrations of impurities and radioactive materials in the Primary Coolant below the values
that would not compromise operation of the nuclear power plant,(v) The system capable of removing the residual heat generated in the reactor pressure vessel during reactor shutdown(including the time of total AC power loss of short duration, and(vi) The system that can transfer the heat removed by the system of the preceding item to an ultimate heat sink.
Article 16-2 (Reactor Coolant Pressure Boundary)Components that compose the Reactor Coolant Pressure Boundary shall be provided in such a way that they can withstand ashock associated with a damage of the Primary Cooling System etc. and increase in their load due to change of the reactor
core reactivity, etc.
Article 16-3 (Leakage of Reactor Coolant Pressure Boundary etc.)
The Reactor Coolant Pressure Boundary shall be provided with isolation devices in order to restrict the runoff of the Reactor
Coolant.
(2) The Reactor Facility shall be provided with devices to detect the leakage of the Reactor Coolant from the Reactor Coolant
Pressure Boundary.
Article 17 (Emergency Core Cooling System)
A nuclear power plant shall be provided with the emergency core cooling system.
(2) The emergency core cooling system shall have the following functions;(i) The system shall be capable of preventing the temperature of cladding tube from increasing above the temperature thatcauses a melt down of fuel material or significant failure of the fuel assembly, and(ii) The system shall prevent generation of a significant amount of hydrogen by the reaction of the cladding tube with thecoolant.
(3) The pump of emergency core cooling system shall have the capability to function normally even under the severest
condition expected of the pressure and temperature in the reactor pressure vessel or the Reactor Containment.
(4) The emergency core cooling system shall be designed and constructed so that it can be tested during operation of thereactor in order to confirm its operability.
-- Provisional translation for information only --A3-10Article 18 (Discharge of the Primary Coolant)
When discharging the Primary Coolant containing radioactive materials (including the fluid containing radioactive materials
discharged from the equipment of Article 16, item 4) to the outside of the Primary Cooling System during normal operation,
the device to process the fluid safely shall be provided.
Article 19 (Installation of Check Valve)
The piping delivering the fluid not containing radioactive materials to the vessel or piping that contains the Primary Coolant
containing radioactive materials or the piping, or the system that processes radioactive wastes (excluding a stack and thosefacilities provided in Articles 28 and 31; hereinafter the same shall apply in Article 21) shall be provided with a check valve.
However, when there would be no back-flow of the fluid containing radioactive materials to the piping providing the fluid
not containing radioactive materials, this shall not apply.
Article 20 (Instrumentation Devices)
The nuclear power plant shall be provided with the devices to measure the matters described in the following each item, andin this case, when it is difficult to directly measure them, the said devices may be replaced with devices to indirectly measure
the matters concerned;(i) Neutron flux density in the reactor core,(ii) Reactor period,(iii) Position of control rods and the concentration of the liquid control material, in the case when it is used,(iv) Following matters on the Primary Coolant,(a) The concentration of radioactive materials and impurities, and(b) The pressure, temperature and flow rate at the inlet and outlet of the reactor pressure vessel,(v) The water Level in the reactor pressure vessel (pressurizer, in the case when a pressurizer is provided) and steam
generator,(vi) The pressure, temperature, concentration of combustible gases, concentration of radioactive materials and dose
equivalent rate in the Reactor Containment,(vii) The concentration of radioactive materials in the main steam piping and the air-ejector exhaust gas, etc.,(viii) The pressure, temperature and flow rate of the Secondary Coolant and the concentration of radioactive materials in theSecondary Coolant at the outlet of the steam generator,(ix) The concentration of radioactive materials at the stack outlet or at the locations close to it during the stack in service,(x) The concentration of radioactive materials in discharge water at the discharge point or at the locations close to it,(xi) The concentration of radioactive materials in discharge water at the outlet or at the locations close to it, of an drainage
line provided with an opening in the Controlled Area (the Controlled Area excluding the area where only the dosage related
to external Radiation would exceed the dosage specified in a public notice pursuant to the provision of Article 2, item 9;hereinafter the same shall apply) that would be contaminated with radioactive materials,(xii) The dose equivalent rate in the area of the Controlled Area that is always occupied and other area that especiallyrequires the Radiation control (it means the fuel handling area etc.),(xiii) The air dose rate and concentration of radioactive materials in the region adjacent to the Peripheral Monitoring Area,
and(xiv) The wind direction and wind velocity at the nuclear power plant.
(2) The device described in item 6 of the preceding paragraph, to measure the dose equivalent rate shall have redundancy andindependency.
(3) The devices to measure the matters listed in Paragraph 1, item 1 and from item 3 to 14 shall be able to display and record
the measurement results.
Article 21 (Alarm Devices etc.)
The nuclear power plant shall be provided with devices to reliably detect and automatically sound the alarm about thefollowing events: operation of reactor would be significantly compromised due to loss of function of its mechanicalequipment or erroneous operator action; remarkable increase in the concentration of radioactive materials described in item 9of the preceding Article or in the dose equivalent rate described in items 12 and 13 of the said Article; liquid radioactive
waste would significantly leak from facilities process or store the liquid radioactive waste (excluding gaseous one; hereinafter
the same shall apply).
(2) The nuclear power plant shall be provided with devices to display the operating state of the major mechanical equipment
for the reactor, the Primary Cooling System and facilities processing or storing radioactive wastes(3) The nuclear power plant shall be provided with the appropriate communication system to give necessary instructions topersons in the nuclear power plant concerned, in the event of failure or damage etc. of a facility for the Primary Cooling
System.
Article 22 (Safety Protection System)
The nuclear power plant shall be provided with the safety protection system pursuant to the following each item;(i) One capable of not letting the Allowable Fuel Damage Limit be exceeded, working together with the reactor shutdown
system and the Engineered Safety Features, in the event of the Abnormal Transient During Operation or when operation of
the reactor is compromised due to an earthquake etc.,(ii) The mechanical equipment or the trains that compose the system shall have redundancy not to loose their safety
protection functions in the event of a single failure or a single removal from its in-service state,(iii) The trains that compose the system shall be separated from each other, and be independent of each other not to loose
their safety protection functions,(iv) Even in the event of a loss of its driving sources, a system cutoff or other adverse conditions, the safety protection systemshall be able to maintain the states that would not compromise the safety of the Reactor Facility by bringing it to more safe
states or keeping the states concerned,-- Provisional translation for information only --A3-11

(v) When a part of the instrumentation and control system is shared with the safety protection system, the system shall befunctionally separated from the instrumentation and control system in order not to loose its safety protection function,(vi) The system shall enable the conduct of the required tests to confirm its capability during reactor operation, and(vii) The safety protection system shall enable the change of its actuation setpoints according to the operation conditions.
Article 23 (Reactivity Control System and Reactor Shutdown System)
The nuclear power plant shall be provided with the reactivity control system and reactor shutdown system. In this case, thereactivity control system and the reactor shutdown system may be provided not independently from each other.
(2) The reactivity control system shall have the capability to control a reactivity change associated with a planned powerchange without exceeding the Allowable Fuel Damage Limit.
(3) The reactor shutdown system shall consist of two or more independent systems such as control rods, liquid control
material etc., and shall have the following capabilities;(i) Each of the two or more independent systems shall be able to bring the reactor to a non-critical state and maintain the state
under the high temperature condition during normal operation, and also, at least one of the reactor shutdown systems shall beable to bring the reactor to a non-critical state and maintain the state under the high temperature condition during theAbnormal Transient During Operation without exceeding the Allowable Fuel Damage Limit. In this case, the Reactivity
Worth produced by injecting the liquid control material along with an actuation of the emergency core cooling system etc.
can be added,(ii) At least one of the systems shall be able to bring the reactor to a non-critical state and maintain the state under the cold
condition during normal operation and the Abnormal Transient During Operation,(iii) In an accident such as a loss of the Primary Coolant etc., at least one of the systems shall be able to bring the reactor to a
non-critical state, and at least one of the systems is able to maintain the reactor in a non-critical state. In this case, theReactivity Worth produced by injecting the liquid control material along with an actuation of the emergency core coolingsystem etc. can be added, and(iv) In the case of using control rods, the provisions from item 1 to 3 shall be satisfied, even when one control rod with thelargest Reactivity Worth is stuck.
(4) The Maximum Reactivity Worth of Control Rod and the Reactivity Insertion Rate shall not cause a damage of the Reactor
Coolant Pressure Boundary and the reactor core that compromises core cooling in the event of the anticipated reactivity
insertion event (it means an event where a reactivity is abnormally inserted to the reactor).
(5) The control rod, liquid control material etc. shall be able to maintain the required physical and chemical properties under
the severest conditions of the pressure, temperature, and Radiation during normal operation.
Article 24 (Control Material Drive Mechanism)
The mechanism to drive control material shall be provided pursuant to the following each item;(i) It shall be able to drive the control material at the velocity appropriate to the characteristics of the reactor,(ii) It shall not be able to drive control rods at the velocity that exceeds the Allowable Fuel Damage Limit, even in the eventof an abnormal withdrawal of control rods during normal operation of the reactor,(iii) It shall not drive control rods in a direction to increase the reactivity of the reactor in the event of a loss of the powersource for driving control rods, and(iv) The mechanism to drive control rods shall not damage the control rods, fuel assemblies, reflectors etc. by insertion of thecontrol rods or other shocks.
Article 24-2 (Nuclear Reactor Control Room etc.)
The nuclear power plant shall be provided with the reactor control room.
(2) The reactor control room shall be centralized with the facility to operate the equipment for the reactivity control system
and reactor shutdown system, the one to operate the equipment for ensuring safety of the reactor in an emergency such as the
emergency core cooling system, the one to display operating status of major mechanical equipment for the reactor and thePrimary Cooling System, the one to display measurement results of major instrumentation, and other major equipment forsafe operation of the reactor (including the equipment specified in Article 21, paragraph 1) and also, shall be capable ofoperating them properly without an inadvertent action.
(3) The reactor control room and the passageway etc. leading to it shall be provided with the shielding and other appropriate
Radiation protection measures, and isolation of the ventilation system against the toxic gas that will be generated with the fire
etc. outside the control room, and other appropriate protection measures, so that personnel etc. can enter the reactor control
room without any problem and remain there for a certain period of time to take measures such as shutdown of the reactor in
the event of a failure or damage of the facilities for the Primary Cooling System.
(4) The nuclear power plant shall be provided with the device capable of shutting down the reactor and maintaining it in asafe state from a place other than the reactor control room when it is not available due to a fire etc.
Article 24-3 (Emergency Management Office of A Nuclear Power Plant)
The nuclear power plant shall be provided with the emergency management office of the nuclear power plant at a place otherthan the reactor control room to take appropriate measures in the event of a damage etc. of the facilities for the Primary
Cooling System.
Article 25 (Fuel Storage Facility)
The storage facility for fuel assemblies necessary for normal operation or the spent fuel (hereinafter referred to as "fuel")shall be provided pursuant to the following each item:(i) The structure shall not allow the fuel to attain the criticality;(ii) The fuel shall not melt with the decay heat;(iii) It shall have a capacity large enough to store the fuel, as required;(iv) The water pool to store the spent fuel and other highly radioactive fuel shall be pursuant to the followings;(a) The structure shall not allow the water to overflow or leak,-- Provisional translation for information only --A3-12

(b) It shall have the amount of water required to shield the Radiation from the fuel,(c) When the fuel cladding tube would corrode significantly, it shall be prevented,(d) It shall be able to detect a leakage of the water and anomalies of the water temperature in the water pool, and(e) Its functions shall not be compromised at the time of the anticipated fuel drop during fuel handling.
(v) In the case that the radioactive materials released by a fuel damage due to a fuel drop would cause Radiation hazards tothe public, a facility to contain the fuel storage facility and a facility to reduce the release of radioactive materials shall be
provided in order to mitigate the consequences of the radioactive materials to the outside of the nuclear power plant;(vi) The dry cask (hereinafter referred to as "cask") that stores the spent fuel in a nuclear power plant shall be pursuant to
the followings;(a) It shall be able to contain the radioactive materials contained in the spent fuel, and to monitor its functions
appropriately,(b) It shall have an appropriate shielding capability for the Radiation from the spent fuel,(c) It shall be able to prevent the significant corrosion or deformation of the spent fuel cladding tube, and(d) The main body of the cask, etc. shall be of the appropriate materials and structures for the temperature, Radiation,load, and other conditions during its service.
(vii) Persons other than persons in charge shall not have unnecessary access.
Article 26 (Fuel Handling Facilities)
The fuel handling facilities shall be provided pursuant to the following each item;(i) It shall have the capability to handle the fuel to be used for normal operation,(ii) It shall have the structures that do not allow the fuel to attain the criticality,(iii) The fuel shall not melt with the decay heat,(iv) It shall not damage the fuel during handling,(v) The container to enclose the fuel shall withstand the shock, heat etc. during its handling, and shall not be damaged easily,(vi) The dose equivalent rate on the surface and the dose equivalent rate at a distance of one (1) meter from the surface of thesaid container in the preceding item with fuel loaded shall not exceed the dose equivalent rates specified in a separate public
notice, respectively to prevent its Radiation hazards. However, This shall not apply to the container only used in theControlled Area, and(vii) In the even of a loss of the power source for handling the fuel during fuel handling, fuel drop shall be prevented byproviding a mechanism to retain the fuel, etc.
Article 27 (Biological Shield etc.)
The biological shield shall be provided at places inside the nuclear power plant and required to prevent Radiation hazards due
to external Radiation, pursuant to the following items;(i) It shall have a shielding capability required for preventing Radiation hazards,(ii) Places with openings or penetrations such as piping shall be provided with measures for Radiation leak, as required, and(iii) It shall be able to withstand its dead weight, additional load and thermal stress.
(2) The Reactor Facility, and steam turbine driven by the Reactor Coolant and its associated systems shall be provided suchthat the air dose rate in the vicinity of the site by the direct gamma ray and sky-shine gamma ray from the facility or systems
concerned during normal operation is less than the dose limit specified in a separate public notice.
Article 28 (Ventilation System)
The ventilation system shall be provided at places inside the nuclear power plant, required to prevent Radiation hazards dueto the air contaminated with radioactive materials, pursuant to the following items;(i) It shall have a ventilation capacity required for preventing Radiation hazards,(ii) It shall be of a structure that could make it hard for the air contaminated with radioactive materials to leak and would not
produce backflow,(iii) In the case of installing the device to clean the air to be discharged, it shall be of the structure easy for the filter to be
decontaminated from radioactive materials or to be replaced, and(iv) An intake port shall be constructed so as to make it hard for the air contaminated with radioactive materials to be taken
in.
Article 29 (Prevention of Contamination by Radioactive Materials)
The surface of walls, floors and other portions inside of buildings frequently occupied at the nuclear power plant, which
would be contaminated with radioactive materials and touched by people, shall be easy to remove the contamination withradioactive materials.
(2) The nuclear power plant shall be provided with the system to remove the contamination with radioactive materials of thematters that would be touched by people.
Article 29-2
A drainage line with an opening in the Controlled Area to be possibly contaminated with radioactive materials, which
discharges the wastewater outside a nuclear power plant, shall be provided with discharge water monitoring equipment andthe system to safely process the wastewater containing radioactive materials.
Article 30 (Waste Treatment System, etc.)
The nuclear power plant shall be provided with the system to process radioactive wastes（including a stack, and excluding
those provided in Article 28 and the following Article）pursuant to the following items;(i) It shall have the capability to process radioactive wastes generated in a nuclear power plant such that the concentration ofradioactive materials in the air outside the Peripheral Monitoring Area and the water at the boundary of the Peripheral
Monitoring Area is below the values specified in a separate Public Notice, respectively,(ii) It shall be provided separately from the facility that processes wastes other than radioactive wastes. However, in the casewhere liquid waste other than radioactive wastes is delivered to the facility that processes liquid radioactive wastes and the-- Provisional translation for information only --A3-13liquid radioactive wastes would not flow back to the facility that deals with wastes other than radioactive wastes, this shallnot apply,(iii) It shall be of a structure that could make it hard for radioactive wastes to leak and chemical agents etc. would not
significantly corrode,(iv) The provision of Article 28, item 3 shall apply mutatis mutandis to the facility that processes gaseous wastes, and it shallnot discharge gaseous wastes at the locations other than the outlet of a stack,(v) The container for transport of the liquid radioactive wastes, and the highly radioactive solid wastes which are generated
from what is provided within the Reactor Coolant Pressure Boundary, within the site of a nuclear power plant shall withstandthe shock, heat etc. during its handling, and shall not be damaged easily. However, for the container only used in theControlled Area, this shall not apply, and(vi) The said container in the preceding item with radioactive wastes loaded shall be able to be shielded so that the dose
equivalent rate at the surface and the dose equivalent rate at a distance of one (1) meter from the surface of the container do
not exceed the dose equivalent rates specified in a separate Public Notice to prevent Radiation hazards. However, this shallnot apply to the container only used in the Controlled Area.
(2) The facility in which the system to process liquid radioactive wastes (limited to a portion where a leakage of liquidradioactive wastes would spread, and hereinafter; hereinafter the same shall apply in this paragraph) is provided shall be
provided pursuant to the following each item;(i) The surface of floors and walls inside of the facility should be of a structure that could make it hard for liquid radioactive
wastes to leak,(ii) The floor of the facility shall be of a structure such that liquid radioactive wastes is led to effluent receiving ports bymeans of the slope of troughs provided at the slope of floor or the floor, and the periphery of the facility that processes liquidradioactive wastes shall be provided with an embankment to prevent leakage of liquid radioactive wastes from spreading,(iii) The entrance leading to the outside the facility or its periphery shall be provided with an embankment to prevent liquidradioactive wastes from leaking outside the facility. However, in the case where the floors of the facility are lower than thefloors of the adjacent facilities or the ground surface and there would be no leakage to the outside of the facility, this shall not
apply, and(iv) It shall be provided such that no floor of the facility exists above the discharge water channel that discharges wastewateroutside a nuclear power plant (excluding those for storm water, which have no opening in the Controlled Area which would
be contaminated with radioactive materials, and those provided with discharge monitoring equipment and the system thatsafely processes the wastewater containing radioactive materials.)(3) The container for transporting the liquid radioactive wastes of Paragraph 1, item 5 shall be provided in the same manneras provided in Paragraph 2, item 3, to prevent the spread of a leakage of liquid radioactive wastes. However, this shall not
apply to the container only used in the Controlled Area and of a structure that would not cause leakage.
Article 31 (Waste Storage Facility etc.)
The facility to store radioactive wastes shall be provided pursuant to the following each item;(i) It shall have the capacity to store the radioactive wastes to be generated during normal operation,(ii) It shall be of a structure that could make it hard for the radioactive wastes to leak, and(iii) It shall withstand the heat generated by the decay heat and irradiation, and shall not significantly corrode by chemical
agents etc.
(2) The facility provided with the facility to store solid radioactive waste shall be provided such that the contamination withradioactive wastes would not spread.
(3) The provision of paragraph 2 of the preceding Article shall apply mutatis mutandis to the facility provided with thefacility to store liquid radioactive wastes. In this case, "the facility that processes liquid radioactive wastes" shall be deemed
to be replaced with "the facility that stores liquid radioactive wastes".
Article 32 (Reactor Containment Facility)
The nuclear power plant shall be provided with a Reactor Containment facility pursuant to the following each item, in orderthat the leak rate in the event of a failure or damage of the components for the Primary Cooling System would not cause
Radiation hazards to the public:(i) The reactor Containment shall be pursuant to the followings;(a) It shall withstand the maximum pressure and maximum temperature anticipated in the event of a failure or damageof the components for the Primary Cooling System,(b) In the case of providing an opening to the Reactor Containment, the airtightness shall be ensured, and(c) The penetrations of the Reactor Containment and the entrance shall be able to undergo the leak test according to
the leakage anticipated.
(ii) The piping to be installed penetrating the Reactor Containment shall be provided with isolation valves (shutoff isolation
valve (limited to that with a locking device) or automatic isolation valve (excluding the check valve without isolation
function); hereinafter the same shall apply) pursuant to the followings;(a) The piping to be installed to the Reactor Containment and penetrate the Reactor Containment shall be provided
with one isolation valve at the locations close to the inside and outside of the penetration concerned;(b) Irrespective of the preceding item (a), it may be pursuant to the followings;
1. The piping that does not have an opening within a facility for the Primary Cooling System and the Reactor
Containment and would not be damaged in the event of a damage of a facility for the Primary Cooling
System or the piping that would not cause a leakage of radioactive materials contained in the Reactor
Containment due to the remaining liquid inside because of its structure in the event of a damage of a facility
for the Primary Cooling System shall be provided with one isolation valve at the location close to the insideor outside of the penetration concerned, and-- Provisional translation for information only --

A3-14

2.

