Nuclear Safety2021

Long-Term Management

and Actions for a Severe
Accident in a Nuclear
Power Plant
Status Report

NE A

Nuclear Safety

Long-Term Management and Actions for a Severe
Accident in a Nuclear Power Plant
Status Report© OECD 2021
NEA No. 7506
NUCLEAR ENERGY AGENCY
ORGANISATION FOR ECONOMIC CO-OPERATION AND DEVELOPMENTORGANISATION FOR ECONOMIC CO-OPERATION AND DEVELOPMENTThe OECD is a unique forum where the governments of 37 democracies work together to address the economic,
social and environmental challenges of globalisation. The OECD is also at the forefront of efforts to understandand to help governments respond to new developments and concerns, such as corporate governance, theinformation economy and the challenges of an ageing population. The Organisation provides a setting where
governments can compare policy experiences, seek answers to common problems, identify good practice andwork to co-ordinate domestic and international policies.
The OECD member countries are: Australia, Austria, Belgium, Canada, Chile, Colombia, the Czech
Republic, Denmark, Estonia, Finland, France, Germany, Greece, Hungary, Iceland, Ireland, Israel, Italy, Japan,
Latvia, Lithuania, Luxembourg, Mexico, the Netherlands, New Zealand, Norway, Poland, Portugal, Korea, theSlovak Republic, Slovenia, Spain, Sweden, Switzerland, Turkey, the United Kingdom and the United States. The
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Nuclear Energy Agency.

NUCLEAR ENERGY AGENCYThe OECD Nuclear Energy Agency (NEA) was established on 1 February 1958. Current NEA membershipconsists of 34 countries: Argentina, Australia, Austria, Belgium, Bulgaria, Canada, the Czech Republic,
Denmark, Finland, France, Germany, Greece, Hungary, Iceland, Ireland, Italy, Japan, Luxembourg, Mexico, theNetherlands, Norway, Poland, Portugal, Korea, Romania, Russia, the Slovak Republic, Slovenia, Spain, Sweden,
Switzerland, Turkey, the United Kingdom and the United States. The European Commission and theInternational Atomic Energy Agency also take part in the work of the Agency.
The mission of the NEA is:–

to assist its member countries in maintaining and further developing, through international
co-operation, the scientific, technological and legal bases required for a safe, environmentally soundand economical use of nuclear energy for peaceful purposes;–

to provide authoritative assessments and to forge common understandings on key issues as input togovernment decisions on nuclear energy policy and to broader OECD analyses in areas such as energy
and the sustainable development of low-carbon economies.

Specific areas of competence of the NEA include the safety and regulation of nuclear activities,
radioactive waste management and decommissioning, radiological protection, nuclear science, economic andtechnical analyses of the nuclear fuel cycle, nuclear law and liability, and public information. The NEA Data
Bank provides nuclear data and computer program services for participating countries.

This document, as well as any [statistical] data and map included herein, are without prejudice to the status of or sovereignty
over any territory, to the delimitation of international frontiers and boundaries and to the name of any territory, city or area.
Corrigenda to OECD publications may be found online at: www.oecd.org/about/publishing/corrigenda.htm.
© OECD 2021
You can copy, download or print OECD content for your own use, and you can include excerpts from OECD publications, databases
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Cover photos: Three Mile Island NPP, Chernobyl NPP (hujgcemj70, Creative Commons) and Fukushima Daiichi NPP (IAEA).

TABLE OF CONTENTS

Table of contents

List of abbreviations and acronyms .................................................................................................... 9
Executive summary .............................................................................................................................. 11
Chapter 1. Introduction ........................................................................................................................ 13
1.1. Background ................................................................................................................................ 13
1.2. Objectives and scope ................................................................................................................ 13
1.3. Structure of the report ............................................................................................................. 14
Chapter 2. State of the art of long-term management and actions .............................................. 15
2.1. Feedback from Three Mile Island unit 2, Chernobyl and Fukushima Daiichi
accidents .................................................................................................................................... 15
2.2. Status of long-term post-accident management and actions in NEA member
countries .................................................................................................................................... 92
Chapter 3. Approach to long-term post-accident management and actions ........................... 101
3.1. Definition and scope .............................................................................................................. 101
3.2. Long-term controlled state functions and monitoring ...................................................... 103
3.3. Possible accident/plant damaged states classification for long-term management ........ 105
3.4. Methods and tools for risks and issues ranking for long-term management ................ 114
3.5. Action identification and ranking table exercise ............................................................... 134
Chapter 4. Summary of challenges and main issues for long-term management ................. 163
Regulatory and organisational aspects ........................................................................................ 163
Relation to the public during long-term management .............................................................. 164
Risk evaluation for long-term management................................................................................ 164
Radioprotection in long-term management ................................................................................ 165
Coolable configuration and confinement of radioactivity ......................................................... 165
Development and use of unique systems, equipment and instrumentation ......................... 166
Contaminated water treatment .................................................................................................... 167
Site clean-up, de-contamination and waste management ....................................................... 168
Fuel retrieval .................................................................................................................................... 168
Chapter 5. Recommendations for enhancing long-term post-accident management
and for future research ................................................................................................. 171
5.1. Recommendations related to knowledge consolidation for cross-cutting issues ......... 171
5.2. Recommendations related to provisions development for cross-cutting issues ........... 174
5.3. Recommendations related to knowledge consolidation for specific long-term
management ........................................................................................................................... 177
5.4. Recommendations related to provisions developments for specific long-term
management and actions ...................................................................................................... 181

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Appendix A. Questionnaire on long-term management and actions for a severe
accident in a nuclear power plant ........................................................................... 183
Appendix B. Simplified application of SA-LT categorisation from level 2 probabilistic
risk assessment release category figures of merit ............................................... 187
Appendix C. Fault tree analysis and event tree analysis ........................................................... 195
Appendix D. SA-LT management schematic guiding procedure ............................................... 197
References …........................................................................................................................................ 201

List of figures
2.1.

Contaminated water in the basement of the reactor building ........................................ 18

2.2.

Processing and decontamination of radioactive waste water by the EPICOR-II
system ..................................................................................................................................... 19

2.3.

Processing and decontamination of high-level radioactive waste water by the SDS ... 19

2.4.

Processed water storage tanks to hold processed water from EPICOR-II and SDS
clean-up systems ................................................................................................................... 20

2.5.

High pressure spray decontamination................................................................................ 22

2.6.

Strippable coating for floor decontamination .................................................................... 22

2.7.

Remote control robotic high pressure spray washing system ......................................... 23

2.8.

Rubble bed at the bottom of the reactor ............................................................................. 24

2.9.

Full-scale defuelling work platform .................................................................................... 24

2.10. Rotating defuelling platform operation .............................................................................. 24
2.11. Core boring machine ............................................................................................................. 26
2.12. Canisters for fuel material, knockouts and filter tubes .................................................... 27
2.13. Canisters placed in a submerged storage rack in spent fuel pool ................................... 28
2.14. Manual debris clean-up operation....................................................................................... 29
2.15. Temporary solid waste storage facility for radioactive wastes ....................................... 29
2.16. Model 125-B rail shipping cask used for transporting defuelling canisters ................... 30
2.17. Independent spent fuel storage installation at INEL......................................................... 31
2.18a. Ruined Unit 4 after graphite fire ended .............................................................................. 36
2.18b. Photo of the destroyed reactor in the first hours after the accident: April 26, 1986 ..... 37
2.18c. Photo made from helicopter on May 3, 1986 ...................................................................... 37
2.18d. Installation of supporting structures of the original confinement (shelter),view during shelter construction......................................................................................... 37
2.18e. External protective structures of shelter before new confinement installation ........... 37
2.18f. View of the new confinement that covered the old one on 29 November 2016 ............ 38
2.19. Kinetics of integrated daily radioactive release during the initial phase of theaccident ................................................................................................................................... 41
2.20. Left: location of Chernobyl lavas according to first measurements (in red),
top right: photo of LFCM cluster in room 305/2, bottom right: photo of LFCM
covered by concrete ............................................................................................................... 43
2.21. Section of shelter (sarcophagus) isolating the destroyed unit 4 ...................................... 44
2.22. Different types of corium (fuel-containing materials) detected in the Chernobyl
shelter ...................................................................................................................................... 46
2.23. Scheme of horizontal and inclined boreholes drilled for sampling andinstallation of instrumentations .......................................................................................... 48
2.24. Scheme of melt flows and locations of lavas in different compartments ..................... 48

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2.25. Typical views of samples retrieved through the boreholes ............................................. 49
2.26. Examples of (Left) Brown Ceramic, (Centre) Black Ceramic and (Right)
Chernobylite microstructures (scale given in µm) ............................................................ 49
2.27a. Locations of sensors of “Finish-R” and “Signal” systems for monitoring nuclear
safety of fuel-containing materials ..................................................................................... 50
2.27b. Neutron detector readings (count/s) in the room # 304/3 from the morning of
29.06.90 to the evening of 30.06.90....................................................................................... 51
2.28. Leaching rates of different radionuclides from polychromic ceramic, brown
ceramic and black ceramic ................................................................................................... 53
2.29. New formations on LFCM surfaces: (top left) in steam release corridor (top right)
in steam discharging valve (bottom)black ceramics ......................................................... 53
2.30. Cs and Sr specific activities in groundwater (Bq/m3) near the shelter ........................... 55
2.31. Dynamics of daily (a) and annual (b) releases of β-active aerosols with airleakages from 1998 to 2015 ................................................................................................... 56
2.32. Temperature trends during five years after accidents in Fukushima Daiichi
units 1, 2 and 3 ....................................................................................................................... 64
2.33. Dose rate trends during five years after accidents in Fukushima Daiichi
units 1, 2 and 3 ....................................................................................................................... 65
2.34. Possible leak paths in Fukushima Daiichi units 1, 2 and 3 ............................................... 65
2.35. S/C water level measurement in Fukushima Daiichi unit 2............................................. 66
2.36. Current status in the three damaged units of Fukushima Daiichi units 1, 2 and 3 ...... 67
2.37. Muon tomography visualisation in Fukushima Daiichi unit 2 ........................................ 68
2.38a. “PMORPH” robot dose rate measurements in unit 1 PCV ................................................. 68
2.38b. “PMORPH” digital images in unit 1 PCV .............................................................................. 69
2.38c. “PMORPH” radiation dose levels in unit 1 PCV................................................................... 69
2.39a. Robot inspection of the unit 2 control rod drive rail area ................................................ 70
2.39b. Robot inspection of the unit 2 entrance of pedestal area ................................................. 71
2.39c. Robot inspection of the unit 2 pedestal area ...................................................................... 71
2.40a. Unit 3 internal investigation using underwater remote operated vehicle ..................... 72
2.40b. Unit 3 PCV internal investigation results............................................................................ 72
2.40c. Unit 3 PCV internal investigation results............................................................................ 73
2.41. Implementation actions for specified nuclear facilities and progress ........................... 75
2.42. Measures for mid-term risk reduction at TEPCO’s Fukushima Daiichi NPP
as of July 2017 ......................................................................................................................... 77
2.43. Ranking of hazard and likelihood risk components at Fukushima Daiichi ................... 78
2.44. PCV gas monitoring system at Fukushima Daiichi reactors ............................................ 80
2.45. Flow diagram of the contaminated water treatment........................................................ 82
2.46. Countermeasures against contamination of the marine environment ......................... 82
2.47. Optimised personal protective equipment......................................................................... 86
2.48. Strategy for debris characterisation .................................................................................... 88
2.49. Logic tree diagram for reduction of risks related to fuel debris retrieval ....................... 89
3.1.

Generic approach proposed for long-term management issues identification
and risk ranking ................................................................................................................... 115

3.2.

General structured layout for issues identification......................................................... 117

3.3.

LTM goals for the risks and issues identification as derived from the LTM
definition ............................................................................................................................... 118

3.4.

Fundamental actions affecting the long-term management goal of fueldebris heat removal ............................................................................................................. 120

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3.5.

Aspects considered for deciding on the likelihood of success of a given
long-term management action .......................................................................................... 136

B.1.

Level 2 probabilistic risk assessment flowchart .............................................................. 187

B.2.

Not-bypassed release category event tree ........................................................................ 191

D.1.

Long-term management flowchart ................................................................................... 197

List of tables
2.1.

Estimated masses of materials within the reactor location and masses
involved in Chernobyl lavas ................................................................................................. 45

2.2.

Masses of fuel and corium (LFCM) in different shelter compartments .......................... 45

2.3.

Averaged oxide content of Chernobyl lavas ...................................................................... 47

2.4.

Masses of dusts in different compartments of shelter ..................................................... 47

2.5.

Average radionuclides contents in fuel particles .............................................................. 48

2.6.

Average yearly data on contamination of water in the shelter ....................................... 55

2.7.

Timeline of events and operations at the Fukushima Daiichi NPP following the
11 March accident up to July 2017 ....................................................................................... 62

2.8.

Summary of investigations performed in units 1, 2 and 3 ............................................... 73

3.1.

Summary table of simplified classification of SA plant damaged states andcorresponding long-term management functions and challenges ............................... 112

3.2.

Issue Identification and risk ranking methods applicable to long-term
management and actions (LTMA)...................................................................................... 131

3.3a. Classification of challenges associated with implementation of long-term
management actions ........................................................................................................... 137
3.3b. Logical connectives used to fill in Table 3.4 ..................................................................... 137
3.4.

Combination of classification of long-term management challenges andderived connection with “likelihood of success” and “technology gaps” .................... 138

3.5.

Maintain coolable configuration AIRT .............................................................................. 158

3.6.

Waste water, solid waste and effluent management AIRT ............................................ 159

3.7.

Site clean-up and decontamination AIRT......................................................................... 159

3.8.

Defuelling of damaged reactors AIRT................................................................................ 160

3.9.

Damaged fuel/fuel debris and radioactive waste disposal, SFP fuel removal AIRT .... 160

3.10. Long-term management AIRT............................................................................................ 161

6

B.1.

Release category logic tree .................................................................................................. 189

B.2.

Minimal safety function configuration for non-bypassed SA-LT PDSs ........................ 192

LONG-TERM MANAGEMENT AND ACTIONS FOR A SEVERE ACCIDENT IN A NUCLEAR POWER PLANT, NEA No. 7506, © OECD 2021

LIST OF AUTHORS

List of authors

Lead authors
D. Jacquemain

Institut de Radioprotection et de Sûreté Nucléaire (IRSN), France

S. Basu

Nuclear Regulatory Commission (NRC), United States

M. Salay

NRC, United States

T. Lind

Paul Scherrer Institute (PSI), Switzerland

P. VokacÚJV Řež, Czech Republic

J.-C. De La Rosa Blul

European Commission, Joint Research Centre, the Netherlands

S. Bechta

Royal Institute of Technology (KTH), Sweden

A. Kubarev

KTH, Sweden

P. Ellison

GE Power & Water, United States

Y. Yamamoto

Tokyo Electric Power Company (TEPCO), Japan

K. Takeda

TEPCO, Japan

T. Niisoe

Nuclear Regulation Authority (NRA), Japan

V. Krasnov

Institute for Safety Problems of NPP (ISPNPP), Ukraine

Persons who contributed to the work group
A. Viktorov

Canadian Nuclear Safety Commission (CSNC), Canada

N. Mesmous

CSNC, Canada

J.-H. Song

Korean Atomic Energy Research Institute (KAERI), Korea

H.-Y. Kim

KAERI, Korea

K.I. Ahn

KAERI, Korea

M. KimInternational Atomic Energy Agency (IAEA), Austria

C. Pilleux

Electricité De France (EDF), France

A. Dubreuil-Chambardel

EDF, France

A. Hotta

NRA, Japan

R. Kojo

NRA, Japan

T. Matsuo

TEPCO, Japan

W. Giannotti

Nuclear and Industrial Engineering (NINE), Italy

K. Funaki

NEA

K. HidaNuclear Damage Compensation and Decommissioning Facilitation
Corporation (NDF), Japan

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LIST OF AUTHORS

Organisations that provided information on long-term management status(cf. questionnaire in Appendix A) and for the action identification and ranking table(AIRT) exercise (Section 3.5)
W. Barten

Federal Nuclear Safety Inspectorate, Switzerland

M. Richner

Kernkraftwerk Beznau (KKB), Switzerland

J.-U. Klügel

Kernkraftwerk Gösgen (KKG), Switzerland

A. Dubreuil-Chambardel

EDF, France

T. Nemec

Slovenian Nuclear Safety Administration (SNSA), Slovenia

E. Takasuo

Radiation and Nuclear Safety Authority (STUK), Finland

H. Tuomisto

Fortum, Finland

P. Isaksson

Swedish Radiation Safety Authority (SSM), Sweden

T. van Rompuy

BelV, Belgium

M. Sonnenkalb

Gesellschaft für Anlagen- und Reaktorsicherheit (GRS), Germany

F. Robledo

Consejo de Seguridad Nuclear, Spain

Persons who provided a complete external review of the report
Y. Maruyama

Japan Atomic Energy Agency (JAEA), JapanD. Marksberry

NRC, United States

N. Okuzumi

International Research Institute for Nuclear Decommissioning(IRID), Japan

Persons who provided a review on specific parts of the report
J. Byrne

General Public Utilities, United States, review for RWMC group

H. Chatri

Canadian Nuclear Safety Commission (CNSC), Canada, review for
WGRISK group

Secretariat
M. Kissane

NEA

N. Sandberg

NEA

This publication is dedicated to Martin Kissane who passed away in August 2020. He is sorely missed.