When an isolation valve is to be provided at the inside or outside of the penetration, for the piping at one
side of installation location of which isolation valve function could be considered to significantly deteriorate
due to humidity etc., two isolation valves shall be provided at the location close to the other side of thepenetration.
(c) Notwithstanding the provisions of the preceding (a) and (b), it is not required to provide an isolation valve in thefollowing cases;
1. In the case that providing an isolation valve to the piping of a system required for terminating an accident
would compromise safety and the isolation function of the Reactor Containment could not be lost because of
the piping of the system concerned, and
2. In the case of the piping related to the measurement or control rod drive mechanism, through which theleakage from the piping concerned is controlled to the fully acceptable extent.
(d) The isolation valve shall not lose its isolation function, even in the event of a loss of its driving power source after
it has been closed; and(e) The isolation valve shall be able to undergo the leak test according to the leakage anticipated.
(iii) When the safety of the Reactor Containment would be compromised due to hydrogen and oxygen generated in the eventof a failure or damage of a facility for the Primary Cooling System, a system to reduce hydrogen or oxygen concentration
shall be provided;(iv) In the case that the leakage of gaseous radioactive materials from the Reactor Containment would cause Radiation
hazards to the public in the event of a failure or damage of a facility for the Primary Cooling System, a system to reduce theconcentration of radioactive materials concerned (including the facility to contain the radioactive materials concerned) shallbe provided; and(v) In order to prevent an increase in the pressure and temperature in the Reactor Containment made in the event of a failure
or damage of a facility for Primary Cooling System from compromising safety of the Reactor Containment, a system toremove the heat generated in the Reactor Containment (hereinafter referred to as the " Reactor Containment heat-removal
system ") shall be provided pursuant to the followings;(a) The pump of the Reactor Containment heat-removal system shall be able to function normally even under theseverest condition of the pressure and temperature anticipated in the Reactor Containment, and(b) The Reactor Containment heat-removal system shall be able to undergo the test to confirm its capability during
reactor operation.
Article 33 (Emergency Power Supply System)
At least two lines of the electric lines connected to the nuclear power plant shall be able to be accepted at the nuclear power
plant concerned, be of special high operating voltage exceeding 60,000 volts, and be provided such that the nuclear power
plant concerned could be networked to a power transmission system.
(2) The nuclear power plant shall be provided with power generating equipment with an internal combustion engine as its
generative power or emergency standby power unit with the function equivalent to or more than it, in order to maintain thefunction of the equipment required for ensuring safety in the event of stoppage of deliveries of the electricity from the electriclines described in the preceding paragraph or a generator usually used at the nuclear power plant concerned.
(3) Equipment especially required to ensure the safety of a nuclear power plant shall be provided with a uninterruptible
power supply device or the one with the function equivalent to or more than it.
(4) The emergency power supply system and its associated systems shall have redundancy or diversity and independency,
and sufficient capacity for ensuring the functions of facilities such as the Engineered Safety Features during an Abnormal
Transient During Operation or an accident such as a loss of the Primary Coolant, even in the event of a single failure of themechanical equipment composing the system.
(5) The nuclear power plant shall be provided with batteries that have capacity required to ensure that the facility to safely
shutdown the reactor and cool it down after its shutdown can function even in the event of a loss of all AC power for a shortperiod of time.
Article 34 (Application, Mutatis Mutandis)
The provision of Article 8, paragraph 3 shall apply mutatis mutandis to the steam turbine driven with the Reactor Coolant andits associated systems to be provided at a nuclear power plant,.
(2) The provision of Article 9, item 15 and the provisions of the Chapter 2 of the Ordinance of the Ministry for Establishing
Technical Standards on Thermal Power Generation Equipment (Ordinance of the Ministry of International Trade andIndustry No. 51 of 1997) shall apply mutatis mutandis to the auxiliary boiler to be provided at a nuclear power plant.
(3) The provision of Article 9, item 15 and the provisions of Chapter 3 of the Ordinance of the Ministry for Establishing
Technical Standards on Thermal Power Generation Equipment shall apply mutatis mutandis to the steam turbine and its
associated systems to be provided at a nuclear power plant.
(4) The provisions of Article 25 and Articles 26 to 29 of the Ordinance of the Ministry for Establishing Technical Standards
on Thermal Power Generation Equipment shall apply mutatis mutandis to the internal combustion engine to be provided at anuclear power plant.
(5) The provision of Article 4 of the Ordinance of the Ministry for Establishing Technical Standards on Thermal Power
Generation Equipment shall apply mutatis mutandis to the electric structures to be provided at a nuclear power plant.

-- Provisional translation for information only --A3-15

(3) The Law for the Independent Administrative Agency, Japan Atomic Energy Agency, (Excerpt)(Law No. 155, December 3, 2005)
Latest Revision: Law No. 28, April 20, 2007;
Law No.28, April 20, 2007 (unenforced)(Mission of the Agency)
Article 4. The incorporated administration agency, Japan Atomic Energy Agency (hereinafter called as the " Agency"), in
accordance with the basic policies provided in Article 2 of the Atomic Energy Basic Law, comprehensively, systematicallyand efficiently conducts fundamental and applied researches on the nuclear power, develops fast breeder reactors and therequired nuclear fuel materials for the nuclear fuel cycle, and develops technologies for reprocessing nuclear fuel materials
and disposal of high level radioactive wastes, etc., and performs dissemination of these outcomes etc. so as to contribute inpromoting research, development, and utilization of atomic energy, which helps the welfare of human society andimprovement of the life of the people.
(Scope of Duties)
Article 17. The Agency performs the following duties in order to achieve the missions provided in Article 4;
1 To carry out basic research of nuclear energy,
2 To carry out applied research of nuclear energy,
7 To collect, organize and provide information about nuclear energy, and
8 As other duties than ones described in Item 1 to 3, the Agency performs experiments, researches, surveys, analysis andjudgment on the nuclear energy, when requested by heads of related administrative agencies or local governments who
authorize that they are necessary.

(4) Regulatory Guide for Reviewing Seismic Design of Nuclear Power Reactor Facilities
September 19, 2006
The Nuclear Safety Commission
1．Introduction
This guide is provided to show the basis of the judgment for adequacy of the seismic design policy in the standpoints toensure seismic safety at the Safety Review related to the application for the establishment license (includes the application ofalteration of an establishment license) of the individual light water power reactor.
The former ‘Regulatory Guide for Reviewing Seismic Design of Nuclear Power Reactor Facilities (decided by the Nuclear
Safety Commission “NSC” on 20 July 1981 and revised on 29 March 2001, hereinafter referred to as “Former Guide”) ’ was
the guide which was revised based on the state of arts of evaluating methods of static seismic force etc. by the NSC in July
1981, which had been provided in September 1978 by the Atomic Energy Commission. And it was partially revised in March
2001.
This time, overall revision of Former Guide has been conducted by reflecting accumulated new seismological andearthquake engineering knowledge and remarkable improvement and development of seismic design technology of nuclear
power reactor facilities.
Incidentally, this guide shall be revised to reflect the coming new knowledge and experiences suitably according toaccumulation of new findings.
2．Scope of Application
This guide shall be applied to the nuclear power reactor facilities (hereinafter referred to as “Facilities”).
Nevertheless, basic concept of this guide could be referred to other type nuclear reactor facilities as well as other nuclear
related facilities.
Incidentally, if some part of application contents could not comply with this guide, it would not be excluded if it reflected
technological improvements or developments and seismic safety could be ensured farther than satisfying this guide.
3．Basic Policy
A part of Facilities designated as important ones from the seismic design points shall be designed to bear seismic force
exerted from earthquake ground motion and to maintain their safety function, which could be postulated appropriately tooccur but very scarcely in the operational period of Facilities from the seismological and earthquake engineering standpoints
such as geological features, geological structure, seismicity, etc. in the vicinity of the proposed site.
Moreover, any Facilities shall be designed to bear the design seismic force sufficiently which is assumed appropriately forevery classification in the seismic design from the standpoint of radiological effects to the environment which could becaused by earthquake.
Besides, buildings and structures shall be settled on the grounds which have sufficient supporting capacity.
(Commentary)
I．Regarding Basic Policy(1) Regarding determination of earthquake ground motion in the seismic design
In the seismic design, it shall be based on the principle that ‘ the ground motion which could be postulated
appropriately to occur but very scarcely in the operational period of Facilities and are feared affecting severely to
Facilities’ shall be determined adequately, and that, on the premise of this ground motion, the seismic design shall be-- Provisional translation for information only --A3-16

conducted not to give any risk of serious radiological exposure to the public in the vicinity of Facilities from theexternal disturbance initiated by an earthquake.
This policy is equal to the ‘ basic policy’ in Former Guide which is required to the seismic design with the provisionof ‘ nuclear power reactor facilities shall maintain seismic integrity against any postulated seismic force assumed so
sufficiently that no earthquake would induce significant accidents’.
(2) Regarding existence of “Residual Risk”
From the seismological standpoint, the possibility of occurrence of stronger earthquake ground motion which
exceeds one determined on the above-mentioned (1) can not be denied. This means, in determination of seismic
design earthquake ground motion, the existence of “Residual Risk”(defined as such a risk that, by extension of theeffect of the ground motion which exceeds the determined design ground motion of Facilities, impairing eventswould occur to Facilities and the event in which massive radioactive materials diffuse from Facilities would breakout, or the result of these events would cause radiological exposure hazards to the public in the vicinity of Facilities).
Therefore, at the design of Facilities, appropriate attention should be paid to possibility of occurrence of theexceeding ground motion to the determined one and, recognizing the existence of this “Residual Risk”, every effort
should be made to minimize it as low as practically possible not only in the stage of design basis but also in thefollowing stages.
4．Classification of Importance in Seismic Design
Importance in seismic design of Facilities shall be classified into the followings from the standpoints of the possible impact
of radiation to the environment caused by earthquake corresponding to the categories of Facilities.
(1) Classification on Function
S Class : Facilities containing radioactive materials by themselves or related directly to Facilities containing radioactive
materials, whose loss of function might lead to the diffusion of radioactive materials to the environment,
Facilities required to prevent the occurrence of those events and Facilities required to mitigate the consequences
resulting from the diffusion of radioactive materials in the occurrences of those accidents, and also whose
influences are very significant,
B Class : Facilities of the same functional categories as above S Class, however whose influences are relatively small,
C Class : Facilities except for S or B Class, and ones required to ensure equal safety as general industrial facilities.
(2) Facilities of Classes
Facilities of Classes are shown as follows by the above classification of the importance in the seismic design,
1) S Class Facilities :i)Equipment/piping system composing of the ‘reactor coolant pressure boundary’ (the definition is the same that is
described in other Regulatory Guide for Reviewing Safety Design of Light Water Nuclear Power Reactor
Facilities),ii) Spent fuel storage pool,iii) Facilities to add the negative reactivity rapidly to shutdown the reactor and Facilities to preserve the shutdown
mode of the reactor,iv) Facilities to remove the decay heat from the reactor core after reactor shutdown,v)
Facilities to remove the decay heat from the reactor core after the failure accident of reactor coolant pressure
boundary,vi) Facilities to prevent the propagation of radioactive materials directly as the pressure barrier at the failure accidentof reactor coolant pressure boundary,vii) Facilities, except for those in the category ⅵ) above, to mitigate the diffusion of radioactive materials to theenvironment at the accident which involves the release of radioactive materials.
2) B Class Facilities :i)Facilities connected directly to reactor coolant pressure boundary and containing radioactive materials bythemselves or have possibility to contain radioactive materials,ii) Facilities containing radioactive materials. Except for those whose effect of radiological exposure to the public
due to their break is smaller enough to compare with annual exposure limit at the outside of the peripheral
observation area, because of its small inventory of containing radioactive materials or of the difference of the typeof storage system,iii) Facilities related to the radioactive materials except radioactive wastes and have possibility to give excessive
radiological exposure to the public and the operational personnel from their break,iv) Facilities to cool the spent fuels,v)
Facilities except for those of S Class, to mitigate diffusion of radioactive materials to the environment at an
accident which involves the release of radioactive materials.
3) C Class Facilities :Those Facilities not belong to above S or B Class.
5. Determination of Design Basis Earthquake Ground Motion
The ground motion to be established as the basis of the seismic design of the Facilities shall be determined adequately as theground motion to be postulated to occur but very scarcely in the operational period of Facilities from the seismological andearthquake engineering point of view relating to geology, geological structure, seismicity, etc. in the vicinity of the proposed-- Provisional translation for information only --A3-17

site, and to be feared making a serious impact to Facilities. (Hereinafter this ground motion is referred to as ”Design Basis
Earthquake Ground Motion Ss” or “DBGM Ss”.)
DBGM Ss shall be determined on the following principles.
(1) DBGM Ss shall be determined as following two types of earthquake ground motions in horizontal direction and verticaldirection on the free surface of the base stratum at the proposed site, relating to (2)”Site specific earthquakes ground motion
whose source to be identified with the proposed site” and (3) “Earthquake ground motion whose source not to be identified”
mentioned below.
(2) Site specific earthquakes ground motion whose source to be identified with the proposed site shall be determined on thefollowing principles.
1) Taking account of the characteristics of active faults and the situation of earthquake occurrences in the past and atpresent in the vicinity of the proposed site, and classifying the earthquakes by the pattern of earthquake occurrence etc.
plural number of earthquakes which are feared making severe impact to the proposed site shall be selected (hereinafterreferred to as “Investigation Earthquakes”).
2) Following items shall be taken into account concerning the ‘characteristics of the active faults around the proposed site’in above-mentioned 1).
i)
The active faults considered in the seismic design shall be identified as the one of which activities since the late
Pleistocene epoch can not be denied. Incidentally, judgment of the faults can depended upon whether the
displacement and deformation by the faults exist or not in the stratum or on the geomorphic surface formed
during the last interglacial period.
ii) The active faults shall be investigated sufficiently by integrating geomorphological, geological and geophysicalmethods, etc. to make clear the location, shape, activity of the active faults, etc. according to the distance from the
proposed site.
3) For any Investigation Earthquakes selected in above-mentioned 1), following evaluations of earthquake ground motion
bothⅰ) with response spectra and ⅱ) by the method with fault models shall be conducted, and DBGM Ss shall be
determined from respective Investigation Earthquakes.
Incidentally, in evaluating the earthquake ground motion various characteristics (include the regional peculiarity )according to the pattern of earthquake occurrences, seismic wave propagation channel, etc. shall be taken into accountsufficiently.
i)
Evaluation of earthquake ground motion with response spectra
For respective Investigation Earthquakes, response spectra shall be appraised by applying appropriate methods
and the design response spectra shall be evaluated on these spectra, and earthquake ground motions shall be
evaluated in considering the earthquake ground motion characteristics such as duration time, time depending
change of amplitude-enveloping curve suitably.
ii) Evaluation of earthquake ground motion by the method with fault model
For respective Investigation Earthquakes, earthquake ground motions shall be evaluated by settling the seismic
source characteristics parameters with appropriate methods.
4) Uncertainty (dispersion) concerned with the evaluation process of the DBGM Ss in above-mentioned 3) shall beconsidered by applying the appropriate methods.
(3) Earthquake ground motion whose source not to be identified shall be determined on the following principle.
Design Earthquake Ground Motions shall be determined by collecting the observation records near the source which are
obtained from past earthquakes inside the inland earth’s crust, of which the source can not be related directly to any active
faults, settling the response spectra based on those records by taking account of the ground material characteristics of the
proposed site, and adding consideration of the earthquake ground motion characteristics such as the duration time, timedependent change of amplitude-enveloping curve, etc. suitably to these results.
(Commentary)
II．Regarding to determination of DBGM Ss.
(1) Regarding the characteristics of DBGM Ss.
In Former Guide, regarding design basis earthquake ground motion two categories of “Earthquake Ground
Motion S1” and “Earthquake Ground Motion S2” were required to be determined, however in this revision boththese motions were integrated, and enhancement of selection of Investigation Earthquakes, evaluation of groundmotion etc. were strived for DBGM Ss.
This DBGM Ss is the premise ground motion of the seismic design to ensure seismic safety of Facilities and, in
determining it , it’s adequacy should be checked sufficiently according to the latest knowledge in the specificexamination.
(2) The interpretation of the terminology regarding determination of DBGM Ss are as follows.
1)‘Free surface of the base stratum’ is defined as the free surface settled hypothetically without any surface
layer or structure and as the surface of base stratum postulated to be nearly flat with considerable expanse
and without eminent unevenness to plan out design basis earthquake ground Motion. ‘Base stratum’
mentioned here is defined as a solid foundation of which sear wave velocity Vs exceeds 700m/s, andwhich has not been weathered significantly.
2)‘Active faults’ are defined as faults which moved repeatedly in recent geological age and have also
possibility to move in the future.
(3) Regarding the principle of determination DBGM Ss
1)
In selecting Investigation Earthquakes, the characteristics of active faults and the situation of earthquake-- Provisional translation for information only --A3-18

occurrence in the past and at present should be investigated carefully, and furthermore existing research
results concerned with distribution of middle, small and fine size of earthquakes in the vicinity of the
proposed site, stress field, pattern of earthquake occurrence (including shape, movement and mutualinteraction of the plate) shall be examined comprehensively.
2)
Investigation Earthquakes shall be selected depending on the classification considering the pattern ofearthquake occurrence etc. as follows.
i)
Inside Inland Earth’s Crust Earthquake‘Inside inland earth’s crust earthquake’ is defined as the earthquake which occurs in the upper
crust earthquake generation layer and includes one which occurs in the rather offshore coast.
ii) Inter-plates Earthquake‘Inter-plates earthquake’ is defined as one which occurs in the interfacial plane of two mutuallycontacting plates.
iii) Inside Oceanic Plate Earthquake‘Inside oceanic plate earthquake’ is defined as one which occurs inside a subducting ( subducted )oceanic plate, and is classified into two types,‘Inside subducting oceanic plate earthquake’ which occurs near the axis of sea trench or in it’srather offshore area, and ‘Inside subducted oceanic plate earthquake (Inside slab earthquake) ’which
occurs in the land side area from the vicinity of the axis of sea trench.
3)
The evaluation method using fault model should be regarded as important in the case of earthquake
whose source is near the proposed site and process of its failure could be supposed to make large impact
to evaluation of the ground motion.
4)
In consideration of ‘uncertainty (dispersion) concerned with the determination process of DBGM Ss’,appropriate method should be applied considering the cause of uncertainty (dispersion) and it’s extent
which are supposed to make large impact directly to plan out DBGM Ss.
5)
The principle of determination of ‘Earthquake ground motion whose source not to be identified’ isimplied that, if the detailed investigation would be conducted sufficiently considering the situation etc. in
the vicinity of the proposed site, it could not be asserted to evaluate all earthquakes inside inland earth’s
crust in advance which could have still the possibility to occur near the proposed site, therefore thisearthquake should be considered commonly in all applications in spite of the results of the detailed
investigation around the proposed site.
The validity of DBGM Ss determined by materializing this principle should be confirmed specifically inchecking on the latest information at the time of each application. Incidentally, on that occasion,
probabilistic evaluation could be referred as the needs arise regarding the ground motion near the source
generated from the source fault which does not indicate any clear trace on the ground surface.
6)
Regarding ‘Site specific earthquakes ground motion whose source to be identified with the proposed site’and ‘Earthquake ground motion whose source not to be identified’, the exceedance probability ofrespective earthquakes should be referred in each safety examination from the standpoint that it isdesirable to grasp that the response spectra of each seismic ground motion planed out correspond to what
extent of the exceedance probability.
7)
In the case that the necessary investigation and evaluation are implemented in selection of Investigation
Earthquakes and determination of DBGM Ss, existing materials etc. should be referred in considering theaccuracy of them sufficiently. If different result would be obtained compared with the existing evaluation
results, its reason should be shown clear.
8)Regarding the ground which supports the structures of Facilities and Facilities themselves, if the peculiar
frequency characteristics could be found in the seismic response, it should be reflected to determinationof DBGM Ss as the needs arise.
(4) Regarding evaluation of the faults which assumed as the source of earthquake1)
As investigation of the active faults is the basis of the evaluation concerning the faults which is assumed
as the source of earthquake, appropriate investigation should be implemented combining adequately thesurvey of existing materials, tectonic geomorphologic examination, the earth’s surface geological feature
examination, geophysical examination, etc. according to the distance from the proposed site. Especially inthe area near the proposed site, precise and detailed investigation should be applied. Incidentally extent of
the area near the proposed site should be decided suitably considering the relation etc. with DBGM Ss
determined as ‘Earthquake ground motion whose source not to be identified’.
2)
Regarding active folds, active flexures, etc. these should also be the object of investigation inabove-mentioned 1) as well as the active faults and should be considered in the evaluation of the faults
assumed to be the source in accordance with their dispositions.
3)
The dispositions of the faults should be evaluated appropriately grasping the under ground structure etc.
depending on the regional situation. Incidentally, the special consideration should be required if theearthquake should be assumed from the dispositions of faults in the area where the faults are indistinct.
4)
In the case, the scale of earthquake shall be postulated from the length of the fault etc. by applying theempirical formula, the scale should be evaluated adequately considering the special features etc. of theempirical formula.
5)
Uncertainty shall be considered appropriately in assumption of the characteristics of the source, in thecase that sufficient information could not be obtained to settle the source characteristics parameter
including the shape evaluation of the fault to be assumed as the source even by implementing-- Provisional translation for information only --A3-19

investigation of the active faults.
6. Principle of Seismic Design(1) Primal Policy
Facilities shall be designed to fulfill the following primal policies of the seismic design for respective categories of Class.
1)
Respective Facilities of S Class shall maintain their safety functions under the seismic force caused by DBGM Ss. And
also shall bear the larger seismic force loading of those caused by “Elastically Dynamic Design Earthquake Ground
Motion Sd” or the static seismic force shown below. (Hereinafter Elastically Dynamic Design Earthquake Ground
Motion Sd is referred to as “EDGM Sd”.)
2)
Respective Facilities of B Class shall bear the static seismic force shown below. And, as for the Facilities those arefeared of resonating with earthquake, the influence shall be evaluated.
3)
Respective Facilities of C Class shall bear the static seismic force shown below.
4)
In respective items shown above, the integrity of upper Class Facilities shall not be impaired by the damage of thelower Class Facilities.
(2) Computation Method for Seismic ForceThe seismic force for seismic design of Facilities shall be obtained by using the methods shown below.
1)
Seismic forces caused by DBGM Ss
Seismic force caused by DBGM Ss shall be computed by applying DBGM Ss in combining horizontal seismic force
with the vertical seismic force appropriately.
2)
Seismic forces caused by EDGM Sd
EDGM Sd shall be established based on DBGM Ss with the technological judgments. And the seismic forces caused
by EDGM Sd shall be also evaluated in combining horizontal seismic forces with the vertical seismic force
appropriately.
3)
Static seismic force
Evaluation of the Static seismic force shall be based on the followings.
i) Buildings and structures
Horizontal seismic force shall be evaluated by multiplying the seismic story shear coefficient Ci by the coefficientcorresponding to the importance classification of the facilities as shown below, and multiplying the weight at theabove height of the story concerned.
S Class
3.0
B Class
1.5
C Class
1.0Here, Ci of the seismic story shear coefficient shall be obtained in putting the standard shear coefficient Co to be 0.2,
considering the vibration characteristics of the buildings and structures, categories of the ground, etc.
As for the facilities of S Class, both horizontal and vertical seismic forces shall be combined simultaneously in themost adverse fashion. The vertical seismic force shall be evaluated with the vertical seismic intensity which is
obtained by putting the seismic intensity 0.3 as a standard, and by considering the vibration characteristics ofbuildings and structures, categories of the ground, etc. However the vertical seismic coefficient shall be constant in theheight direction.
ii) Components and piping system
The seismic force of respective Classes shall be evaluated with the seismic intensities which are obtained bymultiplying the seismic story shear coefficient Ci in above-mentioned i) by the coefficient corresponding to theimportance classification of the Facilities as the horizontal seismic intensity, and by increasing the horizontal seismic
intensity concerned and the vertical seismic intensity in above-mentioned i) by 20% respectively.
Incidentally, horizontal seismic force shall be combined with the vertical seismic force simultaneously in the mostadverse fashion. However, vertical seismic forces shall be assumed to be constant in the height direction.
(Commentary)
III．Regarding the Design Principle(1) Regarding the necessity of establishment of EDGM Sd
In Former Guide, the design basis earthquake ground motion should have been determined classified as two
categories of Earthquake Ground Motion S1 and Earthquake Ground Motion S2 corresponding to the seismic
importance classification of the buildings, structures, components and piping system, however in this revision,
the determination of DBGM Ss shall only be required. In the seismic design concept to ensure seismic safetyof Facilities, it is the basic principle that the safety functionｓ of the seismically important Facilities shall bemaintained under the seismic forces by this DBGM Ss.
In addition to confirm maintenance of seismic safety functionｓ of the Facilities under this DBGM Ss withhigher precision, establishment of EDGM Sd, which is closely related with DBGM Ss from technicalstandpoint, is also required to be prescribed.
(2) Regarding establishment of EDGM Sd
The concept of ‘to bear the seismic force’ which prescribed in the Article 6. in this Guide means that Facilities
as a whole are designed in the elastic range on the whole to a certain seismic force.
In this case, design in the elastic range means to retain the stress of respective parts of the Facilities under theallowable limits by implementing stress analysis supposing the facilities as the elastic body.
Incidentally, the allowable limits shown here, does not require strict elastic limits and requires the situation-- Provisional translation for information only --A3-20

that the Facilities as a whole should retain in elastic range on the whole even though the case in which the
Facilities partially exceeds the elastic range could be accepted.
Although respective S Class Facilities are required ‘to bear the seismic force’ by EDEGM Sd, this EDGM Sd
is established based on the technological judgment.
The elastic limits condition is the condition that the impact which the Earthquake Ground Motion makes to the
Facilities and the situation of the Facilities can be evaluated clearly, and that it makes a grasp of maintenance ofseismic safety functions as a whole of the Facilities under the seismic force by DBGM Ss more reliable byconfirming that the Facilities as a whole retains in elastic limits condition on the whole under the seismic force
by EDEGM Sd.
Namely EDEGM Sd assumes a part of the roles which the Design Earthquake Ground Motion S1 of Former
Guide used to be attained in the seismic design.
EDGM Sd should be established by multiplying DBGM Ss by coefficients obtained on the technologicaljudgment in considering the ratio of input seismic loads for the safety functional limits and the elastic limits forthe respective Facilities and their composing elements. Here, in evaluating the coefficient, the exceedance
probability which is referred in the determination of DBGM Ss would be consulted.
The concrete established value and reason of establishment of EDGM Sd should be made clear sufficiently inrespective specific application. Incidentally, the ratio of EDGM Sd and DBGM Ss ( Sd/Ss ) should be expectedlarger than a certain extent in considering the characteristics required to EDGM Sd, and should be obtained not
to be less than 0.5 as an aimed value.
In addition, EDGM Sd would be established specifically to respective elements which compose the Facilities
depending on the difference of their characteristics to be considered in seismic design.
Incidentally, regarding to B Class Facilities, ‘as for Facilities that are feared resonating with seismic force
loading, the influence shall be evaluated’, the earthquake ground motion applied to this evaluation would beestablished with multiplying EDGM Sd by 0.5.
(3) Regarding the evaluation of the seismic force by DBGM Ss and EDGM Sd
In case that the seismic force by DBGM Ss and EDGM Sd are evaluated based the seismic response analysis,
the appropriate analytical methods should be selected and suitable analytical consideration should be settled
based on the sufficient investigation in considering to the applicable range of response analysis methods,
applicable limits, etc.
Incidentally, in the case ‘free surface of the base stratum’ is very deep compared with the ground level on
which Facilities would be settled, amplification characteristics of the ground motion on the ground level above
free surface of the base stratum should be investigated sufficiently and be reflected to the evaluation of theseismic response as the needs arise.
(4) Regarding Static seismic force
Evaluation of the static seismic force should be depended upon 1) and 2) shown below.
In addition, regarding to the buildings and structures, the adequate safety margin of retained horizontal
strength of buildings and structures concerned should be checked to maintain the retained horizontal strength
required relating to the importance of Facilities, and the evaluation of retained horizontal strength required
should be complied to the 3) shown below.
1) Horizontal seismic forcei)
The datum plane for evaluation of horizontal seismic force should be the ground surface in principle.
However, if it is needed to consider the characteristics such as the constitution of the building and thestructures and the relation to the surrounding ground around Facilities, the datum plane should be
provided appropriately and be reflected to the evaluation.
ii) Horizontal seismic force applied to aboveground part from the datum plane should be obtained to be thetotal of the seismic forces acted on the part concerned in accordance with the height of the building andthe structure and be calculated with the following formula,
Qi = n・Ci・Wi
where,
Qi : Horizontal seismic force acting on the part in question,n : Coefficient in accordance with importance classification of facilities ( Earthquake-proof S Class 3.0,
Earthquake-proof B Class 1.5, Earthquake-proof C Class 1.0).
Ci : Seismic story shear coefficient, it depends on the following formula,Ci = Z・Rt・Ai・Co
where,
Z : Zoning factor (to be 1.0, the regional difference is not considered),
Rt: A value representing vibration characteristics of building to be obtained by the appropriate
calculation methods specified in standards and criteria which are assumed to be adequate forsafety. Here, ‘the appropriate calculation methods in standards and criteria which are assumed
to be adequate for safety’ corresponds to the Building Standard Law etc.
However, if the value which expresses the vibration characteristics and is evaluatedconsidering the structural characteristics of buildings and structures, and the response
characteristics and situation of the ground in the seismic condition would be confirmed to fallshort of the value calculated by the methods in the Building Standard Law etc. it could be
reduced to the evaluated value by this method ( but equal to or not less than 0.7).
Ai: A value representing a vertical distribution of seismic story shear coefficient according to the-- Provisional translation for information only --

A3-21

iii)

vibration characteristics of building, to be calculated by the appropriate methods specified instandards, criteria and the other appropriate methods as is like Rt,
Co: Standard shear coefficient (to be 0.2),
Wi : Total of fixed loads and live loads supported by the part in question.
Horizontal seismic force which acts on the parts of the buildings and structures under the datum plane
should be evaluated by following formula,
Pk =n・k・Wk
where,
Pk : Horizontal seismic force acting on the part in question.
n : Coefficient in accordance with importance Classification of Facilities (Earthquake-proof S
Class 3.0, Earthquake-proof B Class 1.5, Earthquake-proof C Class 1.0).
k : Horizontal seismic coefficient by the following formula,⎡ H⎤
k ≧ 0.1・ ⎢1 − ⎥ ・ Z⎣ 40 ⎦
where,
H : Depth of each under part from the datum plane; 20 ( m ) at depths of >20 m,
Z : Zoning factor (to be 1.0, the regional difference is not considered ),Wk : Summation of dead loads and live loads of the part concerned.
Incidentally, in the case if the value would be calculated in evaluating the vibration characteristics
suitably by considering the structural characteristics of buildings and structures, and the response
characteristics and situation of the ground in the seismic condition, it would be the value calculated bythis method.

2) Vertical seismic force
The vertical seismic force in the evaluation of the static force to Earthquake-proof S Class Facilities should be
evaluated with the vertical seismic intensity by the following formula,
Cv = Rv・0.3
where,
Cv : Vertical seismic intensity,
Rv : A value representing the vertical vibration characteristics of the building , to be 1.0.
However, based on special investigation or study, if it would be confirmed to fall short of
1.0, it would be reduced to be the value based on the results of investigation or study (butequal to or not less than 0.7).
3) Retained horizontal strength required
Retained horizontal strength required should be evaluated specified in the method in standards and criteria
which are accepted to be adequate for safety.
Here, the standards and criteria which are accepted to be adequate for safety corresponds to the BuildingStandard Law etc.
Incidentally, in evaluation of retained horizontal strength required, the coefficient regarding the importanceclassification of the facilities which is multiplied by the seismic story shear coefficient should be settled to be
1.0 in all the case of Earthquake-proof S, B, C Class and standard shear force coefficient Co which is used inthis case should be provided to 1.0.
7．Load Combination and Allowable Limit
The basic concept about combination of loads and allowable limits which shall be considered in assessing adequacy ofdesign principle regarding seismic safety is as follows.
(1) Buildings and Structures
1) Earthquake-proof S Class Buildings and Structures
Combination with DBGM Ss and allowable limiti)Regarding the combination of normal loads and operating loads with the seismic forces caused by DBGM Ss,
the buildings and structures concerned shall have sufficient margin of deformation acceptability (deformation
at ultimate strength)as a whole, and adequate safety margin compared to the ultimate strength of buildings andstructures.
ii) Combination with EDGM Sd and allowable limit
Regarding resulted stress in combining the normal loads and operating loads imposed with the seismic loads
caused by EDGM Sd or Static seismic force, allowable unit stress specified in standards and criteria assumed tobe adequate for safety shall be established as the allowable limits.
2) Earthquake-proof B, C Class Buildings and Structures
Regarding resulted stress in combining the normal loads and operating loads imposed with Static seismic forces,
allowable unit stress in above-mentioned 1) ⅱ) shall be established as the allowable limits.
(2) Components and Piping System
1) Earthquake-proof S Class Components and Piping System
Combination with DBGM Ss and allowable limitsi)
The functions of Facilities shall not be affected by the occurrence of excessive deformations, crack and failure,even if the most part of structures would reach yield condition and the plastic deformation would occur, with-- Provisional translation for information only --A3-22

respective resultant stress due to combined respective loads which occur in the normal operating condition,
unusual transient condition in operation and accident condition with the seismic loads caused by DBGM Ss.
As for the active components etc., acceleration limit etc. for retaining of function shall be established as theallowable limit, which is confirmed by the verification test etc. regarding the response acceleration caused by the
DBGM Ss.
ii) Combination of EDGM Sd with allowable limits
The yield stress or the stress with equivalent safety to this shall be established as allowable limits to respective
resultant loads due to combined loads at normal operating condition, unusual transient condition in operation andaccident condition imposed with the seismic loads caused by EDGM Sd or Static seismic force.
2) Earthquake-proof B, C Class Components and Piping System
The yield stress or the stress with equivalent safety to this shall be established as allowable limits to respective resultant
loads due to combined loads in normal operating condition and unusual transient condition in operation imposed with theseismic loads caused by Static seismic force.
(Commentary)
IV．Regarding Load Combination and Allowable Limit
The interpretation of the combination of loads and allowable limits should be based on the followings.
(1) Regarding ‘respective loads which occur in unusual transient operation and accident’, if the load acted on by theevents which are feared being caused by the earthquake and the loads, even if which are not feared being causedby the earthquake but being caused by the events which continue in long term if they would occur once, should
be considered to be combined with the seismic load.
However, even if the load is ‘a load which occurs in accident’, considering the relation between occurrence
probability of this accidental event and the duration time, and the exceedance probability of the earthquake, theload caused by this event needs not be considered to be combined with the seismic loads if the probability thatthe both of them occur simultaneously is extremely small.
(2) Regarding the allowable limits for combination of buildings and structures with EDGM Sd etc. though it was
required to be established as the ‘allowable unit stress specified in standards and criteria assumed to be adequate
for safety’, this standards and criteria correspond concretely to the Building Standard Law etc.
(3) ‘Ultimate strength’ in the terms regarding combination of the buildings and structures with DBGM Ss means thebounding maximum bearing load in reaching the condition, which is considered as the ultimate condition of thestructures, where deformation and strain of the structure would increase remarkably by adding the load to thestructure gradually.
(4) Regarding the allowable limit of components and piping system, though the basic principle requires to maintainthe resulted stress under the ‘ yield stress or equivalent safety situation’, this situation corresponds concretely tothe situation specified in the ‘Technical Standards on Structures etc. of Nuclear Power Generation Facilities etc.’
which is prescribed in the Electricity Utilities Industry Law.
8．Consideration of the accompanying events of earthquake
Facilities shall be designed regarding the accompanying events of earthquake with sufficient consideration to the following
terms.
(1) Safety functions of Facilities shall not be significantly affected by the collapses of the inclined planes around Facilities
which could be postulated in the seismic events.
(2) Safety functions of Facilities shall not be significantly affected by the tsunami which could be postulated appropriately to
attack but very scarcely in the operational period of Facilities.

(5) Regulatory Guide for Reviewing Technical Competence of Nuclear Operators
16 NSC Decision No. 6
May 27, 2004
Decision of Nuclear Safety Commission

Introduction
This guide is established in order to be used in the review of capability of a person who applies for license for business of
nuclear fuel fabrication, spent fuel storage, spent fuel reprocessing, radioactive waste storage/disposal or establishment
and operation of a reactor (hereinafter referred to as “business, etc”) as provided for by the Law on the Regulation of
Nuclear Source Material, Nuclear Fuel Material and Reactors (hereinafter referred to as “the Reactor Regulation Law”),
and to confirm whether the applicant has adequate technical capability to appropriately perform its business mentioned
above so that the use of nuclear fuel material or a reactor cannot cause any disaster.
Establishment of the guide was prompted by the criticality accident which occurred at a nuclear fuel fabrication facility
on September 30, 1999. The Nuclear Safety Commission decided to start preparation of a guide for examining technicalcapability of license holders, based on the interim and the final reports on the investigation of the accident. The reports
are “The immediate measures for ensuring safety of nuclear power” decided by the Nuclear Safety Commission on
November 11, 1999, and “A basic policy on the immediate measures of the Nuclear Safety Commission” decided by the
Nuclear Safety Commission on January 17, 2000. The Special Committee on General Safety of Nuclear Power of the
Nuclear Safety Commission started drafting the guide and compiled a report titled “A study on establishing a guide for-- Provisional translation for information only --A3-23

technical capability (June, 2003)”. The Special Committee for Nuclear Safety Standards and Guides of the Nuclear
Safety Commission continued drafting the guide and reported to the Nuclear Safety Commission on March 24, 2004. The
Nuclear Safety Commission invited public opinion on the report, discussed it, and decided “The Examination Guide forTechnical Capability of License holders of Nuclear Power” on May 27, 2004.
The technical capability mentioned in this guide includes knowledge, technology and expertise of engineers of theorganization as well as its organizational capability to ensure safety. This guide provides for the basic requirements to be
met in obtaining license for the businesses mentioned above, license for establishment and operation of a reactor or
license for change of the licenses (hereinafter referred to as “license, etc. for business”).
In review of an application, it is necessary to confirm that contents of the application for license for business, etc. meetthis guide. However, even in the case a part of the application does not comply with this guide, the application shall not
be denied if the reason for non-compliance is adequate.
Furthermore, this guide shall appropriately be revised, as experience in the review of, and relevant knowledge of,technical capability accumulate.
I. Scope of application
This guide shall be applied to a person who applies for license for following businesses, etc. as provided for by the law:i) nuclear fuel fabricationii) establishment and operation of a reactoriii) spent fuel storageiv) spent fuel reprocessingv) radioactive waste storage/disposal
The fundamentals of this guide may be referred to in reviewing an application for license for a nuclear facility other thanthose mentioned above.
As the requirements in this guide are broadly divided into “design and construction” and “operation and maintenance”,
characteristics of the individual business shall be taken into consideration when applying each division of therequirements to the business.
II. Requirements
Guideline 1. Organization for design and construction
The license holder shall have established an appropriate organization which has clear assignment of performing designand construction adequately.
Guideline 2. Assignment of engineers for design and construction
The license holder shall appropriately have assigned engineers who have professional knowledge, technology, andexpertise required to carry out design and construction.
Guideline 3. Experience in design and construction
The license holder shall have sufficient experience in design and construction of equivalent or similar facilities in the saidbusiness.
Guideline 4. Quality assurance activities for design and construction
The license holder shall appropriately have established a system for quality assurance activities required to adequately
perform design and construction.
Guideline 5. Organization for operation and maintenance
The license holder shall have established an appropriate organization which has clear assignment of performing operationand maintenance adequately, or shall appropriately present its policy to establish such an organization.
Guideline 6. Assignment of engineers for operation and maintenance
The license holder shall appropriately have assigned engineers who have professional knowledge, technology, andexpertise required to carry out operation and maintenance, or shall appropriately present its policy to assign suchengineers.
Guideline 7. Experience in operation and maintenance
The license holder shall have sufficient experience in operation and maintenance of equivalent or similar facilities in thesaid business, or shall appropriately present its policy to obtain such experience.
Guideline 8. Quality assurance activities for operation and maintenance
The license holder shall appropriately have established a system for quality assurance activities required to adequately
perform operation and maintenance, or shall appropriately present its policy to establish such a system.
Guideline 9. Education and training of engineers
The license holder shall appropriately present its policy for carrying out education and training of engineers for keeping
and improving their professional knowledge, technology, and expertise.
Guideline 10. Selection and assignment of qualified persons
The license holder shall, when required by a law or regulations based on it, have selected and assigned qualified
engineers for carrying out the said business, or shall appropriately present its policy to select and assign such qualified
engineers.
(Annotation)
Guideline 1. Organization for design and construction
1) The scope of “design and construction” mentioned herein means stages until the license holder passes pre-service
inspection relating the license, etc. for the said business. However, the scope of “design and construction” for radioactive-- Provisional translation for information only --A3-24

waste disposal means stages before the first waste package is received by the receiving facility, as pre-service inspection
is not applicable to the business.
2) “The license holder shall have established an organization…” includes the case where a policy for establishing anorganization in accordance with progress in design and construction is appropriately presented.
Guideline 2. Assignment of engineers for design and construction
1) “Professional knowledge” mentioned herein may include the knowledge required by the national qualification system,
related to this business, such as chief engineer of reactor, chief nuclear fuel engineer, supervisor of radiation protection,
chief engineer of boiler and turbine, chief electrical engineer, and professional engineer.
2) “The license holder shall have assigned engineers…” includes the case where a policy for assigning engineers inaccordance with progress in design and construction is appropriately presented.
Guideline 3. Experience in design and construction“The license holder shall have sufficient experience…” includes the case where experience and technology havesufficiently accumulated through dispatching engineers to equivalent or similar facilities in the said business at home orabroad or through training of engineers at related facilities, or the case where a policy for obtaining experiences inaccordance with progress in design and construction is appropriately presented.
Guideline 4. Quality assurance activities related to design and construction
1) “The license holder shall establish a system…” includes the case where a policy for establishing a system for quality
assurance activities in accordance with the progress in design and construction is appropriately presented.
2) “Quality assurance activities” mentioned herein include a scheme where a policy of the top management for ensuringsafety in design and construction is decided, planning, implementation, evaluation, and improvement of the activities are
conducted in accordance with the quality assurance plan, and continuous improvement of the activities is secured
through auditing. Also, these activities shall be documented and kept.
3) The term “system” may include committees for comprehensive deliberation of the quality assurance activities, if
necessary.
Guideline 5. Organization for operation and maintenance1) The scope of “operation and maintenance” mentioned herein means stages at and after the start of operation of the facility
following passing of the pre-service inspection relating the license, etc. for the said business. However, the scope of“operation and maintenance” for radioactive waste disposal means stages at and after the first waste package is received
by the receiving facility, as pre-service inspection is not applicable to the business.
2) The term “organization” may include committees for deliberating matters related to operational safety, if necessary.
Guideline 6. Assignment of engineers for operation and maintenance“Professional knowledge” may include the knowledge required by national qualification system related to the saidbusiness, such as chief engineer of reactors, chief nuclear fuel engineer, supervisor of radiation protection, chiefengineer of boiler and turbine, chief electrical engineer, and professional engineer.
Guideline 7. Experience in operation and maintenance“The license holder shall have sufficient experience …” includes the case where experience and technology havesufficiently accumulated through dispatching engineers to equivalent or similar facilities in the said business at home orabroad or through training of engineers at related facilities.
Guideline 8. Quality assurance activities related to operation and maintenance
1) “Quality assurance activities” mentioned herein include a scheme where a policy of the top management for ensuringsafety in operation and maintenance is decided, planning, implementation, evaluation, and improvement of the activities
are conducted in accordance with the quality assurance plan, and continuous improvement of the activities is secured
through auditing. Also, these activities shall be documented and kept.
2) The term “system” may include committees for comprehensive deliberation of the quality assurance activities, if
necessary.
Guideline 10. Selection and assignment of qualified persons“Qualified persons” mean those who have a license for chief engineer of reactor or chief nuclear fuel engineer, or those
who meet the criterion as person responsible for operation.