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LIST OF ABBREVIATIONS AND ACRONYMS

List of abbreviations and acronyms

AIRT

Action identification ranking table

ALARA

As low as reasonably achievable

ALPS

Advanced Liquid Processing System

BWR

Boiling water reactor

CRPPH

Committee on Radiation Protection and Public Health (NEA)

DOE

Department of Energy (United States)

EC

European Commission

EPRI

Electric Power Research Institute

FCM

Fuel-containing materials

IAEA

International Atomic Energy Agency

INEL

Idaho National Engineering Laboratory

IVC

In-vessel cooling

IVMR

In-vessel melt retention

ISLOCA

Interfacing system loss-of-coolant accident

LFCM

Lava-type fuel-containing materials

LT

Long-term

LTM

Long-term management

LTMA

Long-term management and actions

MAAP

Modular Accident Analysis Programme – Integral severe accident code developed
by Fauske & Associates Inc.

MCCI

Molten corium-concrete interaction

MELCOR

Integral Severe Accident Code developed by Sandia National Laboratory for the US
Nuclear Regulatory Commission

NDA

Nuclear Decommissioning Authorities (United Kingdom)

NDF

Nuclear Damage Compensation and Decommissioning Facilitation Corporation(Japan)NEA

Nuclear Energy Agency

NPP

Nuclear power plant

NRA

Nuclear Regulation Authority (Japan)

OECD

Organisation for Economic Co-operation and Development

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LIST OF ABBREVIATIONS AND ACRONYMS

10

PCV

Primary containment vessel

PDS

Plant damaged state

PEIS

Programmatic Environmental Impact Statement

PIRT

Phenomena identification ranking table

PMORPH

Primary Containment Vessel Internal Survey Equipment

PRA

Probabilistic risk assessment

PWR

Pressurised water reactor

RC

Release category

RCS

Reactor coolant system

R&D

Research and development

RPV

Reactor pressure vesselSA

Severe accident

SAMG

Severe accident management guideline

SARRY

Simplified active water retrieve and recovery system

SDS

Submerged demineraliser system

SED

Safety and environmental detriment

SFP

Spent fuel pool

TCOFF

Thermodynamic Characterisation of Fuel Debris and Fission Products Based on
Scenario Analysis of Severe Accident Progression at Fukushima Daiichi Nuclear
Power Station (NEA)TEPCOTokyo Electric Power Company, Japan

TMI

Three Mile Island

NRC

Nuclear Regulatory Commission (United States)

WENRA

Western European Nuclear Regulatory Association

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EXECUTIVE SUMMARY

Executive summary

This status report deals with the long-term management (LTM) of a nuclear power plant (NPP)
after a severe accident (SA), and has been produced with the objective to i) review the existing
regulations and guidance, practices, technical bases and issues considered in member countriesof the OECD Nuclear Energy Agency (NEA) regarding LTM of an NPP; ii) exhaustively identify,
describe and discuss the main challenges and issues to be tackled; and iii) propose
recommendations and areas for future investigation to improve LTM of an NPP.
In the report, LTM refers to accident management actions implemented after a plant hasreached a stabilised and controlled state following a reactor or spent fuel pool (SFP) severe
accident and up to and including fuel and debris retrieval from the damaged plant, temporary
on-site storage of the fuel and debris, and eventual transportation to off-site permanent storage.
The main LTM actions aim at: i) evaluating the plant damaged state (PDS) from a physical andradiological standpoint; ii) maintaining a stabilised and controlled state of the damaged plantfrom a safety perspective; iii) implementing provisions against further failures; iv) cleaning up
and decontaminating; v) managing accident wastes (conditioning, treatment, packaging andstorage); vi) preparing and achieving fuel and debris retrieval; and vii) protecting plant personnel
from exposure. Off-site long-term management and actions are not discussed, and neither areradiation protection, waste disposal and decommissioning aspects.
Based on a questionnaire that was circulated among NEA member countries it was
concluded that most of these countries do not have specific regulations for the long-term phaseof an SA, rather, it is commonly considered to be covered by the existing regulation and severe
accident management guidelines.
Information from the three major accidents at Three Mile Island (TMI), Chernobyl and
Fukushima Daiichi has been gathered and analysed in order to provide information and insights
specifically for LTM. Whereas TMI-2 LTM can be considered as completed when referring to theabove LTM definition, Chernobyl and Fukushima Daiichi LTM still face tremendous challenges
to complete LTM with important learnings yet to come. The three accidents have shown thatdifferent challenges and issues may arise for LTM depending on accident nature andconsequences with different “entry states” to LTM. For the three accidents, unique national
regulatory and licensing requirements were developed and they have also required thedevelopment of complex technical means (e.g. unique systems, equipment andinstrumentation), actions and organisations. The main challenge in implementing LTM in all
three accidents is that it had to be conducted with limited knowledge of LTM entry state, i.e. thestatus of the core and the plant at the entry to the long-term management phase, of risks to
evolve to a new unstable situation, and of risks related to long-term management and actions(LTMA) implementation. Looking more specifically at damaged fuel diagnostics and retrieval,
the three accidents have resulted in distinctive damaged fuel distributions and characteristics,
even among the three damaged reactors at Fukushima Daiichi. At TMI-2, the fuel retrieval
strategy had to be revised after investigations in the reactor pressure vessel. The fuel retrieval
was performed successfully with a specific defuelling platform where operations were exercised
beforehand to limit workers’ exposure. At Chernobyl, investigations have shown that ageing
through leaching and interactions with atmospheric gas and water may affect the integrity ofthe damaged fuel with possible fuel dusting in the long term. At Fukushima Daiichi,
uncertainties still remain regarding the fuel distribution in the three damaged reactors, corium
composition and its behaviour concerning leaching and ageing effects. The best strategies fordamaged fuel retrieval have not yet been established for Chernobyl and Fukushima Daiichi.

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EXECUTIVE SUMMARY

Next, approaches to long-term management have been discussed from a generalperspective. Firstly, the group developed the LTM definition and scope and identified main longterm controlled state functions and their necessary monitoring for a safe LTM. Then, an
approach has been developed to identify and categorise challenges, issues and risks for LTM
covering a large diversity of accident scenarios. In a first step, possible LTM “entry states”, which
depend on the accident progression up to that point, are systematically classified. LTM
challenges and issues have been identified for each generic entry state. As a second step, a
structured approach aimed at identifying and categorising main issues and risks for LTM wasdeveloped to guide LTM. Within this approach, a risk-informed, plant-specific classification
method of the events, different to the one using generic entry states, is used.
Finally, an action identification and ranking table (AIRT) was developed to identify
knowledge, challenges, open issues and technological gaps related to main LTM actions, suchas maintaining a coolable configuration and confinement integrity, managing water wastes,solid wastes and effluents, site clean-up and decontamination, intact and damaged fuel removalfrom reactors and SFPs and their disposal.
Based on the material and discussions presented in this report, recommendations are givenin the following areas:••

Knowledge development or consolidation for:–calculation tools and methods for analysis of reactor and SFP severe accidents toenhance capabilities to predict the stabilised state and the corresponding PDS;–

status of components, equipment, systems, including passive ones, and structures
after an SA with emphasis on those that contribute to maintaining a stabilised stateon the long-term (LT);–LT phenomena that can affect LTM (e.g. corrosion-erosion reactions, fuel “dusting”
and dispersion);–

methods or expert systems for risk assessment for LTM and LTMA optimisation.

Provision development for:–monitoring of the PDS and its evolution;–

upgrading equipment, components, systems and structures for LTM;–

developing harmonised practices and technical means to limit workers’ occupational
exposure in LTM.

Common between the three accidents reviewed are the serious challenges associated withhandling of contaminated and leaking cooling water. Therefore, it is also recommended todevelop provisions for the optimisation of management of cooling waters to facilitate LTM:

12•during the emergency phase, closed-loop cooling should be implemented as early as
possible;•

strategies for flooding and cooling the corium should as far as feasible avoid transfer ofcontaminated waters outside the confinement;•

use of water with controlled chemistry should be further studied with respect tolimitation of risk for re-criticality, fission products remobilisation, corrosion, clogging,and for facilitation of water management in the long term.

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INTRODUCTION

Chapter 1. Introduction

1.1. Background
Following the accident at the Fukushima Daiichi nuclear power plant (NPP), one of theimperatives for the nuclear safety, science and industry communities is to reassess the safetyof existing NPPs, notably to evaluate the sufficiency of technical means and administrative
measures addressing the management of accidents for the design basis and beyond andincluding long-term phases.
Up to now, international actions primarily addressed lessons learnt from the Fukushima
Daiichi accident for the management of short-term phases (emergency operating procedures
and severe accident management guidelines [SAMGs] domains) and for emergency
preparedness, with exceptions mentioned in Section 1.2, but the long-term management andactions (LTMA) have not been examined in detail.
In recognition of international interest in additional valuable information that could be
gained from the Fukushima Daiichi accident LTMA, NEA Committee on the Safety of Nuclear
Installations (CSNI) recommended at the end of 2014 to launch an action i) that addresses areview of the experience gained from management of long-term phases in Three Mile Island
unit 2 (TMI-2), Chernobyl and Fukushima Daiichi accidents and of technical and organisational
measures and practices in NEA member countries, ii) that would provide recommendations forenhancement for the management of long-term phases of a severe accident (SA) and identify
related safety research knowledge gaps. It was felt that both enhancements in terms oftechnical means and organisation could be proposed to optimise implementation of LTMA.
This report describes the outcomes of this action that was led by a working group of the
NEA Working Group on Accident Management and Analysis (WGAMA) with the final objectiveof providing guidance for the international community to enhance safety related to the longterm management of a severe accident and engage well focused research in the field.

1.2. Objectives and scope
The objectives of the Long-Term Management of Actions for a Severe Accident in a Nuclear
Power Plant (LTMNPP) Working Group were to:•review the experience gained for LTMA from the TMI-2, Chernobyl and Fukushima
Daiichi accidents;•

review envisaged, planned or existing regulations, guidance and practices in NEA
member countries for LTMA for a severe accident in an NPP;•

suggest possible approaches for LTMA following a severe accident including methods forrisk ranking and issues identification;•

identify the main challenges and open issues faced during the long-term phase of asevere accident;•

derive recommendations to better address them in the future, including, for instance,orientation onto where research and development (R&D) efforts should be allocated, or
improvement of methods to assess LTMA;•

derive recommendations to enhance LTMA.

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INTRODUCTION

The report is primarily focusing on on-site LTMA following an SA and is not covering issues
related to off-site long-term management and actions. Specific working groups and actions are
dealing with this issue. Also, radiation protection, waste disposal and decommissioning aspects
have not been treated in detail here as specific NEA working groups and actions were conducted
to treat them – the Committee on Radiological Protection and Public Health (CRPPH) (NEA, 2014)
and Radioactive Waste Management Committee (RWMC) (NEA, 2016) (Expert Group on
Fukushima Waste Management and Decommissioning R&D [EGFWMD] and Working Party on
Decommissioning and Dismantling [WPDD]).
In addressing the long-term management (LTM) action, elements of interest arising from
related NEA working groups such as the NEA Benchmark Study of the Accident at the
Fukushima Daiichi Nuclear Power Station Project and the NEA Senior Expert Group on Safety
Research Opportunities Post-Fukushima (SAREF) have been considered. These are reported anddiscussed in the chapter describing the experience gained from Fukushima Daiichi accident for
LTMA.

1.3. Structure of the report
Chapter 2 provides a description of the state of the art of long-term management and actions
based on severe accidents experience (TMI-2, Chernobyl and Fukushima Daiichi), on
information compiled through a questionnaire on envisaged, planned or existing regulations,guidance and practices in NEA member countries for LTMA for a severe accident in an NPP andon a review of relevant existing reports in the literature.
Chapter 3 suggests possible approaches to long-term management and actions considering
diverse scenarios of SA and spent fuel pool (SFP) accidents and including methods for issues
identification and risks ranking.
Chapter 4 provides a summary and a discussion of main identified challenges and open
issues faced during the long-term phase of a severe accident.
Chapter 5 provides some recommendations for enhancing long-term management andactions better addressing identified challenges and open issues, including the development ofprovisions, of approaches and the design of future research to consolidate relevant knowledge.

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Chapter 2. State of the art of long-term management and actions

2.1. Feedback from Three Mile Island unit 2, Chernobyl and Fukushima Daiichi
accidents
This chapter provides a description of key technical and organisational issues identified fromthe long-term management of the accident at Three Mile Island unit 2 (TMI-2), Chernobyl and
Fukushima Daiichi.