(6) Emergency Preparedness Guide, "Emergency Preparedness for Nuclear Facilities" (Excerpt)(Decision of the Nuclear Safety Commission, June 1980)(Latest Revision: May 2007)

The main points of the latest revision are shown in the following.
In line with international trends, such as International Atomic Energy Agency (IAEA), objectives of the guideline and thesubject facilities etc. were clarified, and the effectiveness of the preventive protective actions was provided. And the overlaps
with the Special Law of Emergency Preparedness for Nuclear Disaster and other guidelines relating to the Nuclear Safety
Commission were arranged. The provisions of main revisions are provided below.
Chapter 1 Preface
1-1 Position of the Guideline

After the accident of the Three Mile Island (TMI) nuclear power station in U.S. that occurred in March 1979, the Nuclear-- Provisional translation for information only --A3-25

Safety Commission (hereinafter called as “the NSC”) studied technical and specialized matters focusing on events unique tonuclear emergency so that the emergency preparedness activities in the surrounding area of a nuclear power station etc. canbe implemented more smoothly, and decided the "Emergency Preparedness of the Surrounding Area of Nuclear Power
Stations etc." in June 1980 (after a partial revision in May 2000 referred to as the "Emergency Preparedness for Nuclear
Installations", (hereinafter called as the "Emergency Preparedness Guideline").
The Emergency Preparedness Guideline is defined in the 10th volume, Nuclear Emergency Preparedness of the Basic Planfor Emergency Preparedness that the guideline shall be sufficiently considered on specialized and technical matters. NSC
established the guideline on the specialized and technical matters concerning emergency measures, to support the national
government, local governments and operators when they develop a plan of the Nuclear Emergency Preparedness andimplement the measures in an emergency.
The Emergency Preparedness Guideline provide basic ideas for the environmental radiation monitoring and emergency
exposure medical treatment in an emergency, and the details are specified in separate guidelines etc. issued by the NSC.
1-2 Scope of the Guideline
The scope of the Emergency Preparedness Guideline shall be the nuclear emergency of the following nuclear installations
provided in the Reactor Regulation Law (limited to the subjects of the Special Law of Nuclear Emergency) and the nuclear
emergency during transportation of nuclear fuel materials etc.
• Reactor facilities (excluding the reactors which are installed in a ship),• Reprocessing facilities,• Fuel fabrication facilities,• Utilization facilities (limited to facilities that use nuclear fuel equal to or exceeding the critical mass), and• Waste disposal facilities and waste storage facilities
1-3 Objectives of Protective Measures
The protective measures described in the Emergency Preparedness Guideline are implemented for the following four
purposes. In addition, in implementing the measures, the principles of justification*1 and optimization*2 provided by theInternational Committee on Radiation Protection (ICRP) etc. should be followed, and it is important to take into
consideration the results of the implementation of the protective measures concerned sufficiently.
• To prevent occurrence of deterministic health effect*3 in residents in the vicinity, in nuclear installation workers, in those
relevant in emergency preparedness, etc.,• To render first aid and to manage the treatment of radiation injuries,• To prevent, to the extent practicable, the occurrence of stochastic health effects in the population, and• Mitigating the anxiety for health of residents in the vicinity, nuclear installation workers, relevant person on emergency
preparedness, etc.
*1 Justification: Implementing a protective action is justified when the benefit is larger than a damage due to a risk or otherimpact caused by implementing the action.
*2 Optimizations: The radiation hazard avoided by each protective action should be balanced to the expense of the action andother damages so that the net benefit attained by the action becomes the maximum.
*3 ICRP, IAEA etc. documents describe health effects of radiation, which are the deterministic effect, although the effectdoes not necessarily occur in all cases, which appears by an exposure of the levels more than a certain dose, and thestochastic effect that appears by an exposure of small dose. In addition, the IAEA document describes objectives ofprotective measures, which are prevention of deterministic effects and reduction of stochastic effects, which are
fundamental principles of radiation protection.
5-3 Indices for Protective Measures
Indices to take protective measures are expressed as the dose (projected dose) expected to receive for individuals if certain
measures are not taken, or measured values as concentration of radioactive materials in foods and drinks.
The projected dose will be presumed from the mode of an abnormal situation, the release situation of radioactive materials orradiations, emergency monitoring information, weather information, SPEEDI network system, etc.*4*4 IAEA documents etc. define indices for protective measures (sheltering / evacuation) with the dose (avertable dose)
avoidable by taking measures. On the other hand, the Emergency Preparedness Guideline applies the projected
doses.
This is because it becomes more safe side decision to implement protective measures by comparing the projected
doses that are obtained assuming a certain period of time than the avertable doses which are obtained defining theperiod of implementing the protective measures, with the indices of protective measures, at the time of
nuclear-emergency generation.

-- Provisional translation for information only --A3-26

Annex 4 The NSC views on, and future actions to take for, the impacts
due to the Niigata-ken Chuetsu-oki Earthquake in 2007
NSC Decision No. 17, 2007
30 July 2007
The Niigata-ken Chuetsu-oki Earthquake in 2007 1 (the Earthquake) on 16 July 2007 hasstrongly shaken the Kashiwazaki-Kariwa Nuclear Power Station of the Tokyo Electric Power
Company, Inc. (TEPCO). The impacts caused include a fire breakout of the Unit 3 transformer, anda release of spilled water containing small amount of radioactive materials to the non-radiation
control area and subsequently to the environment at Unit 6. A joint of the driving shaft of theoverhead crane of the Unit 6 reactor building has been also found to have damaged.
No serious concerns about the environmental impacts have been identified so far. Neverthelessthe Nuclear Safety Commission (NSC) of Japan believes that the damages suffered by the systems
and equipment at the station leave us with big lessons in ensuring seismic safety of nuclear power
plants.
Thorough investigations of the impacts on the nuclear power plants due to the Earthquake areintensively underway, whereas the growing interests are being raised among stakeholders overseas
and in the country. Following are the NSC views, as of today, on the Earthquake impacts andfuture actions to take.
1. Immediate impacts of the Earthquake(1) Ensuring major safety functions such as automatic reactor shutdowns
The Earthquake shook the plants with the maximum seismic accelerations exceeding the values
assumed in the design, but all the units in operation or in power ascension (Units 2, 3, 4 and 7)
were automatically shut down under control. Together with other units in the maintenance mode
(Units 1, 5 and 6), all seven units at the Kashiwazaki-Kariwa Nuclear Power Station are now
brought to the stable cold standby mode. Therefore, the emergency requirements of “shut-down,
cool and contain” have been successfully met for ensuring nuclear safety.
(2) Investigation of impacts due to the Earthquake, and future actions
The incidents caused by the Earthquake are now under in-depth investigation and the total 64
cases on Units 1 to 7 (except four cases of automatic shutdowns due to the Earthquake) have been
1

On July 16, 2007, at 10:13 (JST) there was a M6.8 (preliminary) earthquake at a depth ofapproximately 15km in the off-shore Chuetsu region, Niigata prefecture. Source:
http://www.jishin.go.jp/main/chousa/07jul_chuetsu_oki/index.htm (in Japanese)
A4-1

reported. Fifteen cases out of them are reported to be relevant to radioactive materials, but no cases
have caused concerns of environmental impacts.
Thorough investigations are due for the reactor pressure vessel internals and other major
safety-related components. Comprehensive evaluation of the impacts due to the Earthquake should
follow incorporating such investigation results. The NSC will keep its awareness of the progressfor necessary evaluation, receiving the relevant reports from NISA as their evaluation develops.
2. Ensuring seismic safety(1) Requirements of the revised Seismic Safety Design Examination Guide and safety checks ofexisting nuclear power plantsa) Requirements of the revised Seismic Safety Design Examination Guide
The NSC revised in September last year (2006) the “Regulatory Guide for Reviewing Seismic
Design of Nuclear Power Reactor Facilities (Seismic Guide).” The revised Seismic Guide (theNew Seismic Guide) requires the operators: (1) to investigate in detail the conditions of the active
geological faults using latest technologies; (2) to analyze the ground motions with latest methods;
and (3) to upgrade the formulation of the “Earthquake ground motions without the site specificepi-center.” In doing so, the New Seismic Guide requires the operators to define the groundmotions bigger than those by the earlier version of the Seismic Guide, using the latest knowledge
and data, and to ensure the safety functions thereto.
b) Seismic safety checks of existing nuclear power plants (back checks)
Upon authorization of the New Seismic Guide in September last year (2006), the NSC requested
through NISA all existing nuclear power plants, which had been designed based on the earlier
version of the Seismic Guide, be reevaluated for seismic safety, referring to the New Seismic
Guide (the so-called “Back checks”) by the operators. In response, the back-check work is ongoingby the operators, and some of their results are being reviewed by NISA.
In this reevaluation processes, the complete and earliest checks are important, pursuant to the
New Seismic Guide, on the formulation of the Design Basis Earthquake Ground Motion (DBEGM),
reliabilities of the analytical models used in the design, and the use of latest knowledge madeavailable after the original design. The operators’ evaluation results are subject to the review by
NISA and further by the NSC.
c) The effectiveness of the New Seismic Guide
It is important to make concluding judgment in the seismic safety, without any prejudges, based
upon the scientific knowledge and facts. The necessity of revising again the New Seismic Guide
should be judged after the new DBEGM be formulated in the back-check process and verified
against the actual impacts due to the Earthquake. It is not the time to contend its necessity now.

A4-2

The NSC will judge its necessity appropriately, in view of these verification results as well as theexternal experts’ opinions.
(2) Detail understanding of shakes due to the Earthquake and additional investigation of geologicalfaults around the plant site
The Earthquake has recorded at the Kashiwazaki-Kariwa Nuclear Power Station theinexperienced tremors far exceeding the values assumed in the original design. The operator should
disclose the detail data, as early as possible, relevant to the Earthquake (seismometer recordings,etc.). The NSC will receive the report at its “Project Team on Seismic Safety Investigation(Seismic Safety PT, founded on 5 July 2007)” for necessary evaluation, as soon as the relevant
information is made public.
Detailed investigation is needed concerning the active fault(s), which broke out the Earthquakeat the Kashiwazaki-Kariwa Nuclear Power Station. As soon as the TEPCO investigation plan isfixed on the geological faults and seabed formations around the site, the NSC will receive thereport at its Seismic Safety PT for its evaluation.
(3) Actions to take at all existing nuclear power stationsa) Confirmation of supporting capabilities of structures and systems
The New Seismic Guide prescribes in its basic policy “to install structures and systems on the
ground with sufficient supporting capabilities.” Namely, it requires them to be installed on the
ground with sufficient supporting capabilities for design loads depending on their classes ofimportance, contrary to the earlier Seismic Guide, which had the similar requirements only to thesafety important structures and systems. However, the uneven ground settlement due to the
Earthquake has damaged quite a number of systems, components, piping and ducts, etc.
The NSC requests TEPCO to take necessary measures to meet the requirements of the New
Seismic Guide, including the possible foundation improvement or reinforcement, for the structuresand systems of Seismic Classes B and C in addition to those of Class S, after identifying the realdamages due to the Earthquake.
This request should be applied to all existing nuclear power units in the back-check processes,
not limited to those of the Kashiwazaki-Kariwa Nuclear Power Station.
b) Earliest back-checks and the disclosure of the results
The TEPCO back-check program made available in October last year (2006) foresees thecompletion of the work for the Kashiwazaki-Kariwa Nuclear Power Station by December next year
(2008). However, the NSC requests the operators review their back-check programs and advancetheir geological investigation and formulation of the DBEGM. The NSC will receive the report on
the results at its Seismic Safety PT for evaluation.

A4-3

Further, for the Kashiwazaki-Kariwa Nuclear Power Station, the NSC thinks it important todisclose the results at the earliest practicality. The NSC requested NISA at its extraordinary session
on 17 July 2007 to instruct the operator to submit the evaluation results as soon as individual work
packages are complete.
c) Setting up of seismometers and record-taking of seismic data
Seismometers installed at every reactor building at the Kashiwazaki-Kariwa Nuclear Power
Station after the“Niigata Chuetsu Earthquake in 2004 (Mid Niigata Prefecture Earthquake in2004 2 )” could collect a lot of valuable data. In the meantime, some of collected data were lost
regrettably despite the experience at the Noto Hanto Earthquake in 2007 3 . The seismic data are
extremely valuable in improving seismic safety of all nuclear power stations. It should not belimited to the safety checks for the subject specific earthquake. Adequate measures by all theoperators are requested thereto. The NSC Seismic Safety PT will also check the status ofseismometer installations at each nuclear facility and data keeping measures to prevent their losses.
d) Preparation of revising the safety examination guidelines concerning geologic and groundconditions
The “Safety Examination Guidelines Concerning Geologic and Ground Conditions (Geologic
Guideline)” identifies the items to examine, concerning the geologic and ground conditions of thesite for locating the nuclear reactor, when the safety examination is conducted pursuant to the
Seismic Guide. The Geologic Guideline is being prepared for revision, as specified in the“Revision of relevant clauses on seismic guidelines for power generating nuclear installations (anearlier NSC decision on 19 September 2006). To this end, relevant information is being collected
and analyzed. The work will be advanced and the preparatory revision work will be initiated at an
appropriate timing for incorporating latest knowledge.
e) Earliest incorporation of latest knowledge
New knowledge obtained from the Earthquake should be evaluated at the earliest practicality
and be incorporated in the back-check processes, as needed, including the lateral development toother existing nuclear power units.

2

On October 23, 2004, at 17:56 (JST) there was a M6.8 earthquake with a maximum seismic
intensity 7, at a depth of approximately 10km in the Chuetsu region, Niigata prefecture. Source:
http://www.jishin.go.jp/main/index-e.html
3
On March 25, 2007, at 09:42 (JST) there was a M6.9 (preliminary) earthquake at a depth ofapproximately 10km near the west coast of the Noto Peninsula, Ishikawa prefecture. Source:
http://www.jishin.go.jp/main/index-e.html
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(4) Evaluation of “Residual risks”
The New Seismic Guide requests, in its commentary to the basic policy, the operators to pay due
attention to the “residual risks (the risks of system damages causing dispersion of radioactive
materials and radiation exposure of the public, due to the ground motions exceeding the assumedDBGM)” and to make efforts to minimize them to the maximum practical extent. Upon request, theoperators are currently evaluating them, as part of their back-check processes.
Probabilistic risk assessment of the “residual risks” still has rooms for future development for
applications. But the quantitative evaluation is requested to the operators on the test trial, which
will advance and facilitate future full applications.
(5) Strengthening of safety research relevant to seismic safety
Operators, regulatory bodies and other research institutions are requested to strengthen andreinforce their research programs on seismic safety. In particular, precision improvement in thefault investigation in the ocean and on the land, or that in predicting the magnitudes of earthquakes
should be advanced. Collaboration with the Headquarters for Earthquake Research Promotion 4
would be more than important.
The NSC will hold a nuclear safety research forum “Seismic safety and safety research” for theexchange of relevant information and knowledge as well as opinions on future research needs forseismic safety.
3. Trouble shooting of fires, etc. at earthquakes(1) Trouble shooting frameworks for fires, etc. at earthquakes
The fire of the Unit 3 transformer at the Earthquake developed safety concerns among the public
because of insufficient effectiveness of the private fire brigade, unavailability of fire control
systems, and delayed notification to the external fire station, and as its consequence a lot of timeneeded to bring the fire under control. The operators are requested to establish a system, in which
any necessary equipment and manpower are available at any time including holidays and nights, in
preparation for unavailable assistance from external sources. Such systems should be prescribed in
the respective safety rules of the operators. The NSC will conduct relevant subsequent regulation
reviews in due course.
(2) Strengthening of fire protection measures at earthquakes
The Regulatory Guide for Reviewing Fire Protection of Light Water Nuclear Power Reactor
4

A special governmental organization, attached to the Ministry of Education, Culture, Sports,
Science and Technology, was established in accordance with the Special Measure Law on
Earthquake Disaster Prevention in July 1995 in the wake of the Great Hanshin-Awaji Earthquake
Disaster.
http://www.jishin.go.jp/main/index-e.html
A4-5

Facilities (the Fire Protection Guide) requires the fire control systems to be designed so as not tolose seriously its capabilities under concurrent earthquake conditions, depending on the classes ofsafety importance of reactor facilities. The NSC takes the preparatory actions to revise the Fire
Protection Guide, taking note that the fire control systems did not function properly at the
Earthquake.
4. Reporting system and information dissemination upon troubles(1) Notification to the central and local governments and publicity
People point out that the notifications from the operator to the central and local governments
were delayed, and the publicized contents were not easily understandable to the public. The NSC
requests the operators and NISA to reconsider the notification and publicity system for
improvement.
(2) Accountability to the nation
In order to mitigate nuclear safety concerns among the public, most important is for theoperators and NISA to recover the public trust and foster their understanding. Their activities inthis regard contribute to the further improvement of nuclear safety. The constant activities by theoperators and NISA are requested to ensure the transparency of information, and to disseminate to
the public relevant information on radiation safety. The NSC also accounts for proactively to the
public on the ensuring of seismic safety.
(3) Information sharing internationally
NISA as well as the NSC have been promoting the information sharing with the IAEA and othernations. It is our nation’s responsibility as one of the most earthquake-ridden countries todisseminate the new knowledge learned from the Earthquake to the world for improving safety.
Japan shares the lessons with the IAEA delegation of investigation. NSC also takes actions todisseminate relevant information internationally.
5. Conclusions
The NSC is in a position to avoid any prejudges on nuclear safety, especially in the seismic
safety. The NSC prioritizes the open-mind and learning attitudes of placing primary importance inthe scientific knowledge and facts. The NSC adheres to its philosophy.