Synthesis of relevant Three Mile Island unit 2 experience
The TMI-2 nuclear power plant (NPP) accident that occurred on 28 March 1979, was the mostserious accident in the history of US nuclear power industry (ANS, 1989). The accident resulted
in partial meltdown of the reactor core, and though the reactor pressure vessel did not breach,
there was a small radioactive release with no detectable consequence on plant workers’ health,public health and environment. The nuclear industry had no prior experience in dealing withthis kind of event; neither did the US Nuclear Regulatory Commission (NRC) have priorexperience with regard to its regulatory oversight role. The guiding principle for actions to betaken by both parties in the aftermath of the accident was to ensure protection of public healthand safety, and that of the environment.
The recovery and clean-up of TMI-2 (EPRI, 1992a, 1992b, 1992c, 1990a, 1990b; IAEA, 2014, 1992,
1991a. 1991b; DOE, 1993, 1990, 1988; NRC, 2016), including removal of the fuel from theaccident-damaged reactor, were done for the long-term protection of public health and safety
and of the environment. Implementation of recovery and clean-up activities was the
responsibility of the licensee with support from US nuclear industry, US Department of Energy(DOE), and several international organisations. The NRC was responsible for the regulation andoversight of TMI-2 clean-up operations to ensure the health and safety of the public and theplant personnel, as well as the protection of the environment. NRC’s involvement covered two
major areas: approving the recovery methods employed by the licensee and responding to
public concerns over radiation exposure resulting from the accident and clean-up activities.
The DOE initiated an extensive research programme, as directed by the US Congress, tosupport recovery and clean-up operations (DOE, 1993, 1990, 1988). The damaged fuel from thereactor core was removed and shipped for research to national laboratories which provided
much-needed technical support for the TMI-2 recovery and clean-up programmes.
The NRC accomplished its regulatory responsibilities for all post-accident operations at
TMI-2 through licensing actions; safety evaluations of recovery and clean-up activities;
inspections; daily interactions with the licensee and their contractors; communications withstate and local governments and the public; co-ordination with other Federal agencies involvedin the clean-up; and sometimes direction from the NRC Commissioners. Formal licensing actions
were required for proposed changes to NRC orders, the facility licence, technical specifications,the recovery operations plan, the organisation plan, and exemptions to regulations.
The licensee’s (TMI-2 plant owner) organisational structure and functions changed as theclean-up progressed through various stages of the recovery effort, and as more information
became known about the condition of the damaged reactor core, reliability of plant systems andstructures, and radiological characterisation of accident-generated water, building structures,reactor components, and systems. Various organisations formed working groups to provide
guidance on addressing specific issues, problems, and research activities. Independent
oversight groups reviewed, monitored, and advised on the overall direction of recovery andclean-up plans and activities.
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Changes in the facility’s post-accident mode of operations required unique regulatory andlicensing actions. In order to properly reflect evolving plant status, the NRC-issued orders,modified those orders, approved licence amendment requests, and granted relief from certain
regulatory requirements. The recovery operations plan defined the surveillance requirements
to be performed to ensure equipment operability as required by the plant’s technicalspecifications. Surveillance requirements were approved by the NRC staff. The NRC granted
exemptions from certain requirements of the regulations for NPPs, but only under special
circumstances as permitted in the regulation. Exemptions were necessary at TMI-2 because of
the plant’s damaged configuration and changing status during clean-up.
The commission established a 12-member TMI-2 advisory panel in October 1980 to consult
with, and provide advice to the NRC Commissioners and staff on major activities related to thedecontamination and clean-up of TMI-2. The panel consisted of members from the state
government (Pennsylvania), local government, and the scientific community, as well as
residents in the vicinity of TMI. The NRC TMI Program Office (TMIPO) acted as a liaison between
the NRC and the TMI-2 advisory panel and also provided information to the panel on the statusof the clean-up. The most crucial panel influence on clean-up activities was the increased public
scrutiny of both NRC and licensee decisions and activities. The panel facilitated communication
with the public for both the NRC and the licensee. This communication helped sensitise theagency and the licensee to public concerns.
Advisory and working groups were formed by the General Public Utilities Corporation (GPU),
the NRC, and the DOE to provide advice, and sometimes direction, on important recovery
activities. Key groups that supported long-term recovery of TMI-2 included the following. The
NRC “Advisory Panel for the Decontamination of TMI-2,” was an influential 12-member panelthat provided advice to the NRC Commissioners and staff on major activities related to thedecontamination and clean-up of TMI-2. The TMI-2 “Safety Advisory Board” was established by
the licensee to provide the licensee with an independent appraisal of the recovery programmethat gave particular emphasis to the assurance of public and worker health and safety. Thisboard’s members were national known experts. The TMI-2 “Technical Advisory and AssistanceGroup” was established by the licensee, with the co-operation of the DOE and NRC, to provide
independent technical assessment and advice on the decontamination and defuelling of TMI-2.
The “TMI Information and Examination Program” was established by the DOE to acquire data to
improve current understanding of nuclear plant accident environments and of the phenomena
which contributed to those environments. Membership included the licensee, the Electric Power
Research Institute (EPRI), the NRC and the DOE.
The TMI-2 clean-up effort took about 14 years, and the reactor vessel defuelling operations
spanned a five-year period. A total of about 133 000 kg of fuel, cladding, structural, and control
materials were removed from the reactor vessel during the five-year effort. Unique systems andequipment were designed and installed to remove damaged fuel and structural debris from thereactor vessel. The clean-up activities challenged the management of various forms andconcentrations of radioactive waste. The management of highly contaminated water, fueldebris, and related solid-waste by-products included handling, processing, temporary on-sitestorage, transportation, and final disposal. Decontamination activities resulted in substantial
quantities of contaminated water and organic resins and inorganic zeolites produced fromwater processing systems. Fuel debris that spread throughout the plant created uniqueradiological waste characteristics. Also, some waste did not fit into established regulatory waste
classification categories for transportation and disposal.
The NRC’s Programmatic Environmental Impact Statement (PEIS) related to thedecontamination and disposal of radioactive wastes resulting from the accident was animportant set of guidance documents for the NRC and licensee. The PEIS discussed the options
and associated environmental impacts of four fundamental activities necessary to the clean-up:
treatment of radioactive liquids; decontamination of the building and equipment; removal offuel and decontamination of the coolant system; and packaging, handling, storing andtransporting nuclear waste. These items are discussed below in some details.
In terms of radiological health and safety, there was no known technical reason for theradiological release criteria to be more restrictive than had been acceptable at normal operating
facilities. However, because of the unique characteristics of the clean-up operation that were16

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not considered and evaluated in the pre-accident safety review of the plant, there was a needto define what keeping radiation exposure as low as reasonably achievable (ALARA) meant withrespect to off-site releases and occupational exposures. The PEIS provided the basis for making
that determination.
The following sections briefly describe, after presenting actions taken during the“stabilisation phase”, the key long-term accident management insights and issues identified
from the accident at the TMI-2 plant (ANS, 1989; EPRI, 1992a, 1992b, 1992c, 1990a, 1990b; IAEA,
2014, 1992, 1991a. 1991b; DOE, 1993, 1990, 1988; NRC, 2016). These insights may serve as a basis
for inputs into a long-term post-accident management plan. The items for consideration are:•waste water management;•

decontamination of buildings and equipment;•

defuelling;•

disposal and long-term storage of fuel and other radioactive wastes;•

worker protection.

Actions taken during the stabilisation phase
During the short-term accident management phase of TMI-2, the pressing need was to stabilise
the conditions throughout the plant – hence, this phase was more commonly referred to at thetime as the “stabilisation phase.” This phase, which lasted for just over 16 months, involved
certain essential plant operational and control functions to keep the reactor cool. Specifically,
these functions were: i) maintaining reactor coolant system (RCS) flow and water inventory;ii) controlling RCS pressure; iii) maintaining RCS heat removal; and iv) ensuring the core remain
subcritical.
The severely damaged core and resulting large quantities of non-condensable gas
generation posed a formidable challenge for the primary coolant flow from being blocked. The
operations staff developed procedures to control and remove non-condensable gases, inparticular, hydrogen. The degassing operation was successfully conducted by maintaining thesystem pressure relatively high (6.2-6.9 MPa) and by continually adding reactor coolant makeup water. The pressuriser temperature was maintained at a considerably higher level than the
coolant system, which helped the gas to effervesce and to be removed by venting. Hydrogen
was effectively removed in this manner in about six days.
The decay heat removal system in TMI-2 was designed for 2.4 MPa. Because of the need tocontrol the RCS pressure in the 6.2-6.9 MPa range during the degassing operation, the decay heatremoval system was not used in the first week of the stabilisation phase, which turned out to bebeneficial for keeping the dose rate low. The only viable installed alternative to the decay heatremoval system was to use the steam generators to achieve and maintain cold shutdown.
The RCS pressure and volume control systems were lost as a result of the accident.
Operators worked around the problems temporarily while the construction of a new pressure
control system (standby pressure control system [SPCS]) began. The SPCS consisted of pumps
and accumulator tanks for surge suppression using bottled nitrogen. The construction wascompleted at the end of 1979, and the system remained operational from early 1980 to 1984
when the reactor was depressurised for vessel head removal.
The exact severity of core damage was unknown; however, the control material wasessentially non-existent in the core region as a result of melting. Thus, control rods were notrelied upon in any way for reactivity control and assurance of shutdown. The operators had no
way of measuring the criticality margin. Thus, a high boron concentration was maintained in
the coolant to ensure that the core would not become critical. Later on, it was also determinedthat re-criticality would be virtually impossible because of the non-optimum configuration of
the agglomerated melt in relation to the surrounding water.
Besides the main task of reactor control during the short-term accident management phase(i.e. stabilisation phase), controlling radioactivity was also a high priority task as any off-site
release could potentially have a large environment and health consequence. During the
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accident, ruptured fuel released substantial quantities of gaseous fission products into thereactor coolant. These gases were carried to various low pressure tanks in the liquid clean-up
system, and were the primary source of the problem. These gaseous releases were causing
serious airborne contamination inside the auxiliary building. Several actions were initiated toaddress the problem which include: i) releasing the contents of the waste gas decay tanks into
the containment; ii) finding the source of leakage into the auxiliary building; iii) replacing thecharcoal filter; iv) designing a charcoal bed adsorption system; v) designing and constructing a
new (auxiliary and fuel handling building [AFHB]) ventilation system; and vi) purging thecontainment of krypton-85.
Many of the actions initiated during the short-term accident management at TMI-2 andequipment/procedures – existing or implemented after the accident – were continued beyond
this phase. Examples are maintaining reactor control and having access into the containment
and auxiliary building, waste water processing and storage, initial decontamination effort, etc.

Waste water management
During the first hour of the TMI-2 accident, spilled reactor coolant in the reactor building sump
was automatically pumped into the auxiliary building holding tanks, which then overflowed
along with the sumps. Leakage from the make-up and purification system did contain fueldebris and fission products which subsequently mixed with the water in tanks, sumps, and floor
drains. Over one million gallons (3.8 million litres) of contaminated water ended up in thebasement (Figure 2.1) of the reactor building (creating an over two-metre deep pool) and in tanks
in the auxiliary building.

Figure 2.1. Contaminated water in the basementof the reactor building

Source: NRC (2016).

Contaminated water prohibited access to the reactor building and large areas of theauxiliary and fuel handling buildings. Removal of the contaminated water from the buildings
was the first major decontamination task (NRC, 2016). The NRC’s approval was required todispose of the decontaminated water. On October 1979, an order for modification of licence wasissued to require the licensee to promptly operate the EPICOR-II filtration and ion-exchange
decontamination system (Figure 2.2) to decontaminate intermediate-level radioactive waste
water held in tanks in the TMI-2 auxiliary building. Processing of intermediate-level waste water
inside the auxiliary building began almost immediately and was completed in little over one
year time. During this period, over one million gallons of water was processed, with about halfof this amount being recycled processing. After processing, the water was collected in a cleanwater receiving tank for sample measurements of radionuclide concentrations.

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Figure 2.2. Processing and decontamination of radioactive
waste water by the EPICOR-II system

Source: NRC (2016).

The NRC approved the use of the submerged demineraliser system (SDS) which was
designed and installed to clean-up high-level radioactive accident-generated waste water fromthe reactor building’s basement, reactor coolant system, and reactor coolant bleed tanks(Figure 2.3). The SDS consisted of a liquid waste treatment subsystem, a gaseous waste
treatment subsystem, and a solid waste handling subsystem. The liquid waste treatment
subsystem was designed to remove caesium and strontium from the high-activity waste waterby filtration and ion exchange. The SDS started operation in June 1981.

Figure 2.3. Processing and decontamination of high-level
radioactive waste water by the SDS

Source: NRC (2016).

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From September 1981 to April 1987, EPICOR-II was used to remove residual radioactivity
from SDS effluents and to process miscellaneous wastes from 11 September 1981 to April 1987.
After the SDS was removed from operation in 1988, EPICOR-II was the primary system to cleanup waste water that was mainly generated from building decontamination activities. The
system configuration was the same as before; however, a high integrity container loaded withzeolite resins was placed in the first position to act as a roughing filter to remove gross caesium
and strontium radionuclides.
The processed water from the EPICOR-II and SDS clean-up systems was stored on-site in two
500 000 gallon (2 million litres) tanks (Figure 2.4) which became operational on July 1981.
A recycling system was later installed to transfer the processed water into the reactor building fordecontamination activities. This water was subsequently reprocessed by the SDS and returnedback to the processed water storage tanks for further use or to await ultimate disposal.

Figure 2.4. Processed water storage tanks to hold processed waterfrom EPICOR-II and SDS clean-up systems

Source: NRC (2016).

The processed water disposal system (a closed cycle evaporator) was used to dispose of
2.3 million gallons (8.7 million litres) of processed accident-generated water. The SDS removed
99% of the radioactivity and boiling the decontaminated water twice in the evaporator removed
most of the remaining traces of radioactive particles. However, radioactive tritium could not be
removed and was released into the environment during the evaporation process. The maximum
radiation exposure from the evaporation process was less than one millirem (10 µSv), which isequivalent of about one day of background radiation.

Decontamination of the buildings and equipment
Exposed surfaces in the reactor building and the auxiliary and fuel handling building were
contaminated with radionuclides. Airborne releases entered the ventilation systems and spreadthroughout the auxiliary and fuel handling building. Airborne releases contaminated surfaces
on the upper-level floor and mid-level floor of the auxiliary building, and liquid releases to thedrain system contaminated surfaces on the basement-level floor. The interior of the auxiliarybuilding, including 26 piping systems, was contaminated by radioactive material though lessseverely than the interior of the reactor building. After the accident, the water in the reactor
building’s basement was heated by residual heat from the reactor vessel, evaporated,condensed on the cooler walls, and drained down onto the floors and back into the basement.

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The overall objectives of the TMI-2 decontamination efforts were to maintain access to andoperability of plant systems, to support defuelling preparations and operations, and to permit
the transition of the facility to a long-term storage condition. Shorter-term decontamination
objectives focused on the removal or stabilisation of contamination in order to reduce
occupational exposure and to prevent release of contamination to the environment. The
decontamination objective was to stabilise localised radiological conditions in the plant,
regardless of whether or not access was required for clean-up activities. Longer-term
decontamination objectives ensured that any remaining contamination was stable andsufficiently isolated for long-term storage.
The decontamination objectives for the reactor building working areas were to reduceradiological conditions (general area radiation, airborne gaseous and particulate activities, andsurface contamination levels) to ALARA levels and to maintain those conditions in a way thatwould permit defuelling operations. The decontamination objectives for the auxiliary and fuelhandling building were to permit access without restriction because of surface or airborne
contamination, to reduce radiation exposure from gamma sources to ALARA levels, and toprevent recontamination from other clean-up activities or system leaks.
Clean-up of the auxiliary and fuel handling building started shortly after the accident.
Approximately 510 000 square feet (47 300 m2) of surface in the auxiliary and fuel handlingbuilding required decontamination when clean-up operations began. The reactor building wasan effective barrier against the release of radioactivity off-site and as a result the building’s
interior was extremely contaminated. The decontamination work was in preparation for theremoval of the damaged fuel core from the reactor vessel. Radioactive gases (Krypton 85)
accumulated in the reactor building and had to be removed (vented) before the building could
be safely entered. Clean-up of the reactor building did not begin immediately as the humidity
in the reactor building was 100% causing precipitation in the form of rain (steadily dripping
condensation) inside the reactor building.
Decontamination of building surfaces, systems, and equipment included multiple activities
across the following categories:•

Loose and installed equipment: removal of miscellaneous loose equipment and debris
that were in the facility at the time of the accident, such as ladders, scaffolding, tools,
and portable equipment. Also, decontamination or removal of installed equipment, suchas piping systems, air conditioning and exhaust equipment, cable trays, and electricaland lighting equipment.

•Interior surfaces: decontamination of interior building surfaces consisting of metal andconcrete materials.

•

Sludge and resins: removal of contaminated sediment (sludge) from tanks and sumps in
the auxiliary and fuel handling building and from the reactor building’s basement floor
and sump. Also, removal of highly contaminated resins from the make-up andpurification system demineralisers located in the auxiliary building.

•

Recovery and clean-up of equipment: decontamination of systems and equipment used
for clean-up and defuelling activities. Gross decontamination of refuelling tools andclean-up equipment for reuse or disposal was performed in two temporary equipment
decontamination sites.

•

Support activities: various supporting activities to ensure worker safety and to measure
the effectiveness of the clean-up.

Methods used for decontamination of surfaces inside the reactor building were based on the
results of the gross decontamination experiment and subsequent experience gained indecontamination of the auxiliary and fuel handling building. The following decontamination
methods were reviewed by the NRC and were used:•

Abrasive blasting of steel surfaces with particulate driven at high velocity to remove
contamination. This method was especially suited for small or irregular surfaces that were
not compatible with other decontamination techniques. This method was also used in theon-site special decontamination facilities for tools, equipment, and other materials.

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Chemical decontamination of external surfaces of pipes, tanks, and system internals.
The gross use of chemical agents was limited because of the potential for drain-off ofsuch agents to damage water processing systems. This method was also used in the onsite special decontamination facilities for tools, equipment, and other materials.

•

Dry vacuuming to remove powdered contaminants and dried residue. This method wasused especially for water-sensitive components that could not be flushed with water.

•

Low pressure water flush at levels between 100 to 1 000 psi (0.7 to 7.0 MPa), flow rates up
to 25 g/min (95 l/min), and water temperatures up to 170°F (77°C). This method was used
to clean equipment and loose surface debris. Examples included the polar crane, steam
generator housing structures, missile shields, refuelling canal, and refuelling bridge.

•

High pressure water flush to remove unbonded surface coatings at pressures between
2 000 to 10 000 psi (14 to 70 MPa) and flow rates of 4 to 30 g/min (15 to 114 l/min).

•Wet vacuuming to remove puddles of contaminated cleaning fluids after flushing.

•

Ultra high pressure water flush to remove rust, scale, nuclear-grade coatings, and surface
concrete at pressures up to 60 000 psi (410 MPa) and flow rates of 1 to 2 g/min (4 to 8 l/min).

•“Scabbling” of walls and floors to aggressively remove concrete surfaces and surface
coatings. The scabblers used pneumatically operated reciprocating pistons equippedwith tungsten carbide bits to pulverise the concrete surface. A vacuum system with ahigh-efficiency particulate air (HEPA) filter was attached.

•Steam and vacuum decontamination system to decontaminate painted and uncoatedconcrete, ductwork, diamond deck plates, lead bricks, penetration covers, piping, conduit
cable trays, and drain covers.

•

Strippable coatings that involved the application of an organic coating which contained
chemicals to aid in the removal of radioactive contaminants from the surface. As thecoating dried, it cracked and peeled away from the surface.

•

Scrubbing to remove loosely held contamination on floors and walls using manuallyapplied or mechanically driven rags, absorbent cloths, brushes, pads, grit, and chemical
agents.

Contamination was washed from the floors, walls, pipes and other areas using high pressure
water sprays (Figure 2.5) and strippable coatings (Figure 2.6). Contamination was also removed
from concrete using air-operated chisels and hydraulic pounding machines to break up the top
layer of concrete.

Figure 2.5. High pressure spraydecontamination

Figure 2.6. Strippable coating for floor
decontamination

Source: NRC (2016).

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Remote controlled robotic vehicles and supporting control equipment were used extensively
to perform work. These vehicles were both versatile and productive, and proved useful in manydifferent tasks, including video camera inspections, radiation monitoring, sediment sampling,acquisition of concrete core samples, high pressure water flushing (Figure 2.7), concrete scabbling,
and debris pickup and removal. Specially designed robots were sent into high radiation areas to
do decontamination work and obtain information on physical and radiological conditions.
In addition to the remote vehicles, fixed-position, remotely operated tools were developed
for work inside the reactor vessel. The tools included a plasma arc cutting system to remove thestainless steel core support assembly, and several manipulator arms for handling damaged fueland structural components. The use of robots at TMI-2 did not require any NRC licensing actions;however, activities in which robots were used, like most recovery and clean-up activities,
required safety evaluations by the licensee and the NRC. Key robots used at TMI-2 include:
remotely operated video enhanced receiver (ROVER) or remote reconnaissance vehicle toperform video and radiation surveys, collect sludge samples from the floor, collect core samples
from the wall surface, and other reconnaissance activities; LOUIE I remote vehicle to measure
the radiation profiles and to remove loose pre-accident debris and salt deposits on the floor;
LOUIE II remote vehicle to perform remote floor scabbling and pulverise the floor surface while
vacuuming loose concrete; WORKHORSE or remote work vehicle for decontamination anddemolition work in the basement of the reactor building; and a Mini- ROVER which was acommercial submarine vehicle modified to remove larger fuel debris inside the pressuriser.

Figure 2.7. Remote control robotic high pressure
spray washing system

Source: NRC (2016).

By the end of the clean-up programme, the floor contamination levels in most areas of theauxiliary and fuel handling building were reduced to those typical of pre-accident conditions.
In the reactor building, radiation levels in frequently accessed areas were reduced by 85%. In
November 1982, upper-level corridors of the auxiliary building became accessible to workers
without the need to wear anti-contamination clothing. Respirators had not been required forentry into the upper corridors since October 1979. On 28 June 1984, workers entered the reactor
building without respiratory protection for the first time since the accident, and subsequent
entries were made without respirators, in accordance with ALARA principles. However, oncethe defuelling operations began, the hazard from discrete (hot) radioactive particles was a major
concern that required respiratory protection. An elaborate system of administrative protective
zones (successive rings of increasing contamination with the defuelling platform at the centre)
was established to contain hot particles.

Defuelling
The TMI-2 accident resulted in severe reactor core damage and migration of molten corematerials onto the reactor vessel’s lower head. Core damage and relocation occurred within fourhours of the accident initiation, after which long-term cooling stabilised the damaged reactor
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core. Various reactor core inspections and observations from early defuelling activities were
documented in an Idaho National Engineering Laboratory (INEL) report which provided thetechnical basis for planning defuelling approaches and necessary equipment. During the periodof exploration of the damaged reactor core, information from new inspections and observations
altered defuelling strategies and tool designs. Camera inspections established that there was
empty space where the top five feet of the reactor’s core should have been. During the accident,
the top section of the fuel had collapsed into a bed of rubble. Later, camera inspections inside
the reactor vessel confirmed the existence of another bed of rubble at the bottom of the reactor(Figure 2.8).

Figure 2.8. Rubble bed at the bottom of the reactor

Source: NRC (2016).

The information gathered from these inspections was applied to the training of defuelling
operators and the development of new tools and equipment. Operators trained for several
months on a full-scale defuelling work platform (Figures 2.9 and 2.10) constructed outside thereactor building. The training and decontamination work reduced radiation exposure to levels
comparable with levels in the refuelling of a normal operating plant.

Figure 2.9. Full-scaledefuelling work platform

Figure 2.10. Rotating defuelling
platform operation

Source: NRC (2016).

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STATE OF THE ART OF LONG-TERM MANAGEMENT AND ACTIONSThe TMI-2 clean-up effort took about 14 years with a collective manpower effort of over
3.6 million person-hours to complete. The reactor vessel defuelling operations spanned a fiveyear period from October 1985 through January 1990 and involved over two million personhours. A total of about 133 000 kg of fuel, cladding, structural, and control materials were
removed from the reactor vessel during the five-year effort. During July and August of 1991, thereactor vessel was drained to make final measurements of the residual fuel remaining in thevessel. An estimated residual fuel quantity that remained in the reactor vessel following
defuelling was approximately 1% of the original 94 000 kg of uranium oxide fuel inventory. The
distribution of core material as documented in the final defuelling report is summarised below:•

Upper core void: the upper core void or cavity consisted of 42 partially intact fuelassemblies standing at the periphery of the void. The void was about 26% of the originalcore volume and measured 1.5 m deep from the top of the original core to the debris bed.

•

Upper debris bed: the debris bed consisted of 26 000 kg of core material, such as wholeand fractured fuel pellets, control rod spiders, fuel assembly end fittings, broken fuelrods, and resolidified debris. The bed rested on top of the resolidified, hardened mass
and was about 0.6 m to 1 m deep.

•

Resolidified mass: the solid metallic and ceramic mass consisted of about 33 000 kg ofcore material. The mass rested on partially intact fuel assembly “stubs”. The massmeasured about 3 m in diameter, 1.5 m deep in the centre, and 0.25 m deep around theedges.

•

Intact assemblies: partially intact fuel assembly stubs located under the resolidified massand the peripheral standing assemblies comprised about 45 000 kg of core material.

•

Upper core support assembly: the upper core support assembly included vertical baffle
plates that formed the peripheral boundary of the core, horizontal core former plates towhich the baffle plates were bolted, the core barrel to which the core formers wereattached, and the thermal shield. The assembly retained about 4 000 kg of loose debris
and resolidified material. A resolidified crust ranging from 0.5 to 4 cm thick was attachedto the bottom of three core former plates.

•

Lower core support assembly: the lower core support assembly structures retained about
6 000 kg of resolidified material in five layers around the circumference of the structures.

•

Lower head region: the reactor vessel’s lower head region contained about 12 000 kg ofloose core debris and 7 000 kg of agglomerated core debris. The debris on the lower head
was 4 m in diameter and 0.75 to 1 m deep. The surface debris had particle sizes which
varied from those of large agglomerated debris (up to 0.20 m) to those of granularparticles.

•

Fuel debris distribution outside the reactor vessel: a total of about 228 kg of fuel debris
were transported through the reactor coolant system. About 95% of the debris settled in
the reactor coolant system. Approximately 10 kg total were deposited in the reactor
building on the basement floor and sump (5 kg) and make-up and purification system
let-down coolers (4 kg). A total of 23 kg entered the auxiliary building, mainly depositedin the three reactor coolant bleed tanks (a total of 15 kg) and in the make-up andpurification system (6 kg).

Before core debris could be removed from the reactor vessel, preparations were required toallow direct access to the damaged reactor core. While many of these activities were routine
during normal refuelling operations, the effects of the severe accident on reactor vessel
components, the reactor building’s environment, and occupational radiation exposurespresented complex challenges. Preparations included consideration of numerous potential
safety issues; for example, occupational exposures; decay heat removal; criticality control;boron dilution; radioactivity releases; hydrogen evolution inside the reactor coolant system;
pyrophoricity (spontaneous ignition in air) of zirconium dust in the reactor vessel; heavy load
drops; polar crane failure; reactor vessel draining; and fire protection. Technical considerations
included potential distortion; warping or physical dislocation of the reactor vessel’s head or
upper plenum; reactor coolant clean-up; reactor coolant system depressurisation; and lowering

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STATE OF THE ART OF LONG-TERM MANAGEMENT AND ACTIONS

of reactor coolant level. In addition, clean-up of the reactor building, which included
atmospheric gases, basement water, and surface contamination, was an important prerequisite
to ensure lower radiation exposures.
Unique systems and equipment were designed and installed to remove damaged fuel andstructural debris from the reactor vessel. In the early defuelling phase, tools were designed for“pick-and-place” in which debris was picked up and placed into fuel containers (baskets) or
specially engineered defuelling canisters. Some long-handled tools had various hydraulicallyactuated fittings to tackle the larger pieces and smaller bits of debris. A core bore machinepreviously used to obtain samples from the core was placed back into service to bore holes in
the resolidified mass to help break apart the previously molten reactor core. Combinations oftools were used to assist defuelling the lower reactor vessel region, such as the core bore
machine and plasma arc torch (Figure 2.11). Using long-handled tools, operators used
underwater television cameras to remotely load spent fuel canisters with fuel debris.

Figure 2.11. Core boring machine

Source: NRC (2016).

The defuelling systems were designed before the extent of core damage and radiologicalconditions were fully understood. Three types of defuelling canisters (Figure 2.12) andassociated support equipment were specially designed to remove the fuel debris from thereactor vessel and package it for transportation. Water clean-up systems were installed toensure water clarity in the reactor vessel for removing debris and for processing defuelling
canisters. The design of the canisters was decided upon early in the clean-up and dictated thedesign of the defuelling platform and the shipping cask. The design of the canisters was
originally based on the removal of intact fuel assemblies, but later the length was reduced to fit
inside the reactor vessel above the debris bed. Ultimately, very few if any intact fuel assemblies
were removed. The narrow inside dimensions of the cylindrical canister design necessitatedsubstantial cutting of distorted fuel debris in order to load through the opening of the canister.
A radiation analysis programme was undertaken to identify and quantify possible radiation
sources and to design defuelling equipment to achieve dose rate goals in the defuelling area.
The total occupational dose resulting from all clean-up activities was less than 62 person-Sv
over the first 11 years following the accident. The cumulative occupational dose for defuellingand defuelling support activities was below 20 person-Sv. The exposure rate to defuelling
workers averaged less than 0.1 mSv/h. Specific safety issues were addressed in NRC evaluationsof defuelling equipment, systems, and operations.

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STATE OF THE ART OF LONG-TERM MANAGEMENT AND ACTIONS

Figure 2.12. Canisters for fuel material, knockouts and filter tubes

Source: NRC (2016).

The fuel canister was designed as a receptacle for large pieces of core material, which were
picked up and placed either directly into the canisters, or into other containers which would
then be inserted into the canister. The “knockout” canister was designed for use in the fueldebris vacuum system to separate debris particles ranging from about 140 µm up to full pellet
size or larger. The filter canister was designed for use in the fuel debris vacuum system, thedefuelling water clean-up system, and the canister dewatering system. The filter captured
debris dust larger than 0.5 µm on sintered metal filters. Debris buckets were used to configure
the debris before insertion into the fuel canister, to maximise the packing density in a canister.
Two types of disposal buckets, the top loading debris bucket and side loading debris bucket,
were designed to fit into the fuel canisters. An in-vessel vacuum system was designed to pickup smaller fuel debris and pass the same through knockout canisters. Any remaining debris wascollected in filter canisters.
The canister positioning system, a rotating carousel installed in the reactor vessel, was used
to hold up to five fuel and knockout canisters, including knockout canisters for use with thevacuum system. The height of canisters in the canister positioning system could be adjusted tothree discrete elevations to allow them to be placed more closely to the debris bed as the bed
got lower. The fuel handling building’s canister handling bridge lifted the canister from the fueltransfer system to a submerged storage rack or to the canister dewatering station in the spent
fuel pool (Figure 2.13). The defuelling canister storage racks provided storage for loaded
defuelling canisters.
The existing fuel transfer system was modified to transfer defuelling canisters from inside
the reactor building to the adjoining fuel handling building. Canisters were handled in a way
similar to normal fuel assemblies during refuelling operations. The fuel transfer cask used totransfer a defuelling canister from the spent fuel pool to the shipping cask. The Model 125-B
shipping cask was designed specifically to transport the loaded defuelling canisters.

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Figure 2.13. Canisters placed in a submerged storage
rack in spent fuel pool

Source: NRC (2016).

At the completion of defuelling, a report was developed by the licensee to document themeasurements and calculations that were performed to ensure that the plant had been defueled
to the extent reasonably achievable and that the potential for a nuclear criticality had been
precluded during normal and accident conditions. The major objective of TMI-2 post-defuelling
activities focused on preparing the plant for long-term storage. The licensee called the period
preceding the ultimate disposition (either refurbishment and restart or decommissioning) of theplant “post-defuelling monitored storage” (PDMS). During PDMS, the TMI-2 facility would be inlong-term monitored storage, similar to the decommissioning mode SAFSTOR (mothballingwith delayed dismantling), in which the facility is secured, monitored, and maintained in amanner that ensures the protection of the public health and safety for an extended period.
Transitioning to PDMS required the following conditions: i) criticality was no longer possible;ii) potential for fission products movement was eliminated; iii) fuel was removed and shipped
off-site; iv) radioactive waste was shipped or stored; v) radiation levels were reduced
commensurately with the need for access to permit continued plant monitoring and to support
plant-disposition decisions; vi) water was removed from plant systems and spaces, and thepotential for reintroduction of water was precluded; and vii) a safe, monitored plant condition
was established.

Disposal and long-term storage of fuel and other radioactive wastes
The major objective of TMI-2 post-defuelling activities focused on preparing the plant for
long-term storage. The licensee called the period preceding the ultimate disposition of the plant
PDMS. During PDMS, the TMI-2 facility was in long-term monitored storage, similar to thedecommissioning mode SAFSTOR (mothballing with delayed dismantling), in which the facility
was secured, monitored, and maintained in a manner that ensured the protection of the publichealth and safety for an extended period. Placing the TMI-2 facility in monitored storage wasbeneficial in that it eliminated any possible impact of TMI-2 decontamination anddecommissioning efforts on the operating TMI-1 facility.

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STATE OF THE ART OF LONG-TERM MANAGEMENT AND ACTIONSThe TMI-2 accident and subsequent clean-up challenged the management of various forms
and concentrations of radioactive waste. The management of highly contaminated water, fueldebris, and related solid waste by-products included handling, processing, temporary on-sitestorage, transportation, and final disposal. Decontamination activities resulted in substantial
quantities of contaminated water and organic resins and inorganic zeolites produced fromwater-processing systems. Fuel debris that spread throughout the plant created uniqueradiological waste characteristics, and some had to be removed manually using chisels andother handheld tools (Figure 2.14). Also, some waste did not fit into established regulatory
waste-classification categories for transportation and disposal.

Figure 2.14. Manual debris clean-up operation

Source: NRC (2016).