A4-6

Annex 5

Answers to the Synopsis of the Relevant IAEA Safety RequirementsiINTRODUCTION

In the third Review Meeting held in April 2005, the IAEA Secretariat was requested torelate their Safety Requirements to each article of the Convention. In response to this,the Secretariat formulated a synopsis in the Q & A style by referencing to the Safety
Requirements Publications listed below and allocating the requirements to thecorresponding articles. This document’s purpose consists of our reply to the synopsis
based on the Safety Requirements, which reflects its current status in Japan.
The original objective of the synopsis is to be used as a reference when preparing the
National Report and for the Review Meeting, and may be used at the discretion of eachState.
IAEA SAFETY REQUIREMENTS PUBLICATIONS

Reference
Number
GS-R-1
GS-R-2

Title
Legal and Governmental Infrastructure for Nuclear, Radiation,
Radioactive Waste and Transport Safety
Preparedness and Response for a Nuclear or Radiological Emergency

GS-R-3(DS338)
NS-R-1The Management System for Facilities and Activities

NS-R-2

Safety of Nuclear Power Plants: Operation

NS-R-3Site Evaluation for Nuclear Installations

SS115

International Basic Safety Standards for Protection against Ionizing
Radiation and for the Safety of Radiation Sources

Safety of Nuclear Power Plants: Designii

CONTENTS

Article 7: Legislative and Regulatory Framework ····················································· 1
Article 8: Regulatory Body···················································································· 5
Article 9: Responsibility of the Licensee ·································································· 8
Article 10: Basic Policy for Priority to Safety·························································· 10
Article 11: Financial and Human Resources··························································· 11
Article 12: Human Factors·················································································· 14
Article 13: Quality Assurance·············································································· 15
Article 14: Assessment and Verification of Safety ···················································· 19
Article 15: Radiation Protection··········································································· 21
Article 16: Emergency Preparedness ···································································· 22
Article 17: Siting······························································································· 26
Article 18: Design and Construction ····································································· 29
Article 19: Operation ························································································· 32iiiThis page intentionally left blankiv

Article 7: Legislative and Regulatory Framework

GS-R-1,
2.2

GS-R-1,
2.4

Question

Reply(1) What legislative and governmental
mechanisms are in place that define
national
requirementsfortheregulation of the safety of facilities andactivities?

Concerning commercial power reactors, which are the subjectof this IRRS review, the safety of their facilities and activities
is regulated by the Reactor Regulation Law, the Electricity
Utilities Industry Law, and other laws. Other facilities andactivities are also regulated by the Law for the Regulation of
Nuclear Source Material, Nuclear Fuel Material and Reactors,
and other laws.
For an overview of these regulatory measures, please refer tothe Article 7-related part of Japan’s National Report under theConvention on Nuclear Safety.
The details of the substance and procedures are determined bycabinet order, ministerial order, ministerial notice andministerial instruction.

(2) How is the Regulatory Body
established?

The Ministry of Economy, Trade and Industry(METI), one ofwhose assigned duties is the safety regulation of nuclear
power, is established under Law for Establishment of the
Ministry of Economy, Trade and Industry (hereinafter referred
to as the “METI Establishment Law”) (cf. Article 2), and the
Nuclear and Industrial Safety Agency undertakes the assigned
duties (cf. Article 20). Concerning the other Ministries'
jurisdiction, they are defined in each Ministry's Establishment
Law.

(3) How is the responsibility forlicensingnuclearinstallations;
regulatory review and assessment;
inspection and enforcement assigned
to the regulatory body?

The METI Establishment Law states the mission of the
Ministry of Economy, Trade and Industry (METI) as follows:“The Ministry of Economy, Trade and Industry shall engage in
the growth of [a] economy and industries with emphasis on theenhancement of the economic vitality of the private sector andthe harmonized development in Japan’s external economic
relations and [b] shall engage in ensuring stable and efficient
supply of mineral resources and energy” (Article 3). To fulfill
this mission, METI is assigned various duties (Article 4). Oneof these assigned duties is: “[a] matters relating to theregulations for [i] the refining, fabrication, storage,
reprocessing and waste disposal business in nuclear fuel cycleand [ii] the nuclear power installations and [b] mattersrelating to ensure the safety of these businesses andinstallations”(Article 4, paragraph 1, subparagraph 57).
According to the Article 20, paragraph 3, NISA is responsible
for the duties.

(5) How is it ensured that there are no
responsibilities assigned to theregulatory body that may jeopardize or
conflict with its responsibility for
regulating safety?

According to the METI Establishment Law, Article 20,
paragraph 3, NISA is responsible for the duties listed in
Article 4, paragraph 1, subparagraphs 57 through 59, 62 and
64. As duties other than nuclear safety regulation, securingindustrial safety (mining safety; gas safety; thermal, hydroand other power generation safety; and so on) is included.
However, the duties for securing industrial safety is executed
under separate laws, and the divisions responsible for the
execution of such duties are separated from the divisions
responsible for nuclear power related matters. Thus, executing
such duties does not jeopardize nor conflict with the
responsibilities of NISA concerning nuclear power safety
regulation. Moreover, NISA has established common
institutional objectives (securing the safety of the lives of theJapanese people and protecting the environment) and code of
conduct (strong sense of duty, scientific and rational judgment,
transparency in execution of duties, neutrality and fairness)and fulfills its duties accordingly.

How does the legislation:(1) Set out objectives for protecting
individuals,
societyandthe

The Reactor Regulation Law, in Article 1, stipulates that theobjective of said law is to conduct the necessary regulation,1

environment from radiation hazards,
both for present and in the future?

etc., in order to prevent hazards due to nuclear source
material, nuclear fuel material and reactors and to ensure
public safety and so on.

(3) Establish an authorization process?

The Reactor Regulation Law (including the Electricity Utilities
Industry Law in the case of commercial power reactors) sets
forth regulation, taking into consideration the potential
magnitude and nature of the hazard, for each of the following
categories: nuclear facilities and refining, fabricating, storage,
reprocessing, and waste management and disposal. For
example, in the case of commercial power reactors, the Reactor
Regulation Law and the Electricity Utilities Industry Law sets
forth regulation explicitly for each step along the way, fromlicensing of establishment, approval of construction plan,
approval of operational safety program, approval of
decommissioning program, and confirmation of the completionof decommissioning.

(7) Establish a procedure for review of,
andappealagainst,regulatory
decisions(without
compromisingsafety)?

According to the Administrative Complaint Investigation Law,a disposition under the Reactor Regulation Law and theElectricity Utilities Industry Law can be appealed against the
Minister of METI, who has made the disposition. Also,
according to the Code of Administrative Procedure, a lawsuit
to rescind the disposition can be lodged. The lodging of an
appeal or a lawsuit does not in principle impede the effect, theexecution, or the continuation of the disposition, so the lodging
of an appeal or a lawsuit will not endanger nuclear safety.

(8)
Provideforcontinuityof
responsibility when several successive
licence holders carry out activities?

Concerning a commercial power reactor, someone who intends
to take over such a reactor from the reactor establisher,according to the stipulation by the Reactor Regulation Law,
must receive a license from the Minister of METI. The one who
has received such license and taken over the reactor assumes
the legal status of the reactor establisher. Also, when thereactor establisher enters into a merger,
upon receiving alicense from the Minister of METI, the legal entity surviving
the merger or the legal entity to be incorporated upon themerger shall assume the legal status of the reactor establisher.
In these cases, the transfer of the responsibilities of the reactor
establisher shall be recorded as part of the licensing or
approval procedures.

(9) Allow for the creation of
independent advisory bodies toprovide expert opinion to, and for
consultation by, the government andregulatory body?

The METI Establishment Law stipulates that the Advisory
Committee for Natural Resources and Energy shall be placedin the Agency for Natural Resources and Energy. The Advisory
Committee for Natural Resources and Energy surveys anddeliberates important matters on consultation by the METI
Minster and gives its opinion to the Minister of METI.
Concerning nuclear safety, there is a Nuclear and Industrial
Safety Subcommittee under the Advisory Committee for
Natural Resources and Energy. The Advisory Committee andits Subcommittee are state agencies. Therefore, their
independence from undertakers and their associations isassured.

(10) Set up a means whereby researchand development work can beundertaken in important areas ofsafety?

In order to enhance safety regulations, regarding scientific andrational aspects, and utilizing most advanced knowledge,
NISA newly established "Subcommittee on Fundamental
Policies for Nuclear Safety Infrastructure" in 2006 as one of
the branch subcommittees of "The Nuclear and Industry
Safety Subcommittee". And this "Subcommittee on
Fundamental Policies for Nuclear Safety Infrastructure" isexpected to express expert opinions and advices to NISA
through reviewing needs for nuclear safety researches anddevelopment (hereafter referred to as “nuclear safety
research”), verifying agenda of nuclear safety researches andetc.. Considering such opinions and advices, NISA is to conduct
nuclear safety researches in accordance with real needs and to
feedback the results of nuclear safety researches to regulatory2

process.
In addition, in order to conduct research and development inareas important to nuclear safety in an organized andprioritized manner, the NSC establishes approximately every
five years a promotion program for nuclear safety research andcirculates it among the relevant organizations. Most recently,
the July 2004 “Prioritized Nuclear Safety Research Program”
covers 2005-2009 and summarizes the research areas thatdeserve priority and the promotional measures thereof. Basedin this program, NISA provides grants to the Japan Nuclear
EnergySafetyOrganization(JNES;
incorporated
administrative agency) and commission fees to the Japan
Atomic Energy Agency (JAEA; incorporated administrative
agency).
(14) Define what is an offence and thecorresponding penalties?

Concerning commercial power reactors, the Reactor
Regulation Law stipulates what kind of infractions can lead tothe revocation of a license for the establishment of a nuclear
power reactor or the suspension of use of a nuclear powerreactor. The Law also stipulates what kind of infractions canlead to criminal penalties.

(15) Implement any obligations underinternational treaties, conventions oragreements?

Concerning treaties, the Japanese Constitution states: “Thetreaties concluded by Japan and established laws of nations
shall be faithfully observed.” Also, the Vienna Convention on
the Law of Treaties states: “Every treaty in force is binding
upon the parties to it and must be performed by them in goodfaith.” Therefore, Japan performs its obligations under treatiesthat it has entered into by adopting new laws or amending
existing ones.
When issuing orders and other rulings under a law, according
to the Administrative Procedure Law, it is necessary to seek
public comment from the general public. Concerning specificdispositions by regulatory agencies, such as license forestablishment of an commercial power reactor, according to
the Administrative Complaint Investigation Law or the Codeof Administrative Procedures, a party with legitimate interests
can lodge an appeal or law suit to rescind such disposition.

(16) Defines how the public and otherbodies are involved in the regulatory
process?

Among other matters, the relationship with otheradministrative organizations, such as the NSC, to whom the
Minister of METI reports the status of its regulatory activity,
and the relationship with other institutions, such as JNES,
which conducts Fuel Assembly Inspection, are stipulated in
the Reactor Regulation Law and the Electricity Utilities
Industry Law.
(17) Specify the nature and extent of
the application of newly establishedrequirements to existing facilities andcurrent activities?

When introducing new regulation, including laws concerning
nuclear safety regulation, Japanese legislation in generalenacts transitional measures in a “supplementary provisions”
to stipulate how existing facilities and ongoing activities shallbe treated. Also, when determining the date of enforcement,
we make sure that there will be sufficient time to inform thepublic before the new regulation is enforced.

GS-R-1,
3.2

How does the regulatory bodyestablish regulations and guides, andassessment principles and associatedcriteria upon which its regulatory
actions are based?

Ordinances are issued in the name of the Minister of METI,
and documents are issued in the name of the Director-Generalof NISA. NISA strives to make them know and understandthoroughly, and to apply them strictly. For details, see Article
7 of this report.

GS-R-1,
3.3

How does the Regulatory Body:(9) Ensure that its regulatory
principles and criteria are adequate
and take account of international
standards and recommendations?

In deliberating the transition to performance-based technicalstandards for nuclear power generation equipment and othermatters, NISA considers internationally endorsed standards
and recommendations, for example by seeking to harmonize
them with international standards, taking into considerationsuch standards as the IAEA Nuclear Safety Standards (NUSS)
and the US NRC 10CFR.
Also, in developing its regulatory principles and criteria, NISA

3

follows the safety examination guidelines established by the
NSC, and the regulatory principles and criteria of the safety
examination guidelines are based on the standards andrecommendations
of
ICRP
etc.
as
needed.
In addition, the Radiation Review Council, in the efforts
toward the domestic assimilation of the International
Commission
onRadiologicalProtection(ICRP)
Recommendation of 1990 (Pub.10), has issued the “TechnicalStandards Concerning the Assimilation of the New
Recommendation (pub.60) of the International Commission on
Radiological Protection”, and otherwise takes international
standards and recommendations into consideration whendeliberating domestic countermeasures for radiation
protection.

4

Article 8: Regulatory Body
8.1

GS-R-1,
3.1

GS-R-1,
4.1

Question

Reply

How does the Regulatory Body define
policies on its regulatory actions?

NISA, while recognizing that the primary responsibility of
nuclear safety belongs to the operator, upholds the following
three as regulatory ideals:
1. Safety regulation is clear and open to the public.
2. Safety regulation is effective regulation that reflects thelatest scientific knowledge and experience.

How is the Regulatory Body
structured to discharge its
responsibilities?

3. Safety regulation works proactively with international
trends.
Please refer to the Article 8 (8.3 (2)).

How is the structure and size matchedto the extent and nature of facilities
and activities it regulates?
GS-R-1,
4.2

If the Regulatory Body consists ofmore than one authority, whatarrangements are there to ensure that
duplication or omissions are avoided
and conflicting requirements are notplaced on the license holder?

The Nuclear and Industrial Safety Agency has sole
responsibility for regulating commercial power reactors.
In terms of the relationship with the other authorities
concerned, see 8.7 of this report.

How are its main functions organized
to ensure consistency and to enable
feedback and exchange of information?
GS-R-1,
4.3

Is the Regulatory Body self-sufficientin all technical and functional
expertise?
If not, how does it seek advice orassistance that is independent of thelicense holder?

GS-R-1,
4.5

How has the Regulatory Body
established and implemented
arrangements for quality
management?

GS-R-1,
4.6

How does the Regulatory Body ensurethat it employs a sufficient number of
personnel with the necessary skills toundertake its functions andresponsibilities?

GS-R-1,
4.7

How does the Regulatory Body ensurethat its staff has the relevant
competencies?
What education and training program
does the regulatory body have for its
technical and professional staff?

Regarding review and assessment, NISA is entirely
self-sufficient, while it sometimes has JNES, established by
the state based on the legislation, conduct computer analysis
in order to confirm the appropriateness of computer analysis
submitted by an applicant concerning anticipated transientsduring operation, accidents, radiation exposure, etc., asrequired. Regarding inspection, NISA is also entirely
self-sufficient. Inspection is conducted by NISA or JNES, notconsultants.
NISA solicits views of experts in the Nuclear and Industrial
Safety Subcommittee as necessary.
NISA spares no pains to disclose information to the public,
accept third-party judgment, and has continuously worked to
improve its regulation activities by continuously reviewing its
activities and institutional framework.
Moreover, in January 2007, in order to ensure further
improvement in the quality of its regulations andtransparency and fairness thereof, NISA has have establishedthe “Quality Management Manual” in compliance with the
IAEA GS-R-3 management standards, and have begun
executing its operations accordingly.
NISA has approximately 800 officials, of whom about 330 are
responsible for nuclear safety.(as of end of January 2007)
NISA provides training so that they can acquire sufficientcapabilities required to execute their duties.
As shown in Table 11-1 (Article 11), in order to have NISA
officials acquire and maintain the knowledge and skills toappropriately execute examination and inspection activities,
NSA has established a nuclear safety human resources
development program. Under this program, officials undergo
training according to their levels of knowledge and skill sothat they can acquire the proper skills, knowledge aboutstate-of-the-art technology, and new principles and concepts
that are required in regulatory activities. Specifically, NISA

5

GS-R-1,
4.8

GS-R-1,
4.9

How does the Regulatory Body ensurethat their staffs have sufficientexpertise to either perform regulatory
reviews directly, or evaluate the workof consultants?

Does
the
regulatory
bodyorGovernment use advisory bodies togive independent advice?
How is it ensured that the advice does
not relieve the regulatory body of its
responsibilities to make decisions andrecommendations?

GS-R-1,
4.11

How are arrangements established forthe exchange of safety related
information, bi-laterally or regionally,
with relevant intergovernmentalorganizations to fulfill safety
obligations and promote cooperation?

actively promotes human resources development for nuclear
safety administration through special, dedicated training by
the Nuclear Safety Training Office in the Training Institute of
Economy, Trade and Industry, the training institution for
METI officials, as well as through on-the-job-training at 21
Nuclear Safety Inspector Offices around the country,
long-term training at JAEA, overseas training at safety
regulation agencies in the US, UK and France, and othermeans.
Safety Examiners in the Nuclear Power Licensing Divisionand other divisions undertake examination and evaluation
activities as their sole responsibilities.
Senior experts are assigned to core examination positions suchas Senior Safety Examiner and a leader of examiners.
Meisters are assigned to positions that oversee the wholeexaminations and hold responsibility for them such as
Director for Safety Examination. It requires in generalapproximately 10 years to become a senior expert.
The Nuclear and Industrial Safety Subcommittee has been
established in the Advisory Committee for Natural Resources
and Energy as the institution to deliberate and makeproposals concerning nuclear safety regulation. This isestablished to hear the high-level and specialized opinions ofexperts, and a Cabinet Decision states that ultimate
policy-making decisions are made under the responsibility ofadministrative organs.
Japan does the following to exchange safety-related
information.
I. Bilateral and regional level:
Japan has bilateral nuclear agreements for the peaceful use of
nuclear energy with the United States, France, Germany,
Sweden, Republic of Korea, China, Australia, Canada, the UK
and Russia. Japan also has bilateral agreements concerning
nuclear power generation safety with the United States,
France, the UK, Republic of Korea and China, and conductperiodical meetings for the exchange of safety-related
information.
II. Neighboring states and other interested states
Japan is a party to the Convention on Early Notification of a
Nuclear Accident and the Convention on Assistance in theCase of Nuclear Accident or Radiological Emergency.
Japan and China shall promptly notify each other in the caseof serious accidents concerning nuclear power generation
plants under an agreement with the objective of promoting thesafety level of commercial power reactors. Japan and theRepublic of Korea shall establish an early notification networkand conduct cooperation activities based on intergovernmental
consultations. Moreover, Japan is consulting with China andthe REPUBLIC OF Korea with the objective of further
enhancing nuclear power generation safety in Northeast Asia
as whole measures, such as the development of a framework
for regional cooperation, to reinforce coordination between the
nuclear power safety regulatory agencies in Northeast Asia
nuclear power generating countries.

8.2
GS-R-1,
4.1

How does the Regulatory Body
maintain its independence in thegovernmental infrastructure?

The Nuclear and Industrial Safety Agency (NISA) wasestablished on January 1, 2001, within the reorganization ofnational administrative organs as a “special organization” at
the Ministry of Economy, Trade and Industry, anadministrative organ of the Japanese Government, to conductsafety regulation of nuclear. NISA is effectively independent
from the Agency for Natural Resources and Energy, which is6responsible for nuclear power technology development andfacilities and so on, through mechanisms such as inter alia
provisions in the decision-making mechanisms andsupervision and auditing by the Nuclear Safety Commission(the NSC), and this independence of NISA is secured.
GS-R-1,
2.2(2)

How is it ensured that the regulatory
body is effectively independent oforganizations or bodies charged withthe promotion of nuclear technologies
or responsible for facilities or
activities?

Same as above.

7

Article 9: Responsibility of the Licensee

GS-R-1,
2.3

Question

Reply

How is the prime responsibility forsafety assigned to the license holder
for siting; design; construction;commissioning and operation?

It is stipulated in laws and ordinances that the operator has
the primary responsibility for the safety concerning acommercial power reactor.
1) Reactor Regulation Law
When there is intent to establish a nuclear reactor, theoperator must receive a license from the Minister of METI
for matters including siting, and the securing of safety inbasic design and basic design policies.

How do legislative and governmental
mechanisms ensure that compliancewith the requirements imposed by theregulatory body does not relieve theoperator [licensee] of its prime
responsibility for safety?
How does the operator [licensee]
demonstratetotheregulator’ssatisfaction that this responsibility forsafety has been and will continue to bedischarged?

The operator must establish operational safety program(including the details of the operation and control of nuclear
reactor, operation limits, and safety education) and havethem approved by the Minister of METI before starting
operations.
The operator has the obligation to take measures necessary
to ensure safety concerning each of the following:(1) Maintenance of reactor facilities;(2) Operation of reactors; and(3) Transportation, storage, and disposal of nuclear fuelmaterial or material contaminated by nuclear fuel material.
Where there is a violation of these requirements, theMinister of METI can issue orders including suspending useof reactor facilities.
2) Electricity Utilities Industry Law•
When one intends to conduct construction of an electric
structure (includes nuclear reactor facility; same
hereafter), one must receive the approval of the
Minister of METI for the construction plan for electric
structures including items concerning the securingsafety in detailed design and during construction. Also,
the operator may not use the electric structure until it
has been inspected and certified by the Minister of
METI.
•
The operator must maintain an electric structure sothat it is in compliance with technical standards. If
there is an infraction, the Minister of METI can issueorders including suspending use of that electric
structure.
In this way, the undertaker has been assigned the primary
responsibility for safety regulation of a commercial powerreactor.
Beyond these measures, the system for the operator to
regularly review the operator’s organization for theoperator’s review of the main facilities (Electricity Utilities
Industry Law) and the system for the review of theobservance of operational safety program (Reactor
Regulation Law) require that the primary obligation of theoperator are ensured and that they will continue to beappropriately maintained in the future.