Several on-site facilities were constructed for temporary storage of solid radioactive waste
products from clean-up activities that were being readied for transportation (Figure 2.15). Solid
waste included spent EPICOR-I and EPICOR-II resin liners; contaminated clothing, tools, andequipment; and decontamination materials. The spent fuel canisters were stored in wet pools
until technology was developed to dry the fuel; subsequent to repackaging and storing in dry
storage casks.

Figure 2.15. Temporary solid waste storage facility
for radioactive wastes

Source: NRC (2016).

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The canister handling and preparation for shipment programme included all activities
necessary to prepare and transfer a loaded defuelling canister from its storage rack in the spent
fuel pool to the shipping cask; to insert the canister into the shipping cask; and to verify thatthe shipping cask was prepared for transport in accordance with its NRC-issued certificate ofcompliance. Canister preparations included dewatering and purging the defuelling canisterwith an inert cover gas, verification of final canister weights, verification that the catalytic
recombiners installed inside the canister were functioning, and verification that the canister
had been dewatered sufficiently to ensure that the catalytic recombiners remained operable
regardless of canister orientation.
The canisters containing the fuel core debris were shipped in specially designed rail casks
to a Department of Energy facility in Idaho for storage. The fuel casks (Model 125-B) consistedof five major components: the outer containment vessel, inner containment vessel, upper andlower canister impact limiters, canister shield plugs, and cask impact limiters (Figure 2.16). Eachcask had its own transportation system of a skid and rail car. Gross shipping weight of theshipping cask was about 183 000 pounds (83 000 kg). Up to 21 defuelling canisters could be sent
in a single rail shipment (seven canisters per cask, three casks per shipment). The fuelshipments began on July 1986, and were completed on May 1990. There were 22 rail shipments
for a total of 342 canisters of core debris transported to the INEL. The final fuel shipment fromThree Mile Island to the INEL started on 15 April 1990.

Figure 2.16. Model 125-B rail shipping cask used
for transporting defuelling canisters

Source: NRC (2016).

Once a rail shipment arrived at the Central Facilities Area at INEL, both impact limiters were
removed from each end of the shipping cask and the cask and skid were lifted from the rail car
and placed on a tractor trailer. The cask was then transported to wet storage at the INEL Test
Area North. Here each canister was removed from the shipping cask, filled with water, placed
into a storage module, and transferred to the storage pool in a designated location. A vent tube
was installed on each canister for continuous venting. The empty shipping cask was surveyed
for contamination and prepared for rail shipment as regular freight back to TMI-2. The
342 stainless steel fuel canisters of core debris were stored in underwater storage from 1986 to
2001 at the INEL. During the 2000 to 2001 period, these canisters were transferred to the TMI-2
independent spent fuel storage installation (Figure 2.17), also located at INEL, for interim storageof the TMI-2 core debris.

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STATE OF THE ART OF LONG-TERM MANAGEMENT AND ACTIONS

Figure 2.17. Independent spent fuel storage installation at INEL

Source: NRC (2016).

Worker protection
Post-accident radiological conditions at TMI-2 were substantially different from those normallyencountered at commercial operating nuclear plants because of the magnitude and specific mixof the radionuclide contamination. Radiation surveys made shortly after the accident showedthat general area radiation readings ranged from 1.5 to 5 mSv/h in the fuel handling building,and 0.5 to 50 mSv/h in the auxiliary building. Hot spots were measured in the auxiliary building
reaching up to 1.25 Sv/h, and exceeding 10 Sv/h in some cubicles. During the first entry into thereactor building in July 1980, dose rates at the 305-foot (93 m) entry-level elevation ranged from4 to 6 mSv/h. Localised areas of high radiation were measured at 180 mSv/h over the open
stairwell and 20 to 50 mSv/h at floor drains. Surveys performed during the second entry at the
next-higher level, the 347-foot (106 m) operating-floor elevation, showed general radiation
readings of 1 to 4 mSv/h. Below the 305-foot (93 m) entry-level elevation was the 282-foot (86 m)
basement-level elevation, which was flooded with highly contaminated water and sludge.
A telescoping radiation detector was inserted down through one of the reactor building
stairwells and measured 400 to 450 mSv/h at 5 to 7 feet (1.5-2.1 m) from the surface of thebasement water. Once the water was drained and processed through the submerged
demineraliser system, dose rates from the remaining sludge ranged from 10 mSv/h to 10 Sv/h
per hour, depending on location and distance from the floor.
A concern at TMI-2 was high-energy beta contamination from fission products in the reactor
coolant. Areas in the auxiliary building that had experienced coolant leakage were measured in
the 0.10 to 1 Sv/h gamma range with associated beta dose rates in the 1.0 to 100 Gy/h range.
Similar gamma-to-beta ratios were measured on surfaces in the reactor building. The highenergy beta emitters present at TMI-2 required special radiological protection practices, such as
monitoring equipment, personnel dosimetry, heavy protective clothing, procedures, and training.
Many programmes and activities helped to improve radiation protection practices at TMI-2.
New approaches were needed in a number of basic worker protection and dose reduction areas,
including protective clothing, respiratory protection, dosimetry, radiation field andcontamination characterisation, exposure-tracking systems, dose reduction planning,procedures, training, and robotics. In the summer, when the temperature in the reactor building
approached 33°C (90°F), an ice vest was initially used by workers to control heat stress andextended work period. An extensive heat stress programme and protocol was established that
pioneered the use of cool suits and similar technologies. Also, an air conditioning system wasinstalled inside the containment. Several respiratory protection breathing apparatuses were
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developed or adapted to extend stay-times in the reactor building, including a power airpurifying respirator and a power air-purifying hood. Some of the radiological protection
programmes and activities are summarised below:
Dose reduction techniques included i) shortening the transit time of workers in the reactor
building by opening both personnel airlocks and modifying the ingress/egress paths;ii) decontaminating by water flushing discrete radiation sources, such as the air coolers, elevator
shaft, and enclosed stairwell; iii) eliminating other discrete radiation sources by removal of trash
and contaminated equipment; and iv) placing shielding at the 305-foot elevation (93 m), such aslead curtains around the core flood tank, lead sheets on the covered floor hatch, and watercolumns and bladder shields around the open stairwell, elevator, and enclosed stairwell. Someof the more complex dose reduction activities included extensive pre-task planning and mockup training for each task, decontamination of selected surfaces with chemicals, removal of paint,
and scabbling (the mechanical removal of a thin layer) of concrete floors. These efforts resulted
in significant reductions in the dose rate in the reactor building. In July 1984, workers entered thereactor building without respiratory protection for the first time since the accident, and, in
accordance with ALARA principles, subsequent entries were made without respirators.
Eventually, the most effective reduction in the dose in the reactor building was the efforts
that reduced dose to the defuelling crew on the defuelling platform. This included establishing apath to the work platform by shortening, cleaning, and shielding the path; and thoroughly
reducing the dose in the area or the work platform where the defuelling operators worked. The
dose reduction effort did not try to decontaminate the entire reactor building, but focused on
areas where people worked. As the result, cumulative exposure of the defuelling campaign was
roughly only one-third of the total exposure during the first 11 years following the accident.
Special instrumentation, systems, and techniques were developed or modified to measure
and characterise the unique radiation situation at TMI-2 for ensuring worker safety anddetermining the effectiveness of decontamination processes. Key instruments included:i) thermo-luminescent dosimeter pseudo cores to take beta radiation measurements of thebuilding floors; ii) wall and floor sampler to mill the concrete surface and collect the sample in a
filter for off-site analyses; iii) modified handheld ion chamber detector to provide
omnidirectional detection for gamma measurements; iv) modified handheld tungsten-shielded,
Geiger-Mueller detector with a conical lead collimator on the face of the probe for rapid andaccurate directional exposure measurements; and v) mobile radiochemistry laboratory to
perform transuranic and radionuclide analyses of high activity liquid and solid samples.
In November 1982, upper-level corridors of the auxiliary building became accessible toworkers without the need to wear anti-contamination clothing (respirators had not been required
for entry into the upper corridors since October 1979). The following month, the auxiliarybuilding’s corridors in the basement (at the 281-foot elevation, 86 m) became accessible without
the need to wear respiratory protection masks. Overhead areas, such as ceiling and cable trays,
were decontaminated only to the extent that they would not re-contaminate the floor below. In
such cases, a radiation work permit was required to access ceiling areas.
Although worker activities at TMI-2 have been quite different than those at operating power
plants, the cumulative doses at TMI-2 since the accident had been lower than the average doses
experienced at operating reactors. By the end of 1989, with the clean-up about 99% completed,the collective dose to all workers was 62 person-Sv. This was comparable to the collective
occupational radiation exposure that was estimated in the original PEIS.

Insights and lessons learnt
Because the Three Mile Island accident was the first time that a large power plant experienced
severe core damage much of the necessary equipment, procedures, and regulatory mechanisms
to handle both the short-term and long-term aspects of the accident simply did not exist. Mostof the equipment and procedures for long-term management activities had to be specificallydesigned and developed following the accident.
The TMI-2 clean-up effort took about 14 years, and the reactor vessel defuelling operations
spanned a 5-year period. A total of about 133 000 kg of fuel, cladding, structural, and control
materials were removed from the reactor vessel during the five-year effort. Unique systems and

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STATE OF THE ART OF LONG-TERM MANAGEMENT AND ACTIONS

equipment were designed and installed to remove damaged fuel and structural debris from thereactor vessel. The clean-up activities challenged the management of various forms andconcentrations of radioactive waste. The management of highly contaminated water, fueldebris, and related solid waste by-products included handling, processing, temporary on-sitestorage, transportation, and final disposal. Decontamination activities resulted in substantial
quantities of contaminated water and organic resins and inorganic zeolites produced fromwater processing systems. Fuel debris that spread throughout the plant created uniqueradiological waste characteristics. Also, some waste did not fit into established regulatory waste
classification categories for transportation and disposal.
At the completion of defuelling, it was shown that the plant had been defueled to the extent
reasonably achievable and that the potential for a nuclear criticality had been precluded during
normal and accident conditions. The major objective of TMI-2 post-defuelling activities focusedon preparing the plant for long-term storage. The period preceding the ultimate disposition of
the plant was called “post-defuelling monitored storage” (PDMS). During PDMS, the TMI-2
facility would be in long-term monitored storage.
Due to the unique situation and conditions, many activities related to the long-term
management of TMI-2 accident were new, and new methods and processes had to be developed.
The following insights can be mentioned.



New processes and communication with the public•

The DOE, in conjunction with other Federal agencies, implemented national policy
affirmed by two US presidents, in support of activities related to the clean-up of theTMI-2 accident. The DOE activities, including acceptance and transport of the TMI-2 coredebris, were frequently reviewed by Congressional committees through the process oftestimony on technical progress and DOE budget authorisations. Accordingly, there wasapproval at the highest levels of government for the clean-up activity.

•

New regulatory practices were applied as changes in the facility’s post-accident mode ofoperations required unique regulatory and licensing actions. In order to properly reflect
evolving plant status, the NRC-issued orders, modified those orders, approved licenceamendment requests, and granted relief from certain regulatory requirements. The NRC
granted exemptions from certain requirements of the regulations for nuclear power
plants, but only under special circumstances as permitted in the regulation. Exemptions
were necessary at TMI-2 because of the plant’s damaged configuration and changing
status during clean-up.

•

The TMI-2 programme did a credible job in preparing a public relations plan before thecampaign started. The programme accepted and enhanced all established public
relations procedures, prepared and distributed programme briefs and videos, co-hosted
a media day, made public announcements, performed pre-notification activities, andmet with some state and public officials.

•

Gaining public trust was a significant challenge. In many cases it was more difficult toobtain public trust and acceptance than to find a technical solution. One case was thedisposition of tritium-contaminated water. Although a technical solution could be found
for the safe disposition of tritium-containing waste water it was more difficult to obtain
public trust and acceptance for its disposal.

Decontamination and management of radioactive waste and effluents•Management of large amounts of contaminated water was one of the major challenges.
A large volume of accident-generated contaminated water accumulated in-plant tanks
and containment basement floor and sump. Although most of the radioactive
contaminants could be removed from water, tritium could not. This water was stored
and used for plant decontamination purposes and shielding of equipment located in the
spent fuel pool (no spent fuel at TMI-2). Eventually, this water was evaporated over a
2.5-year period. The evaporation was completed 13 years after the accident.

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

34

•

Shortly after the core degradation, gaseous fission products collected in the reactor
coolant in the liquid clean-up system caused serious airborne contamination inside theauxiliary building and required mitigative actions including diverting effluents into
containment, designing new filtration and ventilation systems, and venting containment
gases.

•

Decontamination posed challenges as radioactive contamination penetrated many
concrete and other porous surfaces requiring surface removal for decontamination.
Eventually, with the various decontamination methods the contamination levels in theauxiliary building were reduced to those typical of pre-accident conditions.

•

Recontamination limited decontamination efforts. Sources of recontamination included
wastewater from flushing, migration of radioactive materials through epoxy coatings,
and redistribution of radioactive particulates by the air handling system.

•

All of the fuel and debris removed from accident was shipped off-site for federal
sponsored research and temporary storage until a permanent waste disposal site isestablished. The TMI-2 site was decontaminated to the extent that any potential for asignificant release of radioactive material to the environment was eliminated.
Radioactive material was removed and other sources of radioactivity were isolated sothat any potential radioactive release was to be within 10 CFR 50 Appendix I, “Numerical
Guides for Design Objectives and Limiting Conditions for Operation to Meet the Criteria‘As Low as Reasonably Achievable’ for Radioactive Material in Light-Water-Cooled
Nuclear Power Reactor Effluents”, guidelines for off-site dose consequences. Radiationlevels inside most areas of the plant were reduced to levels through decontamination toallow for safe entry for routine monitoring by survey teams. The water processing
programmes removed all water containing forms of radiation to the extent practical.
Aregular monitoring programme was established to ensure the continued safety of theplant in the post-defuelling monitored storage state.

•

The shipping campaign of the debris from Three Mile Island to Idaho was a very successfulproject. It has been used as a model for transporting spent fuel since then. There were anumber of things which helped the success of the shipping campaign. Keys to success were
teamwork, attention to detail, emphasis on prevention of and/or early detection ofproblems, and a stringent quality assurance aimed at preventing problems.

Risk management and occupational safety•

Defuelling safety concerns included criticality, pyrophoricity, radioactive gas and dust
production, radiolytically generated hydrogen and oxygen, and steam generation in-coredebris canisters in the event of a fire. Defuelling operations took precedence over datacollection requiring prioritisation in data to be collected.

•

Because the exact severity of core damage was unknown at the time and the operators
had no way of measuring the criticality margin the control rods could not be relied upon
for reactivity control and assurance of shutdown. Therefore, a high boron concentration
had to be maintained in the coolant to ensure that the core would not become criticalgiven the knowledge that was available. Boric acid was chosen from different potential
poisons as part of criticality control but it also raised concerns of boron dilution events.

•

Potential pyrophoric of metallic zirconium (spontaneous combustion) was an early
concern during core cutting operations. Evaluations concluded that the formation of
zirconium powder during the accident was extremely unlikely due to the dynamics ofthe accident. In addition, testing of residue material from the reactor coolant system,
control rod mechanism leadscrew, and reactor vessel plenum cover resulted in no
pyrophoric reaction.

•

In-containment dose reduction was emphasised over decontamination for defuelling
operations since early decontamination efforts proved to not be effective. Dose reduction
techniques involved installation of shielding, choosing entry and transit routes based ondose, and use of decontamination techniques where necessary.

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STATE OF THE ART OF LONG-TERM MANAGEMENT AND ACTIONS•The use of robotics to perform monitoring and decontamination reduced radiation
exposure to workers. The use of robotics could also enable work to be performed inradiation fields that were too high for humans to practically work in.

•

For defuelling, a radiation analysis programme was undertaken to identify and quantify
possible radiation sources and to design defuelling equipment to achieve dose rate goals
in the defuelling area. Training and decontamination work reduced radiation exposure
to levels comparable with levels in the refuelling of a normal operating plant. This
involved operators training to defuel using a full-scale mock-up to work out potential
issues and to minimise time involved in actual defuelling. The mock-up was also used
to test defuelling tools before being used in the containment. The mock-up included aworking defuelling work platform, large water tank and a quarter section of the massive
lower core support structure.

•

Gaseous fission products which could lead to increase personal exposure was a concern
during defuelling. An off-gas system created airflow that prevented radioactive gases
from collecting under the work platform from reaching personnel working on theplatform.

•More than 99% of the fuel and debris in the reactor vessel was removed to eliminate thepotential for re-criticality.

•

Heat stress was an issue for workers in protective gear especially in summer months.

Method and equipment development•

Special instrumentation, systems, and techniques were developed or modified tomeasure and characterise the unique radiation situation at TMI-2 for ensuring worker
safety and determining the effectiveness of decontamination processes.

•

As-built dimensions of the TMI-2 reactor vessel internals did not always match designed
dimensions. This required inspections of the as-built dimensions to ensure that recovery
equipment would work.

•

To ensure that spatial material information about different core materials and fission
products was obtained, DOE initiated a sampling programme using a core boring machine
since this information would have been lost by the proposed defuelling processes.

•

The core bore samples provided insights into fission product release from the fuel,fission product retention in the core, maximum temperature during the accident, andreactor core material interactions. The core bore machine later proved useful to break up
the solidified monolith in the core region and drill through and cut portions of the lower
core support assembly.

•

It was difficult to obtain data about core conditions sufficient to define debris removal
tooling requirements, Camera surveys were insufficient for this purpose. Grab samples
were required. Additional tools were needed (handheld and pneumatic tools, core bore
machine, plasma arc cutting torch) when unexpected conditions were encountered.

•

The melt solidified into a heterogeneous mass of substantially-oxidised core constituents(fuel [uranium, zirconium], structural materials [Fe, Ni, Cr], and control rod materials [Ag,In, Cd]) that resisted efforts to break it apart. Having been unsuccessful with breaking thefrozen mass apart with impact chisels and wedges, the core bore drilling machine thatwas originally used to retrieve bore samples from the core region was refitted to do so.

•

The plasma arc cutting tool, which chosen over several other methods for cutting up
components and material in-vessel, experienced some difficulties including frequent
torch burnouts, difficult starting, loss of torch functionality due to borated water ingress
requiring removal and reassembly, torch gases lifting debris into cutting tool positioning
machinery requiring disassembly and cleaning of the machinery.

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STATE OF THE ART OF LONG-TERM MANAGEMENT AND ACTIONS•The extent of degradation of the core was not known prior to defuelling. As suchdefuelling canisters were prepared for both intact fuel bundles and debris. It was difficult
to estimate the number of defuelling canisters required because less mass was loaded
into each canister than expected. Around 340 canisters were used. The initial estimate
was for 240 canisters.

•

Although using robotics for defuelling was considered, a manual “pick-and-place”
method was chosen since it was more flexible and years were expected for thedevelopment of the proposed robotic system. An existing robotic arm was brought onsite but not used due to complexity in use, the potential for damage to controlling wires
and hoses, and expected difficulties in decontamination.

•

The hydraulic fluid in the defuelling hydraulic system was changed from one containing
boric acid to one containing a borate ester mixture because boric acid precipitation was
damaging hydraulic tools.

•

The transport equipment interfaces, both at the Idaho National Laboratories and TMI-2,
took a lot of planning, teamwork, and honest and open communication. The integrated
test of all cask handling and cask loading equipment was very valuable for confirming
cask-to-handling equipment fit up, training of clean-up personnel, development ofprocedures, and generally proving system performance. The test resulted in a muchsmoother installation and start-up of equipment at the TMI-2 facility.

•

In order to keep the fuel rubble in a safe physical dimension (for criticality safety), a filler
material needed to be used to ensure this physical configuration. It was decided to use alow density concrete for this filler material. The use of a low density concrete in thecanister as filler material was not the best choice. The presence of water in the concrete
mixture resulted in a surplus production of hydrogen that had to be vented. As such, thecanisters could not be permanently sealed during subsequent storage.

Synthesis of relevant Chernobyl experience
The accident at unit 4 of the Chernobyl NPP (Figure 2.18) at 01:23:40 of 26 April 1986 was one of
the most serious accidents in the history of nuclear power. This reactivity initiated accident(NEA, 2010; De Geer et al., 2017) resulted in huge and rapid increase of neutron power that
disrupted reactor and destroyed significant part of the reactor building (Figure 2.18a). Manyother structures, safety barriers and safety systems were also affected.

Figure 2.18a. Ruined Unit 4 after graphite fire ended

Source: Krasnov et al. (2016).

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Figure 2.18b. Photo of the destroyed
reactor in the first hours after theaccident: April 26, 1986

Source: Krasnov et al. (2016).

Figure 2.18d. Installation of supporting
structures of the original confinement(shelter), view during shelter construction

1) Girder B1; 2) Roll-up of pipes; 3) Reinforced upper part of thewall along the axis #50; 4) Ventilation shaft; 5) “Mammoth” girder;6) West pillar of the “Mammoth” girder; 7) East pillar of the“Mammoth” girder; 8) “Octopus” girder.
Source: Krasnov et al. (2016).

Figure 2.18c. Photo made fromhelicopter on May 3, 1986

Source: Krasnov et al. (2016).

Figure 2.18e. External protective structures ofshelter before new confinement installation

1) Roof of reactor hall; 2) Roof of turbine hall; 3) Cascade wall; 4) West(large) counterfort wall; 5) South shields; 6) South “brassy” shields; 7)
North “brassy” shields; 8) North (small) counterfort wall.
Source: Krasnov et al. (2016).

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STATE OF THE ART OF LONG-TERM MANAGEMENT AND ACTIONS

Figure 2.18f. View of the new confinement that covered
the old one on 29 November 2016

Source: Krasnov, ISPNPP, http://dazv.gov.ua/images/pdf/ukraino-japonskyj-komitet/fifth/Viktor%20KRASNOV.pdf.

Extremely high releases of fission products, long-distance transport of aerosols and hot
particles, and global contaminations of land were stimulated by reactor explosion and fine fuelfragmentation on 26 April 1986 as well as by the strong graphite fire which resulted in very highplume visible from 27 April to 9 May (Figure 2.18b and 2.18c).
A significant fraction of the nuclear fuel (3.5%) was released, most of which was depositedin the vicinity of the damaged plant. There is some discrepancy in element specific releasefractions published in the literature (Abagyan et al., 1986; IAEA, 1986; Legasov et al., 1988; Lewis,
1986; llyin and Pavlovskij, 1988a, 1988b). Recent published data (Borovoy and Velikhov, 2012)indicate the following releases:•

noble gases: 100%;•

iodine radionuclides, including 131I: (50-70)%;

•

137

•

Cs and 134Cs: (33 ± 10)%;low volatile nuclides, such as 90Sr and 144Ce, and non-volatile nuclides, such as 239Pu andAm, in the dispersed fuel – up to 5%.

241

In the long term, 137Cs and 90Sr are the most hazardous nuclides contributing to thecontamination of land. Thousands of square kilometres of land were affected with the highest
contamination of Cs deposited by rainfalls.
Contamination by dispersed fuel worsened the radiological situation on the site during theearly phase. Besides 137Cs (γ-emitter) and 90Sr (β-emitter), the most hazardous isotopes over the
first 10 to 20 years after the accident are Pu isotopes (α-emitters) and then 241Am (241Pu
transforms into 241Am by β-decay).
About 30 years have passed since the accident and work on the mitigation of its
consequences has been performed continuously over the years. Many aspects rendered thelong-term mitigation of Chernobyl accident highly challenging, specifically during the
emergency phase right after the accident. They were:•

38

High on-site contamination levels due to fuel fragments and dust dispersal. Even if thefuel fraction released from the damaged unit was limited, the high specific activities offragments and dust resulted in very high dose rates.

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Highly damaged structures and systems due to explosions and fires, complete loss of themeans of surveillance.

•

Facing a completely new situation, absence of adequate protection devices, remotedevices for cleaning the site, systems for removing rubble and finally adequate vehicles
to transport waste materials.

•

Difficulties to assess existing risks (radiological, criticality, thermal [i.e. corium-concreteinteraction]) with the knowledge existing at that time.

Characterisation of dose rates, contaminations and releases
According to Borovoy and Velikhov (2012), there was a lack of reliable information about dose
rate at the site during the first hours after the accident because:•

The operating staff had no appropriate dosimeters. They had dosimeters with an uppermeasurement limit of 3.6 R/hr (0.01 mSv/s). Dosimeters with wider measurement ranges
were locked in a room that operators were not able to open.

•

The NPP top management was reluctant to accept and broadcast that a severe accident
had occurred in unit 4. During the night of 26 April, the plant director reported toauthorities: “Reactor survived; we continue water cooling; the radiation level is at thenormal level”. In fact, the results of the radiation measurements carried out by the headof rescue team in the vicinity of unit 4 indicated the upper limit of 200 R/hr (0.556 mSv/s)of the DP-5 dosimeter he was using. Early morning of the next day, the director reported
to the authorities that radiation was at the level of 3-6 R/hr (0.0083-0.0167 mSv/s) andthat the reactor survived.

This miss-information resulted initially not only in a wrong understanding of the danger
posed by the accident to the public but also in deaths of several people who tried to inspect theunit either on their own initiatives or at the request of the management of the plant.
The accurate in-site radiation monitoring started at 14h00 on 27 April, i.e. about 13 hours
after the beginning of the Chernobyl accident. A team of external experts came to Pripyat, thetown where the plant personnel lived. The first dose maps prepared with participation of
military staff showed dose rates between 100 and 1 000 R/hr (0.278 and 2.778 mSv/s) at different
locations close to unit 4. Twenty-nine dosimeter units were established on 29 April in order toimprove radiation monitoring. Later 36 dosimeter stations were created under the supervisionof the Soviet Army.
Extensive efforts were made during the active phase of the accident to sample the air above
the reactor and in the vicinity of the site. For many reasons (unstable nature of the releases,changes in weather, high radiation fields, work undergoing around the damaged unit), theaccuracy of the measurements was not high. However, these samplings provided a significant
result: besides the volatile fission products releases (noble gases, iodine, caesium and tellurium
isotopes), most other nuclides (niobium, zirconium, ruthenium, cerium) were also released as
small solid particles.
Efforts were also made during the active phase of the accident to measure depositions offission products around the site. Although the dose rates could be easily measured, thecontribution of γ, β and α emitters could not be easily determined as these required complex
radio-chemical analyses. However, it was established readily that the main part of the fuelparticle fallout occurred in the vicinity of the damaged plant and that the on-site contamination
was essentially due to those fuel particles.
Studies were conducted to determine the amount of fuel remaining in the damaged unit.
Based on many measurements and samplings, it was concluded that approximately 3.5% of thefuel was released and deposited around the damaged plant with 0.3% on the site and 1.5% within
the exclusion zone. Even though it was a small fraction of the fuel inventory, fuel particles,
pieces of assemblies and fuel pins on the site were the main contributors to the measured dose.

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STATE OF THE ART OF LONG-TERM MANAGEMENT AND ACTIONSExtensive off-site measurements were organised by the government with participation of
1 130 institutions of Ukraine. Fifteen thousand dosimeter stations and 94 mobile laboratories werecreated (Kachalovsky, 1989). The measurements obtained from this extensive network enabled:•

Decision making for evacuation of people from the city of Pripyat, from a 10 km zone andfinally from a 30 km zone, where land surface contamination by Cs, Sr and Pu exceeded
15, 3 and 0.1 Ci/km2 (555 000, 111 000 and 3 700 Bq/m2) respectively. With these criteria forevacuation, 49 360 people were evacuated from Pripyat on 27 April 1986, 10 799 people
from the 10 km zone on 2 May and 3 and 30 136 people from the 30 km zone between
4 and 7 May. Altogether, almost 100 000 people from 49 settlements were evacuated in1986. Between 1989 and 1991 additional evacuation of people from 11 settlements was
performed: the criterion used was that of a dose limit of 350 mSv for the whole life of aperson.

•Planning and realisation of the radiation decontamination work.

•

Determination of the protection measures in the so-called “controlled zones” which would
be much less contaminated. Such controlled zones were created in 77 settlements werepeople continue to live.

More details about the on-site and off-site radiation measurement results and dataassessments may be found in IAEA, 2006; Izrael et al., 1990; Rimsky-Korsakov et al., 2009; Ivanov
et al., 1994; Kononovich et al., 1994; Borovoi, 1989; Devell et al., 1986; Kashparov, 1994; Borovoy,
1996b; Izrael, 1998.

Countermeasures undertaken in the short term and their effects
Implementing adequate countermeasures during the emergency phase was found to beparticularly difficult since it was not possible to understand the situation and to assess theefficacy of countermeasures.
In the immediate days following the accident, authorities largely feared further
uncontrolled escalation of the accident. They therefore focused on firefighting, the preventionof re-criticality and of the penetration of molten corium into the pressure suppression pool,located below the reactor cavity in this particulate design. It was feared that melt could
penetrate further through the basement into the soil, to pose the risk of groundwatercontamination. Many countermeasures were implemented to reduce these risks. They resulted
in high exposure of radiation to personal involved in the countermeasure operations.
The large radioactive releases occurred from 27 April to 10 May 1986 (Figure 2.19). Duringthat period, a large amount of different materials was dumped from helicopters on the top of
the destroyed reactor. A total of 5 000 tonnes of materials were dumped until 10 May (Krupny,
1996). These included 40 tonnes of B4C to reduce re-criticality risk, 800 tonnes of dolomite tostop the graphite fire, 1 800 tonnes of clays and sand on top of the debris to reduce radioactive
aerosols emission and 2 400 tonnes of lead to cool down the molten fuel. Approximately
14 000 tonnes of solid materials were dumped in April-May 1986. Besides these, 140 tonnes ofliquid polymers and 2 500 tonnes of tri-sodium phosphate were dumped in the open top of thedamaged reactor to reduce formation of radioactive dusts.
As it is shown in Pasukhin (1997), and as it is common in almost all helicopter drops, only asmall fraction of these materials reached the target. However, it helped to stop the graphite fire,
which was maintaining high temperatures for the release of the volatile fission products fromdamaged core. The environmental releases and dose rates above the debris and in the vicinityof reactor building were dramatically reduced. Perhaps natural phenomena, such as reduction
in decay heat with time, natural cooling of debris and the reduction in the amount of graphiteavailable for burning in the hot zone, also helped to decrease the radioactivity releases(Figure 2.19).

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Figure 2.19. Kinetics of integrated daily radioactive releaseduring the initial phase of the accident

First peak of release

Notes: Estimated uncertainty is ± 50%. Activity is normalised for the radioactive decay
at 6 May. 1Ebq (Exa-Becquerel) = 1 018 Bq.
Source: Adapted from OECD (2002).