GS-R-1,
2.1

How does the operating organization
as licensee retain prime responsibility
for safety when it delegates authority
to the plant management for the safe
operation of the plant?
In such cases what resources andsupportdoes
theoperatingorganization provide for the plant
management?

GS-R-1,How does the operating organization

The operator, in taking measures required for safe operationof reactor facilities, must establish a quality assurance
program to comply with the national requirements andclarify the organization in charge of its implementation. The
quality assurance program must include the provisions
concerning “human resources”, “nuclear installations” and“work environment”.

The licensee has the primary responsibility for the safety of

8

3.1

with overall responsibility for safety,
ensure that interfacing organizations
engaged in activities important tosafety meet their responsibility toensure that safety matters are given
the highest priority?

nuclear installations and must observe legislation, etc.
provided by the regulatory body. For details, see 9.1 of thisreport.

GS-R-1,2.14 - 2.18

How is the interface between theoperator and the regulatory body withregard to the responsibility for safety
organized?

NISA strives to have sufficient opportunities to exchange
opinions with licensees in order to have them understand its
views on safety regulation and also to see their views on it.
For details, see 9.3 of this report.

9Article 10: Basic Policy for Priority to Safety
Question
NS-R-2, 2.2;
DS338, 2.2

What are the policies of the operatingorganization giving safety matters thehighest priority?

Reply
The operator, in taking measures required for safe
operation of reactor facilities, must establish a quality
assurance program to comply with the national
requirements, give the highest priority to safety mattersin the program and clarify the policy in “Commitments” inthe quality assurance program.

NS-R-2, 2.6

How does the operating organization
make sure that its safety policy isapplied by all site personnel?

In order to ensure safety of nuclear installations, safety
culture in organization is vital. Licensees take various
efforts to establish safety culture. For details, see 10.2 ofthis report.

NS-R-2, 2.3

How are management objectives set
and how are these objectives related tothe policy for nuclear safety andquality?

The operator, according to the quality assurance program
giving the highest priority to safety matters, must
establish the quality goal to fulfill the work requirementsin each division and level in the organization.

NS-R-2, 2.3

How is safety monitored and followed
up on a regular basis, timely corrective
actions taken and opportunities for
improvements used?

NS-R-1, 3.1

How is it ensured that the designorganization takes into account thecurrent state of the art for safety, andthat the safety of any design change isproperly considered?

The operator, according to the quality assurance program
giving the highest priority to safety matters, must
implement management review to monitor and identify
non conformity and confirm the implementation of
preventive measures and corrective measures.
The operator, according to the quality assurance program
giving the highest priority to safety matters, must
implement review, validation and verification in analteration management process of design and receive anapproval prior to its implementation.

NS-R-2, 2.5

How are proposed changes to the
management structure and associatedarrangements which might besignificant to safety systematicallyreviewed by the operatingorganization and submitted to theregulatory body for review?

The items concerning operations management are
specified in the operational safety program and analteration to them must be submitted to and reviewed by
the regulatory body.

NS-R-2, 2.6

How are clear lines of authority
established to deal with plant safety
matters?

Operational Safety Program states matters relating toduties and organizations, and clarifies the authority todeal with plant safety matters.

NS-R-2,2.10

How is it ensured that all activities
that may affect safety are performedby suitably qualified and experienced
persons?

When licensing for establishment, technical capabilities oflicensees are reviewed. Duties, education and capabilities
required are stipulated in Operational Safety Program
and the quality assurance program.

NS-R-2,2.11

How is it ensured that all activities
that may affect safety and which canbe planned in advance are conductedin accordance with establishedprocedures?

As described in 19.3 of this report, the operational safety
program covering such activities is established by theoperator and reviewed and approved by the regulatory
body. Then, to confirm observance of the operational safety
program, the operational safety inspection is implemented
by the regulatory body.

NS-R-2,
2.12

What are the procedures to deal withactivities that are not included in thenormal procedures?

As described in 19.3 of this report, a special review
committee is established to examine any off-normal
activity such as alteration of an operational safety
program.

NS-R-2,2.13

How is it ensured that an appropriate
safety consciousness and safety
culture prevail in plant operations?

Operational Safety Program states matters relating toduties and organizations, and clarifies the authority todeal with plant safety matters.

10Article 11: Financial and Human Resources
11.1

DS338,
4.1

Question

Reply

How does the licensee determine the necessary
material and financial resources to carry out theactivities of the organization?

When issuing license of a nuclear installations,financial basis is reviewed.
The operator, in taking measures required for safe
operation of reactor facilities, must establish aquality assurance program to comply with thenational requirements and clarify the organizationin charge of its implementation. The quality
assurance program must include the provisions
concerning “human resources”, “nuclear
installations” and “work environment”.
The operator, according to the Electricity Utilities
Industry Law, etc., deposit reserves for waste
storage outside of reactor facilities anddecommissioning activity, as described in 11.1 (2) ofthis report.
The licensee, according to the quality assurance
program giving the highest priority to safety
matters, must clarify the competence required forthe personnel, implement education and training
and assess them.

How are financial resources made available toperform safety improvements?
How are financial resources made available tocope for any waste management activities
resulting from the operation of the facility?
How are financial resources made available tocope for decommissioning activities after thetermination of the operation of the facility?

DS338,
4.3

How are competence requirements determined
for individuals at all levels?

The regulatory body classifies the competence
requirements for staff into four levels (Entry,
Expert, Senior Expert and Meister). The regulatory
body specifies the contents of study and training
required for the achievement of each level ofcompetency, and implements it.
11.2
DS338,
4.3

How does the licensee provide training or takesother actions to achieve the required level ofcompetence?

NS-R-2,
3.1, 3.4

How does the license holder define thequalifications and experience necessary for
personnel performing duties that may affectsafety?
What provisions are in place to select suitably
qualified personnel and given the necessary
training and instruction to enable them toperform their duties correctly, including
managerial and supervisory skills?

NS-R-2,
3.3

What programs are in place for training
personnel before their assignment to safety
related duties?

The licensee, according to the quality assurance
program giving the highest priority to safety
matters, must clarify the competence required forthe personnel, implement education and training
and assess them.
The licensee, according to the quality assurance
program giving the highest priority to safety
matters, must clarify the competence required forthe personnel, implement education and training
and assess them.

As described in 11.2 (2), 2) of this report, the items
concerning the education for the operation andmanagement personnel are specified in an
operational safety program.
The relevant legislation and regulation are as
follows.
1. The Rules for the Installation, Operations, etc. ofCommercial Power Reactors, Article 7-3, “Quality
Assurance” stipulates that a licensee must
establish a quality assurance program. To comply
with the requirement, a licensee establishes aquality assurance program in the operational safety
program. The quality assurance program specifies
the operational management of resources including
items such as required competence, education and11training.
2. The Reactor Regulation Law, Article 37,“Operational Safety Program” and the Rules for the
Installation, Operations, etc. of Commercial Power
Reactors, Article 16, “Operational Safety Program”
stipulate that the operational safety program mustspecify the items about an education on operationalsafety for the operating personnel and managementof the reactor installation. To comply with therequirement, the operational safety program
specifies that a plan for education on operationalsafety be established and implemented. The
situation of the implementation is regularly
confirmed in the operational safety inspection by
the regulatory body.
NS-R-2,
3.4

How does the license holder ensure that all
personnel who may be required to perform
safetyrelateddutieshave
sufficientunderstanding of the plant and its safety
features?

The licensee, according to the quality assurance
program giving the highest priority to safety
matters, must make the personnel understand theimplication and importance of their activities andtheir contribution to the achievement of quality
goal through education and training process.

NS-R-2,
3.5

How does the license holder ensure that thequalifications and training of external
personnel performing safety related duties areadequate for the functions to be performed?

The licensee, according to the quality assurance
program giving the highest priority to safety
matters, must evaluate contractors based on
established standards in a procurement process
and conduct necessary inspection to ensure that the
procurement requirements are satisfied.

NS-R-2,
3.6

What provisions are there for periodic
confirmation of the competence of personnel
and for refresher training?

The licensee, according to the quality assurance
program giving the highest priority to safety
matters, must evaluate the effectiveness ofeducation and training.

NS-R-2,
3.7

Who provides the training organization with
the necessary resources and facilities?

The licensee, according to the quality assurance
program giving the highest priority to safety
matters, must clarify the resources required fornuclear safety and provide them. And, the operator
must implement education and training to ensure
the required competence.

Who determines the need for training, andensures that operating experience is taken into
account in the training?
How is it ensured that production needs do notinterfere with the conduct of the training
program and the need for personnel to be
trained?
NS-R-2,3.9

How is it ensured that training instructors arecompetent in their assigned areas of
responsibility and have the necessary
instructional skills?

The licensee, according to the quality assurance
program giving the highest priority to safety
matters, must clarify the competence required for
personnel with duties that may influence nuclear
safety and evaluate its effectiveness.

NS-R-2,
3.11

What simulator facilities are used for the
training operating personnel on operational
states and for accidents?

Please refer to Tables 11-2 and 11-3 (Article 11) of
the Main Report.

NS-R-2,
3.12

What instruction is given to plant staff on the
management of accidents beyond the designbasis?

The licensee must implement an education on
accident management and provide a training coursewith a training simulator.

NS-R-2,
3.13

What is in place to assess and improve thetraining programs, and modify and update thetraining facilities and materials to ensure thatthey accurately reflect plant conditions?

The licensee, according to the quality assurance
program giving the highest priority to safety
matters, must identify duties throughout the wholeprocess from work plan to its implementation
including training, etc., and implement them in acontrolled manner.

NS-R-2,
3.14

How is operating experience of events at theplant and relevant events at other plants
factored into the training program?

The licensee shall, by in-house nonconformity
management process for any abnormal event in own
plant and by the nuclear information library(NUClear Information Archives: NUCIA) for any
information in other plants, examine an action
including any alteration to a training program and12

implement it.
NS-R-2,
2.4

What provisions did the license holder take toestablished liaison with organizations fordesign, construction, manufacturing and plant
operation and with other organizations(national and international) as necessary toensure the proper transfer of information,
expertise and experience to respond to safetyissues? Are the national resources for services
and technical support adequate?

The licensee, according to the quality assurance
program giving the highest priority to safety
matters, acquires information, expertise andexperience from suppliers as a part of procurement
management. In addition, the operator can sharenecessary information through an owners' grouporganized in cooperation with plant manufacturers.

GS-R-1,
5.13, (3)

Are the competence requirements, thequalification,
training
andre-training
activities of the licensee subject to regulatory
inspection?

Specialists such as the Chief Engineer of Reactors,
the Chief Electrical Engineer and the Chief
Engineer of Boiler and Turbine are qualified by thecompetent authorities having a jurisdiction overrespective license. While the operator implements
education of its employees, the state of theeducation and training is confirmed by theregulatory body through the inspection on theobservance of operational safety program.

13

Article 12: Human Factors
Question

Reply

NS-R-1
5.48, 5.49

How are designs made ‘operator friendly’ andlimit the effects of human errors by plantlayout, work areas, working environment andprocedures (administrative, operational andemergency), including maintenance andinspection, in order to facilitate the interface
between the operating personnel and theplant?

As described in Article 12 of this report, humanfactors are taken into consideration indesigning the main control room and in an
operations management aspect.

NS-R-1,
5.50

How is consideration of human factors and thehuman–machine interface take into account,provided?

As described in Article 12 of this report,
human factors are taken into consideration indesigning the main control room and in an
operations management aspect.

NS-R-1,
5.51

How does the human-machine interface
provide the operators with comprehensive,
easily manageable information, compatible
with the necessary decision and action times?

As described in Article 12 of this report, humanfactors are taken into consideration indesigning the main control room and in an
operations management aspect.

How are similar provisions made for the
supplementary control room?
NS-R-1,
5.52

How are verification and validation of aspectsof human factors included at appropriate
stages to confirm that the design adequately
accommodates all necessary operator actions?

As described in Article 12 of this report, humanfactors are taken into consideration indesigning the main control room and in an
operations management aspect.

NS-R-2,
3.2

Is there a program to ensure fitness for duty?

There is no specific regulatory requirement.
The operator must take appropriate measures
timely to assure compliance with related
legislation and an operational safety program
while giving priority to safety.

14

Article 13: Quality Assurance

Question
DS338,
2.1

How does the management systembring together all therequirements for managing thenuclear installations actions toprovide confidence that theserequirements are satisfied andthat quality requirements are notconsidered separately from safety
requirements?”

Reply
Please refer to 7.3, 10.2, 13.1 and 19.3 of this report.
As for the activities of the regulatory body,(1) i) NISA set forth the Management Manual, which
determines NISA's management system in January 2007,
and began administration of its management system.
ii) The Manual adheres to the "Management Policy", which
lays out NISA's objectives, etc. NISA's management
system, which is based on the manual, also adheres to
NISA's organizational objectives.
(2) i) “Work Management Guidelines” which specifies
NISA's management system, referring to objective
evaluation standards such as IAEA (GS-R-3), etc.,
integrate all the requirements to operate and administrate
the regulatory body in a consistent manner.
(3) i) The “Work Management Guidelines” establish theplanned and systematic measures to appropriately
implement the requirements of IAEA (GS-R-3) etc. : forexample, the establishment of “Work Management
Committee” within NISA, with the Director-General as its
head; the establishment of “Work Implementation Plan” forindividual divisions and NISA as a whole; and theestablishment of a mechanism to implement, inspect andassess the duties under the plan.
(4) The management system of NISA is in the early stage of
its operation, to be improved successively as the experience
accumulates.
The operator must establish the quality assurance in theoperational safety program.

DS338,
2.6

How are management system
requirements graded to deploy
appropriate resources relative tothe safety significance, hazards
and risks, and possible
consequences of failure?

As for the activities of the regulatory body,i) Under the Quality Management Manual, which
determines the NISA management system, the “Quality
Management Committee” is established within NISA, with
the Director-General as its head. Also established underthe Quality Management Manual are the “annual workplan” for individual divisions and NISA as a whole. A
mechanism to implement, inspect and assess activities
under the work plans has been established.
ii) Under the “annual work plan”, items for priority
operational improvement from a mid- to long-term point ofview (operational improvement priority) have been selected
in order to prioritize activities. (Quality Management
Manual Chapter 10 Improvement of Work, (2) Priority for
Improvement of Work)iii) Activities are prioritized at the individual officials’ level
as well as through such means as the establishment ofoperational goals in conformance with the “annual workplan”.
iv) Therefore, through the measures such as prioritizationof works in the aforementioned “Work Implementation
Plan”, activities of “Work Management Committee” toinspect and assess the plan and implementation of “workassessment”, a graded classification of duties, which
contributes to the appropriate allocation of resources, canbe realized.
The operator shall, establishing a quality assurance
program giving the highest priority to safety to comply

15

DS338,
3.1

DS338,
3.8

How does management
demonstrate its commitment toimplementation, assessment andcontinued improvement ofmanagement systems, including
allocation of adequate resources?

How does management establish
goals, strategies, plans andobjectives (sometimes known as
the business plan)?

with the national requirements, clarify and assure theresources required for nuclear safety.
As for the activities of the regulatory body,(1) i) Directors and the Director-General each creates an
Annual work plan for his/her division and NISA, andexecutes, assesses, and makes activities improved underthe Annual work plan, based on Quality Management
Manual, Chapter 3 Duty of Organization, (3) Planning; andChapter 5 Work Management.
ii) As members of the Quality Management Committee,
senior management and directors participate in theoperation of the management system and execute their
responsibilities, based on the responsibilities and authority
stipulated by the Quality Management Manual, Chapter 3,
Duty of Organization, (4) Responsibility and authority.
(2) i) In order to facilitate the execution of NISA’s duties,
the Director-General also establishes NISA’s institutionalframework that includes the appropriate allocation of
resources and exercises overall direction and supervision
over the execution of NISA’s duties.
ii) Other senior management members and directors
participate in the operation of the management system
including the appropriate allocation of resources, andexecute such responsibilities as members of the Quality
Management Committee or by responsibilities andauthority stipulated by the Quality Management Manual
Chapter 3 Duty of Organization, (4) Responsibility andauthority.
The operator, according to a quality assurance program
established to comply with the national requirements,
must specify continual improvement of the effectiveness ofa quality management system and the assured
implementation of review to maintain appropriateness.
As for the activities of the regulatory body,i) The Quality Management Manual, which defines the
NISA management system cites the “Management Policy”
and sets forth NISA objectives and code of conduct andother matters of institutional policy.
ii) The Director-General determines the Management
Policy. The Policy states NISA’s fundamental policies suchas “Fundamental Philosophy of NISA”, “Basic viewconcerning Risks, Safety and Public Confidence”,“Reforming Institutional Management”, and “”How to
perform Activities -commonly followed by all individuals”.
iii) Moreover, the “annual work plan” that the individual
divisions and the Quality Management Committee eachestablishes contains the mid-term objectives and theannual work plan that the divisions and NISA as a wholeeach undertake. Each such annual work plan is inconformance with aforementioned institutional policies.
iv) Concerning an annual work plan, the Quality
Management Committee, with the Director-General aschairman, is established and examines and evaluates thecontents of the annual work plan. The conformance
between the institutional policy and the annual work plancan be confirmed through the Committee.
The operator, as described in 10.2 (2) of this report, mustclarify the organizational goal, policy, plan and objectives
for safety by the top management in a quality assurance
program.

DS338,

How are individuals given

As for the activities of the regulatory body,16

3.12,
and 3.13

responsibility and authority within
the management system, including
when external organizations are
involved in the system?

DS338,
4.1

How does the license holder
determine the resources necessary
to establish, implement, assessand continually improve themanagement system?

DS338,
5.1

How are management systemprocesses identified and theirdevelopment planned, assessedand continually improved?

Senior management fulfills its duties by participating in
the management of the management system within NISA
as members of the Quality Management Committee andthrough the responsibilities and authority.
Each division implements work allocated to it. The directorof each division is responsible for executing and managing
the allocated work.
The operator must specify organization and duties for the
personnel in the operational safety program.
The operator, pursuant to a quality assurance program
established to comply with the national requirements,
must clarify and assure the resources required for nuclear
safety.
As for the activities of the regulatory body,
The Quality Management Manual establishes the “Quality
Management Committee” within NISA, with the
Director-General as chairman, and inter alia mandates thecreation of “annual work plan” for individual divisions and
NISA as a whole, and establishes a mechanism toimplement, inspect and assess activities under the work
plans. In this manner, the process of the management
system has been established.
i) Directors and the Director-General each creates an
Annual work plan for his/her division and NISA, andexecutes, assesses, and makes activities improved under
the plan.
ii) The “annual work plan” shall be accompanied by chartsthat show an overall outline and procedures on therespective division (organization and process charts).
Processes and other measures shall be undertaken as
appropriate.
As for the licensee, refer to 13.3 of this report.

DS338,
5.11 - 5.28
DS338,
6.3

DS338,
6.7

What processes are covered by the
management system?

What independent assessments ofthe management system are there?

How is the system reviewed toensure continued suitability andeffectiveness, and enableaccomplishment of objectives?

As described in Article 13 of this report, the QMS isestablished. The QMS covers all the items including
document management, product control, recording
management, procurement control, communication andmanagement of organizational alteration.
As for the activities of the regulatory body,NISA activities are periodically evaluated by third parties(Nuclear and Safety Subcommittee). NISA is alsosupervised and audited by the NSC.
The licensee, pursuant to the quality assurance program
established to comply with the national requirements,
must assure the objectivity and fairness of an overseeing
process and let an auditor to audit duties other than its
own duty. And, the operator must receive the operationalsafety inspections by the Nuclear and Industrial Safety
Agency for the items specified in an operational safety
program.
As for the activities of the regulatory body,i) It is stipulated that the Quality Management Committee
can establish and amend the Quality Management Manual,
which determines the NISA Management System.
ii) The Quality Management Manual is to be reassessed asnecessary, taking into consideration appropriateness,
effectiveness and other relevant factors, and following
review by the Quality Management Committee.
The operator, pursuant to a quality assurance program
established to comply with the national requirements,
must review a quality management system in prescribedinterval to ensure that the quality management system of17

DS338,
6.11

DS338,
6.17

How are non-conformances andremedial actions dealt with?

How are opportunities for
improvement of the management
system identified and, whereappropriate enacted?

the organization continues to be appropriate, valid andeffective.
As for the activities of the regulatory body,i) When a division becomes aware of a non-conformance,
the division shall expeditiously report it to the Quality
Management Committee.
ii) When the Quality Management Committee receives thereport, it shall expeditiously investigate the cause andimplement corrective actions proportionate to theseriousness of the event.
The operator, pursuant to a quality assurance program
established to comply with the national requirements,
must implement nonconformity management andcorrective measures.
As for the activities of the regulatory body,i) Concerning each division’s annual work plan, it isstipulated that the division should conduct a
self-assessment at the end of the fiscal year, whose results
shall be subject to a hearing by the Quality Management
Committee.
ii) NISA’s work is constantly reviewed, and shall be
continuously improved, after review by the Quality
Management Committee.
The operator, pursuant to a quality assurance program
established to comply with the national requirements,
must investigate a possibility to improve the quality
management system and also assess a need to alter thequality management system including policy on quality
and quality goal, by management review process.