Dumping of large masses of solid material from high altitude had also several negative
effects (Kluchnikov et al., 2011). It damaged roof plates of turbine hall and supporting structuresof de-aerator vessel. It also created additional radioactive dust, which deposited in the north ofthe destroyed reactor building (Petelin et al., 2003) and formed a new area for decontamination.
The risk of steam explosion after melt penetration into the compartments, which wereoriginally filled with water, was found to be low, as it was determined that these compartments
were almost dry by the time of corium melt arrival.
It was decided to construct a protective concrete slab below the reactor building basement
to reduce the risk of groundwater contamination. Its construction started on 3 May 1986. A totalof 388 mine-workers were involved in the construction of a tunnel of 138 m length and 1.8 m
diameter which was used for radiation protection of workers and concrete slab building.
Extraction of soil and pouring of ~2.5 m thick concrete slab having lateral dimensions of ~30 ×
30 m was accomplished. A water cooling system made of a layer of 100 mm diameter tubes anda temperature control system were installed inside the concrete slab to prevent structure
ablation by the core melt. This work was completed by the end of June 1986. Later on, it wasseen through boreholes that only one concrete horizontal wall, out of the four in the reactor
building on the path of the corium melt/debris had been eroded by the melt.
Construction work was performed to provide passages and the site was cleaned before themitigation measures were started. This phase was very tough since no adequate protection forworkers and equipment was available to reduce radiation exposure to acceptable levels. The
situation was continually improved by installing protection and remote devices. The site clean-up

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consisted of: i) removing rubble and damaged equipment, ii) decontamination of external surfacesof buildings, iii) removal of the upper layer of soil (5 to 10 cm), iv) laying concrete slabs and clean
sand and gravel on the ground and v) covering the surfaces by films. Thanks to these measures,the site dose rate was reduced to 5-10% of the pre-existing dose by 10 May.
At the end of May 1986 additional on- and off-site measures were undertaken for furthersite decontamination and preparation of the restart of Chernobyl units 1 and 2. These included
construction of decontamination facilities, preparation of temporary storage places forradioactive wastes, protection of rivers and lakes, construction of new roads and supply systems,
and construction of the isolation walls between unit 4 and unit 3 (Kochetkov et al., 1989).
Later on, extensive construction work was performed to prevent the contamination of the
Pripyat and Dniepr rivers, on whose banks about 32 million people lived. The most critical times
for possible contamination of these rivers were the autumn of 1986 and the spring of 1987. The
Soviet Government constructed a total of 131 filtering and isolating dams (a) in the lower part of
Pripyat River (b) along the Dniepr River and (c) along the Kiev artificial lake. A total of 5 million
tonnes of soil were employed in these construction projects. In addition, four bottom traps andfive underwater dams were also built. These measures reduced contamination of hydrosphere(Kachalovsky, 1989) to about 15% of the pre-existing level.
An additional measure (Borovoy and Velikhov, 2013) for protection of rivers and lakes wasbuilding of the ground wall around the NPP site, i.e. leak-tight wall in the soil having design
thickness of ~1 m, depth of 30 m in some places and length of 8.5 km, to isolate the hydrospherefrom the contaminated groundwater. However, the works have been stopped after confirmationof no melting through of the fundament plate of unit 4 and new evidences of insignificant
contamination of Pripyat River with the groundwater. The main contamination source wasproven to be surface washing and surface contaminated waters. Only 2.8 km of wall were built
along the river. Later, it was found that this wall noticeably increased the natural level ofgroundwater at the site.
The above-listed projects resulted in quite high exposures for the civil and military personnel
engaged in these projects. Altogether 6 000 persons participated in this urgent (day and night)
decontamination and emergency work. Two thousand of the six thousand persons reached theirradiation dose limit of 25 rem (250 mSv) (Nosovsky et al., 2006) before the beginning of September
and they were replaced. Extremely high collective dose of 22 877 person-rem (228.77 person-Sv)
is reported by Kochetkov et al. (1989) during the 8 months of 1986, but in 1987 and 1988 it was
reduced, respectively, to 7 242 and 4 953 person-rem (72.42 and 49.53 person-Sv). The number of
persons receiving the 5 rem (50 mSv) dose limit was reduced by a factor of 12 between 1986 and1987. Almost a factor of 47 reduction is reported for the exposures above 25 rem (250 mSv) during
the same time period.

Fuel debris bed characterisation and isolation
One of the major concerns during the first days after the accident was the possible re-criticalityof the debris bed. Measurements of the ratios of 134I/131I activity were carried out in parallel todumping boron carbide into the remainder of the core material contained in the core cavity inthe reactor. These continuously repeated measurements indicated no formation of short living
iodine isotopes, which confirmed sub-criticality of the debris.
The second urgent task was to identify the mass of the fuel and activities of fission products,
which remained in the debris bed, as well as the properties of the fuel (corium) debris, in orderto predict further accident progression and to provide necessary mitigation measures.
Up to 1988, the dose measurements were taken mainly at the periphery of the reactor. It was
imperative to find methods and means of approaching the places where fuel had accumulated,in order to examine potentially dangerous locations and to collect fuel and aerosol samples.
During the summer of 1986, preliminary results on fuel distribution were obtained based on
the external sample studies showing that more than 96% of reactor fuel remained in the debris
bed, i.e. less than 4% of the fuel released during the accident. Fuel release of 3.5% was reported
to the IAEA (Lewis, 1986). Later, after more complex studies it was concluded that more than 95%,
i.e. more than 180 t of the fuel inventory, was in the debris bed (Borovoy and Velikhov, 2012).

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STATE OF THE ART OF LONG-TERM MANAGEMENT AND ACTIONSStudies of highly contaminated compartments and sampling of surface contamination,
aerosols and fuel debris were started. These remote measurements and samplings were
performed through the existing pipes, e.g. inlet and outlet collectors of control rod coolingsystem, and through boreholes drilled from the accessible and decontaminated rooms. Visual,photographic and video-observations were collected. Radiation and thermal measurements
performed to detect the location and the radioactivity of fuel and to obtain more accurate
information on the condition of the engineering constructions. More than 40 holes were boredand a unique procedure for extraction of highly active core samples was developed. Many
boreholes were drilled and later used for installation of gamma and neutron detectors to obtain
information on corium re-criticality control, as well as measurements obtained from TCs andother sensors (Belyaev, Borovoj and Gagarinskij, 1989). Temperature and radiation
measurements complemented by the detailed sample studies enabled primary detection of thelocations (Figure 2.20) and masses/activities of fuel debris and lavas formed during molten
corium-concrete interaction (MCCI).
Immediately after the accident, the decision was made to construct a confinement building
around the damaged unit to prevent further releases and to provide some shielding against
gamma-radiation. This construction, called the sarcophagus or shelter (Figures 2.18e, 2.18f and
2.18g), used to the maximum extent the remaining structures of the damaged unit. Most part of
the fuel is contained in the sarcophagus in the form of active fragments scattered and deposited
during the accident, as remnants of fuel in-core channels or fuel assemblies in southern spent
fuel pool (SFP), as of finely dispersed fuel particles deposited on surfaces in the sarcophagus andas solidified lava. Some amount of U and Pu is also found in water accumulated in the sarcophagus.
Design and construction of the sarcophagus (Figures 2.18d, 2.18e and 2.21) were completed
extremely fast – in three and in six months respectively. However, the strength of the structure
supporting the sarcophagus was not known at that time and the built confinement was not leaktight. Uncontrolled water penetration/leakages and aerosol releases took place and affected thenear-environment. The safety of the sarcophagus was of prime concern for the long-term
management of the highly radioactive materials on the site of Chernobyl plant. Sarcophagus
failure hazards were increasing with time (Figure 2.18e) and after 30 years, i.e. in November 2016,
the original sarcophagus building was completely covered by a more reliable and leak-tightconfinement, designed for 100 years. This new confinement was designed by an internationalteam of Architects and Nuclear Engineers. With the new confinement (Figure 2.18h), provisions
were set to start the dismantling of the sarcophagus and the recovery of the damaged fuel anddifferent wastes. The safety of these actions should now become a prime objective of the longterm management of the Chernobyl site.

Figure 2.20. Left: location of Chernobyl lavas according to first measurements (in red), top right:
photo of LFCM cluster in room 305/2, bottom right: photo of LFCM covered by concrete

Source: Krasnov et al. (2016).

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Figure 2.21. Section of shelter (sarcophagus) isolating the destroyed unit 4

New concrete
Steel structures
Old concrete
Sand and gravel
Debris in reactor hallDebris in the room 305/2
Corium lavas (LFCM)
Water

1 - pipes of bottom water mains (BWM); 2 - northern additional support; 3 – water tank of containment - scheme «L»; 4 – peripheral rangeof cooling channels; 5 - inclined ferroconcrete slab (a fragment of separator box wall); 6 - metal cladding of heat protection of separator
box; 7 - scheme «OR»; 8 – fuel channels in reactor vault; 9 - ferroconcrete structure; 10 - wall of «loose» FCM; 11 - steam-discharging valve;
12 - LFCM cluster in pressure suppression pool No 2 (PSP-2); 13 - LFCM cluster in pressure suppression pool No 1 (PSP-1); 14 - southern
additional support; 15 – through fusion penetration in sub-reactor slab; 16 – fusion penetration in room 304/3; 17 - steam-discharging
pipe; 18 – main circulating pump; 19 - inclined gallery («Dosifeyev stair»); 20 – tie cladding; 21 - «Octopus» beam; 22 – lead sheets; 23 inclined columns; 24 - western support of «Mammoth» beam; 25 - «Mammoth» beam; 26 - wall on axis 50; 27 – block of B1 beams; 28 –block of B2 beams; 29 - pipe sheathing; 30 – light roofing; 31 – reloading machine (RM); 32 – RM carriage; 33 – RM bridge; 34 – diagnostic
buoy; 35 - upper metal structure of reactor - scheme «E»; 36 - pipes of upper steam-water mains (SWM); 37 - steam-separator; 38 northern buttress wall; 39 – water tank of containment - scheme «D»; 40 - cascade wall.
Source: Krasnov et al. (2016).

Activities associated with the safety of the sarcophagusCharacterisation of fuel-containing materials

Most part of the reactor fuel and fission products is contained in the sarcophagus in thefollowing forms:•

fragments of fuel, fuel rods, assemblies and core structures;•aerosols, hot particles and dusts;•

corium (lavas or lava-type fuel-containing materials [LFCM]);•water solutions and suspensions.

Chernobyl lavas were formed during interactions of molten fuel with reactor structural
materials and concrete (Table 2.1), but also with unknown fractions of materials dumped fromhelicopters on top of the debris during first days of the accident. Their formation and properties
were influenced by the decay heat and the heat of Zr and graphite oxidation reactions. Different
kinds of lavas (see Tables 2.2, 2.3 and Figure 2.22) formed at different temperatures/locations,were sampled and studied in Kluchnikov et al. (2011), Arutyunyan et al. (2010), Borovoy and
Velikhov (2012), and Krasnov and Khan (2015).
Note that sampling of lavas was done several times manually (!) during the first days andlater using robots, but majority of samples were obtained through boreholes drilled fromdecontaminated rooms between 1988 and 1991 (Figure 2.23). Diameter of boreholes was in the

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range of 60 to 150 mm and the lengths were up to 26 m. Main fraction of the drillings was done
from the compartments located in the west part and a smaller fraction from the south to thenorth. Some of the boreholes were drilled with an inclination. The boreholes were used for:•estimation of internal damage of reactor and equipment;•

determination of the main locations of lavas and material sampling (Figure 2.24);•

installation of different instrumentations, including cameras, temperature sensors, γand neutron detectors.

Later, installed instruments were used to develop a complex radiation and safety
monitoring systems, such as “finish” (Borovoy and Velikhov, 2012) and others.

Table 2.1. Estimated masses of materials within the reactor
location and masses involved in Chernobyl lavas
Masses within the reactor location*
after the accident, tonnes

Material

Masses of the materials
in lavas, tonnes

Fuel (U)

120

90

Steel

1 300**< 20***

Serpentinite

580

160

Concrete of under-reactor plate
Concrete of reactor building

Minor
950

130
480

300

280

Sand of lateral reactor shieldZr

No data

45

750

Minor

Graphite

Note: * In the compartments #504/2 (reactor pit) and #305/2 (under reactor pit); ** Non-melted materials are excluded;*** 330 tonnes of steel melted and spread separately of lavas.
Source: Adapted from Borovoy and Velikhov (2012).

Table 2.2. Masses of fuel and corium (LFCM) in different shelter compartments
Compartment (number)Central hall (914/2)
Southern spent fuel pool(505/3)All top rooms, including
central hall (CH)(mark +24.000 and above)
304/3
301/5+301/6+303/3
217/2
Sub-reactor 305/2 and 504/2(to +24.0)
SDC (210/5+210/6+210/7)
PSP2 –(012/14+012/15+012/16)
PSP1 – (012/5+012/6+012/7)
Fuel under cascade wall
Water in all rooms
Fuel on “shelter” object site

Fuel-containing materials
type

Mass of fuel estimatedfor 06.09.2010, t (U)Comments

Fuel, LFCM

More than 22

With consideration of 48 fresh fuelassemblies (5,5 t)

129 spent fuel assemblies

14.8

LFCM
LFCM
LFCM

∼5 on blockage surface
in CH,∼totaly 30
6±2
5 ± 2.5
0.4 ± 0.2

Fuel, LFCM, dust

80 ± 30LFCM

12 ± 6

LFCM

3-12

LFCM
Fuel, dust
Soluble uranium salts, slurry
Fuel, dust, fragments

1.0 ± 0.5
(0.9 ± 0.3)
0.004
0.75 ± 0.25

Fuel and dust

LFCM may be present
Estimated 30 t include surface
contamination inside blockage in
CH and all other rooms«Horizontal lava flow»«Horizontal lava flow»«Elephant foot», «stalactites»
Estimated for 7 FCM clusters.
Start of all LFCM flows.
FCM in a break through between
rooms 304/3 and 305/2«Large vertical flow» and «small
vertical flow»

Note: See compartment and room locations in Figures 2.23 and 2.24.
Source: Klyuchnikov et al. (2011).

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Figure 2.22. Different types of corium (fuel-containing materials)
detected in the Chernobyl shelter
Sampling place

Black ceramic

Sampling place

Brown ceramic

Sampling place

Polychromatic Ceramic

Sampling place

Porous Ceramic

Source: Krasnov et al, (2016).

Chemical (Table 2.3) and phase compositions of lavas are very complex but majority of thelavas are alumino-silicate glass-ceramics with abundant inclusions and pores. Differences
between black and brown ceramics in uranium inventory are visible in Table 2.3. Macro- andmicrostructures of lava are not uniform, see examples in Figure 2.26, and typically represent
large inventory of phases. Main part of the reactor fuel formed (U,Zr)O2-x or (Zr,U)O2-x type solid
solutions with tetragonal and monoclinic structures, UO2, U-rich zircon and a smaller part of U
is dissolved in the other phases.

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Table 2.3. Averaged oxide content of Chernobyl lavas
LFCM typeBlack LFCM 304/3
Black LFCM 217/2
Black LFCM 210/6
Brown LFCM 210/7

Main oxides, mass. %
SiO2
70.6
66.6
62.1
64.0

Al2O3
7.4
8.7
7.2
6.8

Fe2O3
0.25
0.40
2.91
0.64

FeO
0.23
0.36
2.63
0.57

MgO CaO
3.9
6.7
3.8
8.5
5.1
6.0
7.0
6.7

Na2O
6.2
5.6
5.2
5.4

TiO2
0.21
0.27
0.19
0.24

ZrO2
5.8
5.8
5.5
6.6

BaO
0.13
0.15
0.18
0.19

UO2
4.3
5.0
5.8
9.4

MnO
1.9
3.8
0.40
0.53

Cr2O3
0.30
0.33
0.40
0.39

NiO
1.2 ∙ 10-3
0.19
0.39
0.36

Source: Klyuchnikov (2015).

Table 2.4. Masses of dusts in different compartments of shelter
Place (room)

Square, m2

Volume,m3

Dust amount, t of U

Uncertainty, t

2005/2
2014/3
2014/41001/21001/41002/2
1003/31003/41004/21005/21011/11011/2
902/2
903/2
904/3
2005/2
2014/3
904/4
905/3
905/4
906/2
907/2
909/9-10
909/11-12
909/13-16
910/2
911/2
912/2
915/3
916/3
916/4
918/2
919/5-6
919/5-8
926/3
926/4
935/1
935/2
804/3
804/4
612/2
505/4
219/2
061/2
305/2,504/2,404/3, 404/4

100
27
27
54
54
35
105
105
330
35
21
21
17
39
23
100
27
23
30
30
28
32
26
26
72
205
12
81
210
14
14
17
50
50
31
31
4.5
4.5
470
470
80
40
63
45
740
Total

250
140
140
150
145
95
290
290
900
300
70
70
15
100
70
250
140
30
110
110
90
95
70
80
150
400
20
230
700
35
35
45
140
140
130
130
20
20
7 000
6 000
900
500
1 850
1 000
5 300

0.080
0.022
0.022
0.043
0.043
0.028
0.084
0.084
0.260
0.028
0.017
0.017
0.014
0.031
0.018
0.080
0.022
0.018
0.024
0.024
0.022
0.026
0.021
0.021
0.058
0.160
0.010
0.065
0.170
0.011
0.011
0.014
0.040
0.040
0.025
0.025
0.004
0.004
0.380
0.380
0.064
0.032
0.050
0.036
0.590
3.115

0,010
0,003
0,003
0,005
0,005
0,004
0,011
0,011
0,033
0,004
0,002
0,002
0,002
0,004
0,002
0,010
0,003
0,002
0,003
0,003
0,003
0,003
0,003
0,003
0,007
0,021
0,001
0,008
0,021
0,001
0,001
0,002
0,005
0,005
0,003
0,003
0,000
0,000
0,047
0,047
0,008
0,004
0,006
0,005
0,074
0.391

Note: see locations of some compartments and rooms in Figures 2.23 and 2.24; example of isotope specific activities of dust given in Table 2.5.
Source: Krasnov et al. (2016).