NS-R-1,
5.48,
and 5.49

How are designs made ‘operator
friendly’ and limit the effects ofhuman errors by plant layout,
work areas, working environment
and procedures (administrative,
operational and emergency),
including maintenance andinspection, in order to facilitate theinterface between the operating
personnel and the plant?

As described in Article 12 of this report, human factors
shall be taken account of when designing a main control
room and human factors and human error prevention shallbe taken account of in the operational management.

18

Article 14: Assessment and Verification of Safety
14(i)Question

Reply

NS-R-1,
3.10

How is a comprehensive safetyassessment carried out to confirm thatthe design as delivered for fabrication,
as for construction and as built meetsthe safety requirements set out at the
beginning of the design process?

As described in 14.1 and 14.2 of this report, a safetyassessment by the regulatory body is carried out at thereview process of a license for establishment, an approval of
construction plan and a pre-service inspection.

NS-R-1,
3.13

How does the operating organization
ensure that an independent
verification of the safety assessment isperformed by individuals or groups
separate from those carrying out thedesign, before the design is submitted
to the regulatory body?

It is not a regulatory requirement that the operatingorganization have an independent group verify the designfor establishment license.

NS-R-2,
4.1

How is the approval granted by theregulatory body before starting
normal operation?

As described in 7.2 (4) of this report, the regulatory body,
begore the start of normal operation, approves theoperational safety program of the operator, which specifies
concrete methods for operations management. And, as
described in 14.2 (1) of this report, the regulatory body
verifies that reactor facilities fulfill assigned functions andperformance through measures such as pre-service
inspection and the review of start-up test.

NS-R-1,
3.12

How is it ensured that licensees intheir safety assessments use data
derived from the safety analysis,
previous operational experience,
results of supporting research andproven engineering practice?

Please refer to 18.7 of this report, which describes themeasures to assure the technical reliability by tests,
experiences and analyses.

What safety reviews are undertaken if
there is a need to conduct a
non-routine
operation,
testorexperiment?

The licensee, pursuant to the operational safety program,must establish a committee which reviews items significant
to safe operation of nuclear reactor such as an alteration of
the operational safety program or operational procedures(please refer to 19.3 of this report).

What procedures are established by
the operating organization to ensure
proper design, review, control andimplementation of all permanent andtemporary modifications?

7.2 (4), “Regulation at Operational Stage” of this report
describes the measures to be taken for modification or
repair of electric facilities after commissioning.

How has the safety analysis of theplant design made use of the results ofdeterministic and probabilistic
methods?

While safety assessment using deterministic method is
described in 14.1 of this report, the effective use ofprobabilistic method is described in 14.4 and 14.6 of thisreport.

How have the computer programs,
analytical methods and plant models
used in the safety analysis beenverified and validated, and adequate
consideration been given to
uncertainties?

The computer programs, analytical methods, plant models
used in safety analysis and the associated uncertainties are
examined in the process of safety review of the design.

NS-R-2,
5.18
NS-R-2,
7.4

NS-R-1,
5.69,
5.73

NS-R-2,
10.1- 10.6
GS-R-1,
5.11

The operator, when preparing an application forestablishment license, confirms that the design should
comply with the requirements of the Regulatory Guide for
Reviewing Safety Design of Light Water Nuclear Power
Reactor Facilities, etc.

For the management of a temporary modification, refer tothe description of the preceding column on non-routine
operation.

What are the objectives, scope andfrequency of Periodic Safety Reviews
and how are the results are used?

14.3 (2) of this report describes periodic safety review andthe evaluation for aging.

To what extent and how does theregulatory body review and assess
modifications to safety related aspectsof a facility or activity taking into

7.2 (4), “Regulation in Operational Stage” of this report
describes the measures to be taken for modifications andrepairs after commissioning. The Nuclear and Industrial
Safety Agency establishes a procedure for authorization of19account the potential magnitude andnature of the associated hazard?

alteration of the operational safety program.

How does the regulatory body when
performing reviews and assessments
take into account the potential
magnitude and nature of the hazard
associated with the particular facilityor activity?

14.1 (2) of this report describes the evaluation of safety
design.

How have the operating organization
prepared and implemented a programof maintenance, testing, surveillance
and inspection of those structures,
systems and components which areimportant to safety?

As described in 7.2 (4) of this report, the operator mustspecify items concerning operations management of reactor
facilities in an operational safety program and receive
approval from the regulatory body before commissioning of
the reactor facility.

How is it ensured that operational
limits and conditions reflect theprovisions made in the final design?

As described in 7.2 (4) of this report, the operator mustdescribe the operational limits and conditions in theoperational safety program and receive approval from theregulatory body, which evaluates their adequacy.

NS-R-2,5.5

How has the operating organization
established and implemented an
appropriate surveillance program toensure compliance with theoperational limits and conditions, andhow are its results evaluated andretained?

As described in 7.2 (4) of this report, the operator mustspecify the operational limits and conditions in theoperational safety program and receive approval from theregulatory body. Through the operational safety inspection
by the regulatory body, observance of the operational safety
program including operational limits and conditions by theoperator is confirmed.

GS-R-1,
5.14

How does the regulatory body take
into account in its inspection program
the potential magnitude and nature of
the hazard associated with the facilityor activity?

When deciding the depth and extent of an inspection andthe frequency of an inspection, the potential magnitude andnature of the hazard associated with the facility or activity
are taken into account.

GS-R-1,
5.7

14(ii)
NS-R-2,
6.1 - 6.3

NS-R-2,
5.1

The regulatory body confirms observance of operationalsafety program by the operator through the operationalsafety inspection.

Concerning the extent of the inspections, electric structures
for business use that are particularly important to ensuring
public safety are designated as “specific electric structures
for business use” and subjected to inspection before use.
Boilers for power generation, turbines, and like items thatare particularly important to ensuring public safety aredesignated as “specific important electrical structures forbusiness use” and subjected to periodic inspections.
Concerning the frequency of inspections, the inspection
before use is conducted before a facility is placed in service,
the periodic inspections are conducted every 13 months, andthe safety preservation inspections are conducted four times
a year.
In addition, an on-site inspection is conducted as necessary.

20

Article 15: Radiation Protection
Question
NS-R-2,
8.1,

Reply

What program has the operatingorganizations have establishedand implemented to ensure that,in all operational states, doses
due to exposure to ionizing in theplant or due to any planned
releases of radioactive material
from the plant are kept below
prescribed limits and as low asreasonably achievable?

Refer to 15.3 (1) and (2) of this report.

How does the operatingorganization ensure that theradiation protection function in
its organization has sufficient
independence and resources toenforce and advice on radiation
protection regulations, standards
and procedures, and safe working
practices?

The operator must establish a dedicated organization in
charge of radiation control to confirm that the exposure ofradiation workers in reactor facilities complies with thelegislative requirements.

NS-R-2,
8.6

What are the requirements toensure that all site personnel
working in a controlled area or
regularly employed in asupervised area have theiroccupational exposures assessed
and what are the dose limits
required by the regulation?

Refer to 15.2 and 15.3 of this report.

NS-R-1,
6.90

What kind of systems is provided
to treat radioactive liquid andgaseous effluents in order to keep
the quantities and concentrationsof radioactive discharges
controlled and within prescribed
limits? How is the ALARA
principle applied?

Refer to 15.2 (2), 2) of this report.

NS-R-2,8.10

By what means has the operatingorganization demonstrated thatthe assessed radiological impacts
and doses to the general public
are kept as low as reasonably
achievable?

As described in 15.1 of this report, the operator, pursuant to
ALARA guideline, must establish the control target values.

NS-R-2,
8.11

How are the discharges ofradioactive effluents monitored
and controlled?

Refer to 15.2 (2) of this report.

NS-R-2,
8.12

What programs have been
established and implemented, if
required by the regulatory body,for monitoring the environment in
the vicinity of the plant in order toassess the radiological impacts ofradioactive releases on theenvironment?

Refer to 15.2 (3) of this report.

and 8.2

NS-R-2,
8.3

As described in 19.3 of this report, the operator must specify
items concerning radiation control in the operational safety
program and receive approval from the regulatory body.

As described in 19.3 of this report, the operator must specify
items concerning radiation control in the operational safety
program and receive approval from the regulatory body. The
observance of the operational safety program is confirmed
through the operational safety inspection.

As described in 15.2 (2) of this report, the operator must
control release to assure that the control target values are notexceeded.
The amount of release is disclosed to the public.
Released radioactivity of radioactive gaseous waste andradioactive liquid waste is monitored according to“Radioactive Materials Measurement Guidelines” and an
environmental monitoring is implemented in the vicinity of a
nuclear power station according to “Environmental Radiation
Monitoring Guidelines”.

21Article 16: Emergency Preparedness
16.1
Question

Reply

On-site emergency preparedness:
These requirements are covered by
the following quotations.

As described in 16.2 (1), 3) of this report, the nuclear operator
has an obligation to establish license holder’s plan for nuclear
emergency preparedness and to submit the plan to the
Minister of METI before commencing operation of a reactor.

SS 115,
V.4,
V.12,
V.13,
V.17,
V.19

How do the emergency plans include,
as appropriate intervention levels forrelevant protective actions and thescope of their application, withaccount taken of the possible severityof accidents or emergencies that
could occur?

As described in 16.2 of this report, in a case that a certain
specific event occurs in the nuclear power station, the operator
has an obligation to promptly notify the Minister of METI andthe head of local government(s) of it. Then, the national
government issues Declaration of Nuclear Emergency at a
specific level of the event and initiates prescribed actions. The
matters such as the reporting criteria of operator and theconditions, by which the national government issues
Declaration of Nuclear Emergency are described in Table 16-1of this report.

GS-R-2,3.8

How does the regulatory body:• The response to a nuclear emergency of reactor equipment
is described in the operational safety program.

NS-R-2,2.31- 2.38• Ensure that appropriate
emergency preparedness andresponse arrangement are
established when nuclear fuel isbrought to the site, and complete
emergency preparedness is
ensured before operation?
(NS-R-2, 2.36)• Ensure that such emergency
arrangements provide a
reasonable assurance of aneffective response?

• The operational safety program must be approved by theregulatory body prior to fuel loading.

• Require that the emergency
arrangements are tested in an
exercise before thecommencement of operation of a
new nuclear installation,• The licensee must establish an emergency preparedness
action plan and submit it before the commencement ofoperation of a new nuclear installation. However, it is not
required that the emergency arrangements should betested in an exercise before the commencement.

and thereafter:-At what intervals are exercises ofthe emergency arrangements
held?

-Pursuant to the emergency preparedness action plan, thelicensee normally implements the exercise once a year. For
details, see 16.3 of this report.

- Which ones does the regulatory
body witness? (NS-R-2, 2.37)-As described in 16.3 (1) and (3) of this report, the regulatory
body participates in the exercises implemented by the local
governments and the national government.

• Require that emergency plans are
periodically reviewed andupdated? (SS115; V.3)

• Article 7 of the Special Law on Nuclear Emergency
Preparedness stipulates that the plan for nuclear
emergency preparedness be reviewed annually andamended, as necessary.

GS-R-2,
3.12

How do all organizations that may be
involved in the response to an
emergency ensure that management
arrangements are adopted to meet
the timescales for response
throughout the emergency and for aneffective and coordinated response?

As described in 16.2 and 16.3 of this report, relatedorganizations prepare each own plan for emergency
preparedness, ensure effectiveness of the preparedness
through periodically implemented exercise and amend theplan, as necessary.

GS-R-2,4.7

How is transition from normal toemergency operations defined andmade?

As described in 16.2(3) of this report, the notifying criteria forevents recognized as emergency are provided in Table 16.1 ofthis report.

GS-R-2,
4.12

When circumstances necessitate an
emergency response, how do

The Special Law on Nuclear Emergency Preparedness defines
the notifying criteria such as radiation dose, upon which the22operators determine the appropriate
emergency class (see paragraph 4.19)
or the level of emergency response
and initiate the appropriate on-siteactions?
How does the operator notify andprovide updated information, as
appropriate, to the off-site
notification point?

operator must notify the regulatory body of the event, and thecriteria, upon which the government announces the
Declaration of Nuclear Emergency. The operator or thegovernment takes pre-determined actions based on their planfor nuclear emergency preparedness. Further subdivided
classifications for emergency situation are not specified. But,
the events that may lead to nuclear emergency are defined forindividual systems and components, and the operator
transmits those plant information to an off-site center on line,
as described in 16.2 (1), 1) of this report.

GS-R-2,
4.14

How is it ensured that appropriate
emergency response actions are
initiated promptly upon the receipt ofa notification from another State or
information from the IAEA of an
actual or potential transnational
emergency that could affect theState?

For a nuclear emergency occurred in a neighboring country,
our country will respond to it within the framework
established by the Convention on Early Notification of a
Nuclear Accident, the Convention on Assistance in the Case ofa Nuclear Accident or Radiological Emergency and the Basic
Law on Emergency Preparedness.

GS-R-2,
4.20

What criteria for emergency
classification are used to predefine
emergency action levels (EALs) for
abnormal situations (see paragraph4.70)?

The Special Law on Nuclear Emergency Preparedness defines
the notifying criteria such as radiation dose, upon which theoperator must notify the regulatory body of the event, and thecriteria, upon which the government announces the
Declaration of Nuclear Emergency. The operator or thegovernment takes pre-determined actions based on their planfor nuclear emergency preparedness. Further subdivided
classifications for emergency situation are not specified. But,
the events that may lead to nuclear emergency are defined forindividual systems and components, and implementation of
emergency operation starts in a timely manner.

How does the classification system
aid the initiation of a response toallow effective management andimplementation of emergency
operations?
GS-R-2,
4.27

What arrangements have been madefor response organizations to havesufficient personnel available toperform their assigned initial
response actions?

Response organizations are obliged to take measures preventthe propagation of nuclear accidents.

GS-R-2.
4.28

What arrangements have been madeto provide a response to a nuclear orradiological emergency for which
detailed plans could not be
formulated in advance?

An exercise where its scenario has not been provided forparticipants in advance is conducted, as necessary, so that incase of unexpected events they can handle them appropriately.

GS-R-2,
4.39

What arrangements are in place foractions by the operator to prevent an
escalation of the threat, to return the
nuclear installation to a safe andstable state, to reduce the potential
for releases of radioactive material or
exposures and to mitigate theconsequences of any actual releases
or exposures?

As described in 12.2 of this report, the operator must develop
the operational procedures for accidents and required actions
are specified in the procedures. And, training of operators is
implemented according to the operational procedures.

GS-R-2,
4.48

What arrangements are in place formaking and implementing decisions
on urgent protective actions to betaken off the site?

Please refer to 16.2 of this report.

GS-R-2,
4.56

What arrangements have been madeto protect emergency workers?

As described in Table 15-1 (Article 15) of this report, effective
dose limits for emergency workers are established.

GS-R-2,
4.67

How is radiation monitoring andenvironmental sampling andassessment carried out in order to
identify new hazards promptly and torefine the strategy for response?

Please refer to 16.2 of this report.

GS-R-2,
4.68

How is information about emergency
conditions, emergency assessments
and the protective actions
recommended and taken madePlease refer to 16.2 (1), 1) of this report.

Information such as measurements of environmental radiation
monitors installed in the periphery of a nuclear facility aretransmitted online to and indicated in an off-site center near
the reactor facility.

23

available to all relevant responseorganizations throughout the periodof the emergency?
GS-R-2,
4.71

GS-R-2,
4.80

What arrangements are there forpromptly assessing any radioactive
contamination, releases ofradioactive material and doses forthe purpose of deciding on or
adapting the urgent protective
actions to be taken following a
release of radioactive material?
What arrangements are there at thenational level to treat people whohave been exposed or contaminated?

Please refer to 16.2 (1), 1) of this report.

As described in 16.1 of this report, preparation of a medical
treatment system for nuclear emergency is stipulated, andrelated medical institution or emergency hospital is assigned
for each regional block.

GS-R-2,
4.86

What arrangements are there to
manage radioactive waste andcontamination resulting from an
accident?

The management of radioactive waste and contamination
resulting from an accident is defined as a series of actions after
a nuclear disaster according to the Special Law on Nuclear
Emergency Preparedness. The head of Designated national
administrative Agency or the head of Designatedlocally-stationed administrative Agency must implement
series of actions pursuant to their plan for nuclear emergency
preparedness or license holder’s plan for nuclear emergency
preparedness, based on the measurements of concentration or
density of radioactive materials or radiation in the areas
where emergency response are to be taken.

GS-R-2,
5.10

What arrangements are there for thecoordination of emergency response
and protocols for operational
interfaces between license holders
and local, regional and national
governments, as applicable?

As described in 16.2 (1), 2) and 3) of this report, local
governments and the operator have an obligation to prepare
their own plan for nuclear emergency preparedness and theinterface between their activities is explicitly specified in theirplans.

GS-R-2,
5.29

What national emergency facility or
facilities are designated for thecoordination of response actions andpublic information?

An off-site center is established to implement coordination
between related organizations (refer to Figure 16-2, Article 16of this report).

GS-R-2,
5.33

What exercise programs are
conducted on functions required for
emergency response andorganizational interfaces?

Please refer to 16.3 of this report.

How do these programs include theparticipation in some exercises of as
many as possible of the organizations
concerned?
What is done to systematicallyevaluate the exercises and for some
exercises to be evaluated by theregulatory body?
How is the program updating in thelight of experience gained?
16.2
GS-R-2,
4.82, 4.54

[GS-R-2,
3.5]

What steps have been taken by
the appropriate responsible
authorities to provide the publicwith information throughout anuclear or radiologicalemergency?

Such steps are specified in the Basic Plan for Emergency
Preparedness, Chapter1, Section 2, Item 7, “Appropriate
Transmission of Information to Residents in the Vicinity” andthe Emergency Preparedness Guidelines 2-4, “Provision ofInformation to Residents in the Vicinity”.

[What actions are taken by thenational coordinating authority tofoster the implementation of

Such actions are implemented in the framework for bilateralor multilateral cooperation in a timely manner.

24

emergency arrangements byother States?]
GS-R-2, 5.12

What arrangements have been
made to ensure that all States
within defined emergency zones
are provided with appropriate
information for developing theirown preparedness to respond toan emergency and whatarrangements have been made forappropriate transboundary
coordination?

Please refer to 16.4 of this report.

How is any risk (threat)
associated with nuclear
installations in other Statesconsidered?

For a nuclear emergency occurred in a neighboring country,
our country will respond to it within the framework
established by the Convention on Early Notification of a
Nuclear Accident, the Convention on Assistance in the Case ofa Nuclear Accident or Radiological Emergency and the Basic
Law on Emergency Preparedness.

16.3
GS-R-2, 3.15

In the risk assessment how are
populations at risk identified and,to the extent practicable, thelikelihood, nature and magnitudeof the various radiation related
risk considered?

25

Article 17: Siting
17(i)Question

Reply

NS-R-1,
5.18

In determining the design basis of a nuclear
installations how are the various interactions
between the installation and the environment
considered, e.g. factors like population,
meteorology, hydrology, geology, seismology andoff site services (e.g. electricity supply)?

As described in 17.2 and 17.3 of this report, suchinteractions between the installations andenvironment are taken into consideration.

NS-R-3,2.4,
2.14,
and 2.15

How are all site characteristics that may affectthe safety of the nuclear installation
investigated and assessed, including natural
phenomena and human induced situations andactivities in the region of the proposed site?
How are the proposed sites for nuclear
installations examined with regard to thefrequency and severity of external natural andhuman induced events and phenomena that
could affect the safety of the installation?

As described in 17.2 and 17.3 of this report, theconditions to be considered in a siting process
have been established.

NS-R-3,
2.5

How are the following external events, asapplicable, evaluated?
- Earthquakes, paragraphs 3.1-3.4- Surface faulting, paragraphs 3.5-3.7- Meteorological events, including extreme
values, paragraphs 3.8-3.10- Lightning, paragraph 3.11
- Tornadoes, paragraphs 3.12-3.14- Tropical cyclones, paragraphs 3.15-3.17- Floods due to precipitation and other causes,
paragraphs 3.18-3.23- Water waves induced by earthquakes or othergeological phenomena,
paragraphs 3.24-3.28- Floods and waves caused by failure of water
control structures, paragraphs
3.29-3.32- Slope instability, paragraph 3.33
- Collapse, subsidence or uplift of the site
surface, paragraph 3.35-3.37- Soil liquefaction, paragraphs 3.38-3.40- Behavior of foundation materials, paragraphs
3.41-3.43- Aircraft crashes, paragraphs 3.44-3.46- Chemical explosions, paragraphs, 3.48-3.49
NS-R-3,
2.21,
2.17,
and 3.52

As described in Article 18 of this report, detailed
design review guide for earthquake andearthquake induced events has been establishedby the Nuclear Safety Commission (3.1-3.7,
3.24-3.28). The guide was revised in 2006, as
described in 18.5 of this report.
External events other than earthquake are
required to be examined pursuant to the safety
design guides of the Nuclear Safety Commission
and are subjected to appropriate review andassessment process.

What kind of data are use to characterize thesite? How are the necessary data collected (sitespecific, data from other regions that are
sufficiently relevant to the region of interest,
prehistorically and historical data, simulation
techniques,
instrumentally
recorded
information) and how are they analyzed forreliability, accuracy and completeness?

For an earthquake, as shown in the Article 18,
they are specified in the Regulatory Guide for
Reviewing Seismic Design of Nuclear Power
Reactor Facilities which was revised in 2006.

NS-R-3,3.51

How is the region (including all facilities within
the site boundary) investigated for installations
in which materials are stored, processed,transported and otherwise dealt with that, if
released under normal or accident conditions,
could jeopardize the safety of the facility?

Potential effects by the events such as chemical
explosion in a neighboring facility are evaluatedat the stage of license review and are confirmed
to be negligible or insignificant for safety.

NS-R-3,How has the ambient radioactivity in the region

The pre-operational survey of the ambient

26

The reliability, accuracy and completeness of thesafety design guides and the Regulatory Guide
for Reviewing Seismic Design represent thecurrent highest level through the discussion anddeliberation by top level of experts in Japan.

4.15

assessed before commissioning of the nuclear
installation so as to be able to be determine theeffects of the installation and hence provide abaseline in future investigations?

radioactivity for comparison with the data aftercommissioning of a nuclear power reactor facility
is not implemented.

17(ii)
NS-R-3,
2.12,
and 2.22

How is for each proposed site the potential
radiological impacts in operational statesand in accident conditions on people in theregion, including impacts that could lead toemergency measures, evaluated with due
consideration of the relevant factors,
including population distribution, dietary
habits, use of land and water, and theradiological impacts of any other releases ofradioactive material in the region?

Please refer to 17.2 and 14.1 of this report.

NS-R-3,
2.26

In order to prepare for emergency planning,
how is the proposed site and region studied
to evaluate the present and foreseeable
future characteristics and the distribution of
the population of the region?

Please refer to 17.2 and 14.1 of this report.

How are the present and future uses of land
and water in the region evaluated that may
affect the potential consequences ofradioactive releases for individuals and the
population as a whole?
NS-R-3,2.27

How is it ensured that:

Please refer to 17.2 and 14.1 of this report.

(a) For operational states of the facility theradiological exposure of the population
remains as low as reasonably achievable,
economic and social factors being taken into
account?
(b) The radiological risk to the population
associated with accident conditions is
acceptably low?

17(iii)
NS-R-3, 5.1,
and 2.4

How and to what extent are thecharacteristics of the natural and humaninduced hazards as well as the demographic,
meteorological and hydrological conditions of
relevance to the nuclear installation
observed and monitored throughout the
lifetime of the nuclear installation?

Although the natural and human induced
hazards as well as the demographic,
meteorological and hydrological conditions arereviewed
whenconstructinganuclearinstallation, they are not monitored and theirimpacts are not evaluated throughout the
lifetime of the plant after its construction.

NS-R-2,
10.3

How
are
site
characteristics
andcorresponding external events taken into
account in a Periodic Safety Review todetermine to what extent the existing safety
analysis report remains valid?

In the current periodic safety review process, we
do not re-evaluated whether the conditions at thetime of siting assessment remain valid or not.

What arrangements are in place or planned
to ensure that all States within defined
emergency zones are provided withappropriate information for developing theirown preparedness to respond to an
emergency?

As described in 16.4 of this report, arrangements
for information exchange with neighboring
States have been established.

17(iv)
NS-R-2,
5.12

What arrangements are in place forappropriate transboundary co-ordination(participation in the licensing procedure andin environmental impact assessment)?

27

NS-R-1,
4.11

The safety of facilities and activities is ofinternational concern. Several international
conventions relating to various aspects ofsafety are in force.
What kind of arrangements have been
established by your national authorities,
with the assistance of the regulatory body, as
appropriate, for the exchange of safety
related information, bilaterally or regionally,
with neighboring States and other interested
States, and with relevant intergovernmentalorganizations,bothtofulfillsafetyobligations and to promote co-operation.

28

Please refer to the Introduction of this report.

Article 18: Design and Construction
18(i)Question

Reply

NS-R-1,
2.10

How are the five levels of the defense in depth
have been taken into account in the design andoperations of the nuclear installations of aplant to:(1) Prevent deviations from normal operation,and to prevent system failures?
(2) Detect and intercept deviations fromnormal operational states in order to prevent
occurrences from escalating to accident
conditions?
(3) Control design basis accidents to reach safe
shut down of the plant?
(4) Address severe accidents where designbasis may be acceded and insure thatradioactive releases are kept as low aspracticable?
(5) Mitigate the radiological consequences ofpotential releases of radioactive materials that
may result from accident conditions?

For the five levels of the defense in depth, pleaserefer to the following descriptions;Levels One to Three: 18.3 of this report;
Level Four: 18.4 of this report; andLevel Five: Article 16 of this report.

NS-R-1,
5.1,
and 5.3

How are structures, systems and components,
including software for instrumentation andcontrol, important to safety identified andclassified on the basis of their function andsignificance to safety?
How is it ensured they are designed,
constructed and maintained so that their
quality and reliability is commensurate withthis classification?
How is it ensured that any failure in a systems
classified in a lower class will not propagate
into a system classified in a higher class?

The policy for the Regulatory Guide for Reviewing
Classification of Importance of Safety Functions forLight Water Nuclear Power Reactor Facilities is
described in 18.3 of this report.
A requirement that “any failure in a systems
classified in a lower class will not affect a systemclassified in a higher class” is clarified in the item
IV. 4 of the regulatory guide mentioned above.

NS-R-1,
5.8 - 5.20

What internal and external events andcombination of events are been considered in
the design of the nuclear installations?

The events to be considered as subjects for a safetyassessment are described in 14.1 (2), 2) of thisreport.
Namely, by analyzing the failures of equipment or
systems or any operational error of them, the events
causing most severe consequence among a group of
events with similar progression are selected.
Then, these postulated events are classified into
“abnormal transients during operation” and“accident” as shown in the Regulatory Guide for
Reviewing Safety Assessment of Light Water
Nuclear Power Reactor Facilities, depending on
their possibility of occurrence and their potential
severity of impact if they occur.
Finally, these classified events are evaluated fortheir significance to safety according to the criteria
established for each classification.
a. “Abnormal transient during operation” is an
event leading to abnormal situation, which may becaused by a single equipment failure or anerroneous action or a single operational error by
operating personnel, or which may be caused by an
external disturbance, anticipated to occur during
the lifetime of a commercial power reactor. As
examples of such events, 14 events and 12 events
have been selected for a pressurized water reactor(PWR) and a boiling water reactor (BWR),

29

respectively.
In a safety analysis for the events, the integrity of acore and a reactor coolant pressure boundary isverified according to the criteria specified in theRegulatory Guide for Reviewing Safety Assessmentof Light Water Nuclear Power Reactor Facilities, sothat the adequacy of the safety design of importantsafety related equipment such as a safety protection
system and a reactor shut-down system is verified.
b. “Accident” is an abnormal condition exceeding“abnormal transient during operation”. Although
its frequency of occurrence may be extremely low,
its occurrence is intentionally postulated from theviewpoint of evaluating the release of radioactive
materials from a commercial power reactor.
As examples of such accidents, 10 events and 9
events have been selected for a PWR and a BWR,
respectively.
NS-R-1,
5.33,
and 5.34

To what extent are single failures and common
cause failures prevented in the designs of thenuclear installations?

To prevent a single failure and a common cause
failure, the Regulatory Guide for Reviewing Safety
Design of Light Water Nuclear Power Reactor
Facilities, Guideline 9, “Design Consideration for
Reliability” stipulates that a system performing
especially important safety function must
incorporate features such as redundancy, diversity
and independence, taking its structure, operation
principle and the nature of its safety function into
account.

NS-R-1,
5.31

What severe accident vulnerability studies
have been performed and what measures havebeen implemented as the result of the studies.

As described in 18.6 of this report, the measures for
AM have been developed.

How is it ensured that, wherever possible,
structures, systems and components
important to safety:

As described in 18.7 of this report, the measures toassure technical reliability through the feedback ofoperational experiences, tests and analyses havebeen taken.

18(ii)
NS-R-1, 3.6• Are designed according to the latest or
currently
applicableapproved
standards?
• Are of a design proven in previous
equivalent applications?
• Are selected to be consistent with theplant reliability goals necessary forsafety?
• Where codes and standards are used as
design rules, How are they
• Identified and evaluated to determine
their applicability, adequacy andsufficiency?
• Supplemented or modified as necessary
to ensure that the final quality iscommensurate with the necessary
safety function?
NS-R-1, 3.7

Where an unproven design or feature is
introduced or there is a departure from anestablished engineering practice, how issafety demonstrated to be adequate?
How is the development:• Tested before
service?

being

brought

into• Monitored in service, to verify that the
expected behavior is achieved?

30

As described in 18.7 of this report.

NS-R-1, 3.9

How does the design take account ofrelevant operational experience that has
been gained in operating plants and theresults of relevant research programs?

As described in 18.7 of this report.

How is it ensured that the design allows forreliable, stable easily manageable?

Please refer to 18.7 of this report.

How does the design management ensurethat the requirements of the operatingorganization are met and that due accountis taken of the human capabilities andlimitations of personnel?

The operator, according to the quality assurance
program established to comply with the national
requirements, must clarify and enssure theresources including human resources required fornuclear safety. Furthermore, the operator must
implement management of the competence of
personnel engaged in duties that may influence
nuclear safety.
Design parameters required for operation are
handed over from the design organization to theoperating organization, and the operatingorganization reflects the information in“Operational Limits”, etc. in the operational safety
program. Also the operational limits in theoperational safety program may be changed
following a plant modification.

18(iii)
NS-R-1, 4.8,
5.5, and 5.40
NS-R-1, 3.3

How does the design organization supply
adequate safety design information toensure safe operation and maintenance ofthe plant and to allow subsequent plantmodifications
to
be
made,
and
recommended practices for incorporation
into the plant administrative andoperational procedures (i.e. operational
limits and conditions)?

The modified operational safety program
required to be approved by the regulatory body.
NS-R-1, 5.50

How is it ensured that human behavior andthehuman-machineinterfacesystematically is taken into account earlyin the design process?

NS-R-1, 5.51

How is it ensured the design provide
operators with comprehensive but easy
manageable information, compatible withthe decision and action times?

NS-R-1, 5.56

How is it ensured that the need for
operators to intervene on a short-time scale
is kept to a minimum?

is

As described in Article 12 of this report, when
designing a main control room human factors aretaken into consideration, and human factors andhuman error prevention are taken into
consideration in the operation management.
Guideline 41 of the Regulatory Guide for Reviewing
Safety Design of Light Water Nuclear Power
Reactor Facilities requires that a main control roomshould be designed so that;the main control room allows operators
1) to monitor the operational conditions of a reactor
and its associated main facilities as well as theprincipal parameters, and2) to carry out rapid manual actions if they are
required.
The Regulatory Guide for Reviewing Safety
Assessment of Light Water Nuclear Power Reactor
Facilities, Main Text, Explanation II, “Safety
Design Assessment”, 4. “On the Items to beconsidered in Analysis” (3) stipulates followings;
A system and equipment performing safety function
should be generally designed so that it can perform
its required function immediately after theoccurrence of an abnormal condition without any
operator’s action.
In a case that any operator’s action will be expected,
operators must be provided with sufficient time andappropriate information to accurately judge thesituation and to carry out actions with highreliability.
At least 10 minutes are allowed for operators tocarry out any required action after they gain
appropriate information to make accurate
judgment.

31

Article 19: Operation
19(i)NS.R2, 4.6,
and NS-R-1, 3.10

NS.R.2, 4.11

Question

Reply

How does the operating organization
ensure that the commissioning program
reflects the appropriate safety analysis(see Art 14) and includes all the tests
necessary to demonstrate that the plant
as installed:• Meets the design intent, and• Can be operated in accordance withthe operational limits and conditions?
How is it ensured that no tests are
performed which could put the plant into
conditions that have not been analyzed?
How is it ensured that ‘baseline’ data on
systems and components, which areimportant for the safety of the plant andfor subsequent safety reviews, arecollected and retained?

The commissioning test corresponds to apre-service inspection in our terminology. In the
pre-service inspection, testing operation are
carried out by the operator. The test procedures,
their items and contents, are confirmed to beconsistent with the establishment license and theapproved construction plan. Fuels can be loaded
only after the operational safety program, with
operational limits and conditions, has been
approved by the regulatory body.

How is it ensured that reactor criticality
and initial power rising is not authorized
until all tests deemed necessary by theoperating organization and theregulatory body have been performed
and results acceptable to both parties
have been obtained?

Pre-service inspection is divided into two stages,that is, a stage for performance tests of individual
equipment and systems and a stage of power
testing. The criticality testing and the power
testing can be carried out only after all thetesting data required for them have beenacquired at the former stage.

How are operational limits andconditions developed to ensure that theplant is operated in accordance with thefinal design assumptions and intent?

The operational limits and conditions are
specified in an operational safety program, whichis required to be approved by the regulatory body.
Their approval can be granted if the operationalsafety program are consistent with theestablishment license and the approved
construction plan. Therefore, the plant can be
operated in accordance with the final design
assumptions and intent.

The data obtained in commissioning process arepreserved to be available in subsequent safety
review as baseline data.

19(ii)
NS.R.2, 5.1

How do they cover actions to be taken
and limitations to be observed by theoperating personnel?
NS.R.2, 5.2

How are operating personnel directly
responsible for the conduct of operation
made familiar with the intent andcontent of the operational limits andconditions?

The operational limits and conditions are
specified in the operational safety program, andthe operating personnel can be made familiar
with the operational safety program through aprocess such as education and training course inan operator training center.

NS.R.2, 5.5

How does the operating organization
ensure that an appropriate surveillance
program is implemented to ensurecompliance with the operational limits
and conditions, and that its results areevaluated and retained?

An operational safety program includes thesurveillance program, which specifies thesurveillance frequency to confirm the observanceof the operational limits and conditions, and thetime given to restore normal operation when adeviation was identified. The operatingorganization implements the surveillance processaccording to this program.

How are operating procedures developedand implemented for normal, abnormal

As describe in 12.2 (2), 2) of this report, operatingprocedures are developed for the conditions of19(iii)NS.R.2, 5.1132

and emergency conditions, in accordancewith the policy of the operatingorganization and the requirements of theregulatory body?

normal operation, accidents or failures andemergency.

NS-R-2, 5.12

Describe which kind of procedures are inplace for normal operation, abnormal
conditions, design basis accidents andsevere accidents?

As describe in 12.2 (2), 2) of this report, dedicatedprocedures and an accident management
guideline are developed against severe accidents.

NS-R-2, 5.10

Describe the administrative procedure
for the development, elaboration,
validation, acceptance, modification andwithdrawal of operating instructions andprocedures.

Operating instructions and procedures aredeveloped and controlled under the responsibilityof the operator without any regulatory
involvement. The regulatory body confirms some
aspects of the operating procedures which aredescribed in the operational safety program.

NS.R.2, 5.14

How is it ensured that operating
personnel are knowledgeable of, andhave control over, the status of plantsystems
andequipment
foralloperational states?

Operating personnel are educated and trained
according to the provisions described in theoperational safety program. Through theeducation and training, operating personnel are
made familiar with systems and equipment of aplant.

How is it ensured that only designatedand suitably qualified members of theoperating personnel control or supervise
any changes in the operational states ofthe plant?

The operator establishes a qualification systemfor operating personnel, and personnel without
any qualification can not engage in the operationof plant facilities.

How is it ensured that a non-routine
operation, test or experiment, is the
subject of a safety review and specificoperational limits and conditions and a
special procedure?

The operational limits and conditions are
specified in an operational safety program. Anyalteration of operational limits and conditions,with corrective actions from deviation from thelimits, requires regulatory approval.

If, during the non-routine operation, anyof the specific operational limits orconditions is violated, how is it known
what corrective action is to be taken?

Operational Safety Program states that special
safety measures should be taken for non-routine
operation.

What are the operating organization’s
program and types of procedures formaintenance, testing, surveillance andinspection of those structures, systems
and components that are important tosafety?

Work instructions, management procedures,
maintenance manuals and maintenance guides
are developed for individual equipment andsystems and they are periodically re-evaluated by
the operator.

NS.R.2, 5.18

NS.R.2, 6.1,
and 6.6

How often is it re-evaluated in the lightof experience?

19(iv)
NS.R.2, 5.11

How are operating procedures developedand implemented for abnormal andemergency conditions, in accordance withthe policy of the operating organizationand the requirements of the regulatorybody?

As describe in 12.2 (2), 2) of this report, operatingprocedures are developed for the conditions ofnormal operation, accidents or failures andemergency.

NS-R-2, 5.12

Describe which kind of procedures are inplace for abnormal conditions, designbasis accidents and severe accidents?

As describe in 12.2 (2), 2) of this report, dedicatedprocedures and an accident management
guideline are developed against severe accidents.

33

NS.R.2, 5.8

How is it ensured that after an abnormal
event, the plant is brought into a safe
operational state and the appropriate
remedial actions taken immediately?

Please refer to 19.4 and 19.6 of this report(Response in Accidents and Report of Events).

How is it ensured the operatingorganization undertakes a review andevaluation of the case and notifies theregulatory body?

19(v)NS.R.2, 2.10

How is it ensured that all activities that may
affect safety are performed by suitably
qualified and experienced persons, including
activities performed by contractors?

NS-R-2, 2.4

What provisions did the license holder taketo establish liaison with organizations fordesign, construction, manufacturing andplant operation and with other organizations(national and international) as necessary toensure the proper transfer of information,
expertise and experience and the ability torespond to safety issues?

(5),(6)

What adequate resources,
facilities are provided?

servicesandThe operator, according to the quality assurance
program established to comply with the national
requirements, must assess and select a supplierin a procurement process. The operator must
implement actions such as inspections to ensurethat requirements are satisfied.
The operator, according to the quality assurance
program established to comply with the national
requirements, must ensure that appropriate
in-house communication and informationexchange on the effectiveness of a quality
management system be maintained. Also, theoperator must specify requirements for
procurement and ensure requirements to a
supplier through the procurement process.

19(vi)
NS-R-2, 2.17

What procedures are in place for reporting
abnormal events to the regulatory body inaccordance with established criteria?

As described in 19.6 and Table 19-2 of this report,
reports on accidents and failures are stipulated
by legislation and regulations in detail.

How are abnormal events with safety
implications investigated?

Please refer to 19.7 of this report, which describes
processes for recurrence prevention and the feedback of lessons learned from the experiences ofaccidents and failures.

19(vii)NS.R.2, 2.21

How is the outcome of such investigations
converted into recommendations to the plant
management and corrective action?
How is information from such evaluations
and investigations fed back to the plant
personnel?
NS.R.2, 2.22

How does the operating organization use
operating experience at other plants toderive lessons for its own operations?

Please refer to 19.7 of this report, which describes
processes for recurrence prevention and the feedback of lessons learned from the experiences ofaccidents and failures.

NS.R.2, 2.23

How is operating experience examined forany precursors of conditions adverse tosafety, so that any necessary corrective
action can be taken before serious conditions
arise?

The accident sequences leading to the postulated
major accident are analyzed and theirprobabilities are calculated by PSA. A premise of
consideration is that a corrective action should betaken at the early stage of a small event leading
to a significant accident. In a case an event
actually occurs, it will be examined whether
postulated accident sequence was correct or not,and the sequence will be modified, as necessary.

NS.R.2, 2.25,

What mechanisms are used to shareimportant experience with other national
and international organizations?

Please refer to 19.7 of this report.

and 2.4(5)34

NS-R-2, 2.26

How is data on operating experiencecollected and retained for use as input for the
management of plant ageing, for theevaluation of residual plant life, and forprobabilistic safety assessment and periodic
safety review?

The operator, according to the quality assurance
program established to comply with the national
requirements, must implement ageing control
measures and periodic safety review.

NS-R-2, 8.8

How is the generation of radioactive waste
kept to the minimum practicable byoperating practices?

For radioactive waste management, please referto the Japanese National Report for theConvention on the Safety of Spent FuelManagement and the Safety of Radioactive Waste
Management.

NS-R-2, 8.9

What programs are in place to manage
radioactive waste at the site safely, also
taking into consideration conditioning andfinal disposal?

For radioactive waste management, please referto the Japanese National Report for theConvention on the Safety of Spent FuelManagement and the Safety of Radioactive Waste
Management.

NS-R-2, 5.21

How is spent fuel managed at the nuclear
installation?

For radioactive waste management, please referto the Japanese National Report for theConvention on the Safety of Spent FuelManagement and the Safety of Radioactive Waste
Management.

19(viii)- 5.2335

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