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Table 2.5. Average radionuclides contents in fuel particles
90

NuclideActivity, Bq/g (U) (01.01.2015)
NuclideSr

90

5.97 · 108

5.97 · 108

239

Activity, Bq/g (U) (01.01.2015)Pu

5.00 · 106

Y

240

Pu

8.18 · 106

137

Cs

7.23 · 109
241

Am

2.41 · 107

238

Pu

6.19 · 106
241

Pu

2.41 · 108

Source: Krasnov et al. (2016).

Figure 2.23. Scheme of horizontal and inclined boreholes drilled
for sampling and installation of instrumentations

Source: Krasnov et al. (2016).

Figure 2.24. Scheme of melt flows and locations of lavas in different compartments(a) Black ceramic in the room #304; (b) “Elephant foot” in the room 217/2; (c) Brown and black ceramics in steam
distribution corridor 210/6; (d) Lavas with high U content in the room 305/2; (e) Brown ceramics in the pressure
suppression pool #2; (f) Brown ceramics in the pressure suppression pool #2.
Source: Krasnov et al. (2016).

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Figure 2.25. Typical views of samples retrieved through the boreholes

DebrisFragments

Source: Borovoy et al. (2012).

U-rich zircon was attributed as a new technogenic mineral: Chernobylite (Figure 2.26 right),
i.e. a crystalline zirconium silicate (Anderson et al., 1993a) (Zr1-xUx)SiO4, which contains
6-12 mass % of U. Natural crystalline zircon with such a high U content is unknown. Possible
formation mechanism is suggested as isomorphous inclusion of U into the mineral structure
after replacing Zr. A study (Shiryaev et al., 2016) concluded that at least some zircon crystals
rotated during last stages of the lava solidification, indicating that they were formed attemperatures significantly higher than solidus temperatures of lavas.
Microstructure and other properties of the Chernobyl lavas were extensively studied, inparticular, by Anderson et al. (1992, 1993b), Borovoy et al. (1990), Pazukhin (1994, 2002, 2006,2008), and Pazukhin et al. (2002), but some important features related to the prediction of theirlong-term behaviour remain poorly studied.

Figure 2.26. Examples of (Left) Brown Ceramic, (Centre) Black Ceramicand (Right) Chernobylite microstructures (scale given in µm)

10 µm
1. Uranium oxide (white); 2. Inclusions of(Zr, U)Ox (grey and light brown); 3.
Globules of steel.
Source: Arutyunyan (2010).



10 µm
20 µm
Large uranium oxide inclusions (white) andsmall globules of steel (grey)..

BSE image of several inter-grown skeletal
zircon crystals with abundant urania
inclusions. Note that some UO2 dendrites
cut zircons growth zones.

Criticality control of fuel and corium in the shelter (Sarcophagus)

The first attempts of measurements of the multiplication factor in the shelter were made in1991 (Lebedev and Shikalov, 1995). The measurements were performed using neutron pulse
method, which is one of the active methods, along with the standard geophysical equipment of
neutron logging of the bores. Unfortunately, it was not possible to vary the frequency of theneutron pulse emission and the time of registration of the system response was limited to 2 ms.
Consequently, the delayed neutrons were not measured and the precision of the multiplication
factor was not sufficient.

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New monitoring systems, such as “finish”, “pilot” and “KSFCM” were constructed in the
shelter using neutron, γ and temperature detectors which helped to obtain the measurements ofneutron flux density in many locations. Data accumulated before 1990 confirmed conclusions,reached earlier, on the technical justification of the criticality safety of shelter (Belyaev et al.,1990). The Keff measured by active methods was below 0.7 and the values measured by passive
methods were below 0.4, which confirmed that at that time the risk of re-criticality was
insignificant. However, it was indicated in this and in the following study (Borovoy et al., 1996a)that the barriers which prevented water penetration into lava after the accident (elevatedtemperatures, glassy-like lava surface, boron and gadolinium added solutions) were becoming
less efficient with time: lava temperatures had decreased, glass-like structures were slowly
disintegrating, making them water-permeable, boron and gadolinium concentrations in water
were decreasing in time.
Several incidents in 1990 and 1996, which may indicate re-criticality in the shelter, werereported (Borovoy and Velikhov, 2012). The first one happened after the rainy summer of 1990.
The readings of neutron detector installed on the surface of lava in the room 304/3 (Figure 2.23)
through the borehole increased approx. 60 times (Figure 2.27). In the context of detector shift
from its original position as one of the possible reasons of the counting rate increase, thedetector position was visually inspected with a periscope at the beginning of the count rateincrease: on 27.09.90 (before the time window shown in Figure 2.27). No changes were detectedin comparison to the previous view. On the same day the detector and the system channel werechecked with 252Cf neutron source (~1⋅107 neutron/s) temporarily installed in the room.
Response of the measurement system was according to the expectations and there appeared to
be no problems with the measurements.
After the readings reached 160 count/s, gadolinium nitrate solution was injected in two
portions of 80 l each into the west and east parts of the room. First portion reduced the readings
rapidly to 30 count/s. Slow further reduction to 2.4 count/s was observed after the 2nd portion
was injected during the day.

Figure 2.27a. Locations of sensors of “Finish-R” and “Signal” systems
for monitoring nuclear safety of fuel-containing materials

Source: Krasnov et al. (2016).

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Figure 2.27b. Neutron detector readings (count/s) in the room # 304/3
from the morning of 29.06.90 to the evening of 30.06.90

Source: Krasnov et al. (2016).

Three times smaller increases of counting rate occurred on 12 and 16.09.1996. They
happened after the rainy weather also. Studies performed after these data were gathered andshowed that the neutron count rate correlates with the water ingress into the shelter.
Many calculations and analyses of the incidents were performed with reactor and Monte
Carlo codes, but there are still different opinions about the nature of the incidents. Authors(Frolov, 1996) concluded that water penetration into corium debris zone enriched with fuelresulted in reactivity increase with a rate of 10–4 β/centigrade and that re-criticality of this zone
was due to the delayed neutrons at Keff ≈ 1.000009. Alternative reason of reported neutron
anomaly can be connected with possible increase of neutron detector efficiency/sensitivity
caused by softening of neutron spectrum due to water moderation, i.e. with the inaccuracies ofthe measurements.

Studies of debris and lava ageing

Lava ageing can potentially increase the risk of fission product release due to leaching and lava
dispersion. Several phenomena can be responsible for Chernobyl lava ageing, such as matrix
leaching by water and material interactions with atmospheric gases/vapours, self-irradiation,
thermomechanical cracking and fragmentation.
Intensive studies of long-term ageing of natural lavas having chemical and phasecomposition similar to Chernobyl lavas, which can be classified as alumino-silicate glassceramics with abundant inclusions and pores, of borosilicate glasses and 30-year-old lava
samples and aerosols formed in different compartments provided new information reviewed
by Shiryaev et al. (2016), which is useful for assessment of long-term behaviour of the lavas.

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The closest to Chernobyl lavas in silica inventory are natural lavas from Mt. Cameroon
volcano in Africa, which is very high and has a wide range of variation of atmospheric conditions,such as humidity and temperature. According to the long-term observations, the characteristic
thickness of the alteration crust at the lower part of the volcanic mountain does not exceed
10 cm in 100 years, but tephra and pumice can be altered at much higher rates, presumably due
to much higher surface area. These observations confirm importance of leaching and materialinteractions with atmospheric gases/vapours.
Investigations of various types of nuclear glasses have shown significant effects on glass
performance at accumulated doses exceeding ~1018 α-decays/g and/or ~109 Gy. In the case ofChernobyl lava, assuming a homogeneous distribution of α-emitters, such doses can beaccumulated in ~104 years. If lava structures are analogous to nuclear glasses self-irradiation
would not be an extremely severe problem for LFCMs (Pazukhin et al., 2002). However, macroand micro-structural inhomogeneities, e.g. concentration of α-emitters in aggregates andinclusions, and variety of properties of lavas in different locations are detected. Also, LFCM
structure is different and can be less stable in comparison with the accurately synthesised
amorphous borosilicate glasses. Thus, sensitivity of lavas to the internal radiation can be muchhigher. Nevertheless, the direct SEM/EDX studies of several samples collected long ago, stored
in dry conditions and recently examined by Pazukhin et al. (2002), did not reveal catastrophic
cracking, fragmentation and powdering. Only some cracking is reported around the large
inclusions of UO2 and Zr-U-O phase.
Further, temperature gradients in the lavas can reach several tens of K and can be different
at different locations. Such thermal gradients can result in limited thermal stresses. Thermal
cycling of the lava samples in laboratory fatigue tests demonstrated remarkable stability of thebulk material even in the presence of solid inclusions. However, high thermal stresses may haveoccurred during solidification of lavas. These thermal stresses can increase the risk ofmechanical failure and fragmentation.
Another possible mechanism of ceramic fragmentation, which is very typical for porous
natural rocks of Nordic countries, can be associated with water penetration into open micro
and macroporosity and subsequent freezing of water at low temperatures. Such phenomena
are limited by the decay heat of the lavas having noticeable content of fuel and fission products.
It was found in the experiments (ISTC, 2000) with LFCM thermal cycling in the range of(-20)-(+50)°С and parallel microscopy of micro-cracks that such extreme conditions do notcause significant propagation of the micro-cracks. It was explained by the water molecularinteraction with silicate clusters at the micro-level with formation of bounded water whichcannot form ice. However, as it was confirmed by visual monitoring of lavas in the shelter,
water freezing can cause significant damage of macrostructure of the lava when water
penetrates into the large pores.
According to Borovoy and Velikhov (2013), the main phenomena responsible for the lava
destruction and fragmentation/powdering are leaching and material interactions withatmospheric and technogenic gases/vapours. The most affected and fragmented currently areblack and porous ceramics (Figure 2.22). Note that the black ceramic has the highest contentsof fuel and fission products. The most resistant are brown and polychromatic ceramics(Figure 2.22). The damage of different ceramics is controlled not only by the ceramic properties,
but specific location and configuration of lavas allowing or preventing continuous interactionof the ceramic with water. Nevertheless, spontaneous fragmentation of several specimens
collected in the shelter and kept in storage without access of water was observed after 20 yearsof storage in Kurchatov and Radium Institutes.
Leaching rates of different nuclides from Chernobyl lavas by a water solution having
compositions typical for the shelter differ for various lavas as shown in Figure 2.28 (ISP NPP,
2007).

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Figure 2.28. Leaching rates of different radionuclides from polychromic
ceramic, brown ceramic and black ceramic

Note: The damages to the surface layer of lavas in the shelter can be seen visually in some cases,
e.g. presenting dark green or yellow colour surface structures. An example is shown in Figure 2.29. It
contains the following phases: (UO2) CO – Rutherfordine; UO3 • 4H2O – Epijanthinite; UO3 • 16CO2 •1.91H2O; UO4 • 4H2O; Na4 (UO2) (CO3)3; Na4UO2 (CO3)3 • 3Na2CO3.
Source: Krasnov et al. (2016).

Figure 2.29. New formations on LFCM surfaces: (top left) in steam releasecorridor (top right) in steam discharging valve (bottom)black ceramics

Source: Krasnov et al (2016).

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It may be concluded that there are no doubts about ageing, degradation of properties andspontaneous fragmentation of lavas but the degradation rate and related radiological hazards
remain uncertain. Consequently, at the current level of knowledge these phenomena cannot be
quantified or even well predicted. As it is shown by Shiryaev et al. (2016), spontaneous
fragmentation and powdering of lavas create secondary releases of fission products, both withaerosol particles and water solutions/suspensions. Ageing issues should be addressed very
seriously in the safety analyses and management of the unit 4 of Chernobyl NPP.

Spent fuel pools

There are two spent fuel pools in the 4th unit of Chernobyl: the North and the south pools. The
North SFP was empty by the time of the accident; the south pool contained 129 spent fuelassemblies corresponding to 14.8 t of uranium. The south spent fuel pool has a size of 10.6 (length)
x 4.2 (width) x 18 m (height). All fuel assemblies were contained in cylindrical casks.
Five boreholes were drilled in the SFP walls to check in 1990 the status of fuel assemblies
located in the pool. The periscope observations showed that the pool was dry and that its lid wasdestroyed, but materials from the reactor hall did not relocate into the pool. The casks were attheir original positions but debris was visible at the bottom of the pool through the gaps between
the casks. Dose rate close to this debris was at the level of 5 000 R/hr (13.89 mSv/s). The absorbers
installed in the pool for criticality control during normal operation were also visible.
Dose measurements through the boreholes indicated that the dose rate in the upper part of
the SFP decreased from 3 500 R/h (9.72 mSv/s) in 1990 to 370 R/h (1.03 mSv/s) in 2013 but theneutron flux in the lower part increased from 140 to 400 n/(sm2s) during the same time. Thissuggests possible fuel relocation to the pool bottom, e.g. due to fuel overheat and degradationin dry conditions.



Water management in the sarcophagus

Water management in the sarcophagus is considered critical since water can:•fragment fuel-containing material increasing the amount of mobile radioactive particles
inside the sarcophagus;•

cause an increase of the criticality of fuel-containing material in the course of time, as
their cooling and fragmentation proceeds (with generation of nuclear hazardous
compositions);•

contribute to the weakening of building structures;•

lead to groundwater contamination with radionuclides.

The water balance has not been understood completely but its most important
contributions have been identified.
Before implementation of the new confinement in 2016 (Figure 2.18h), the following sourcesof water in the shelter were found (with a total flow rate of approximately 1 600-2 000 t/year
before 2004):

54•

Leakages of atmospheric depositions (rain and snow) through the roof and walls.
Between the years 2004 and 2008 extensive work on shelter reconstruction wascompleted, including hydro isolation of the building and renovation of its roof. More than
600 m2 of roof were replaced. These measures practically eliminated rain and snow
penetration into the shelter.

•

Condensation of humidity of air inside the shelter, which contributes with approximately
600 t/year but with visible variations between the years because of the weather differences.

•

Operation of the system for dust suppression employing spray nozzles under the roof,
which supply 300-400 t/year before 2004 and approximately 100 t/year after it.

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The contaminated water releases from the shelter are due to:•

Leakages into the sump water system of the unit 3, as detected by the trace method withsodium bromide. This water, which is classified as liquid wastes, is accumulated in thesump tank and periodically pumped for reprocessing. The flow rates are between
580 and 1 090 t/year (Krasnov and Khan, 2015).

•

Leakages of water into the soil and mixing with groundwater. The flow rate is unknown.
Systematic monitoring of groundwater contamination at the site has been conductedsince the year 1992. Currently, approximately 50 boreholes are integrated in the survey
grid. In the boreholes, the groundwater levels are measured, gamma-spectrometry
logging is conducted in monitoring mode, and groundwater sampling is made.

•Evaporation and release into atmosphere with air (~380 м3/year). The source term
associated with the water evaporation is unknown but is estimated to be quite low.

The water volumes accumulated in the shelter also differ from year to year. In 1996, 3 000 m3
of water was reported in the shelter, in 2007 – 331 m3, in 2008 – 322 m3, in 2009 – 318 m3 and in2010 – 337 m3. The water volume has remained nearly constant in the last few years.
The water is mainly contaminated with Cs isotopes, 90Sr, U and Pu (Table 2.6). Significant
increase of Pu concentration in water, but also U concentration in water sampled from several
locations, is visible from the comparison of data for the years 2000 and 2010. It confirms
disintegration and fragmentation of lavas in the shelter. Data reflecting increasing trend of Pu
and U leaching are marked in the Table 2.6 with the grey background.

Table 2.6. Average yearly data on contamination of water in the shelter
Sampleplace #
620
30
32Isotope specific activity, Bq/m3
137

90

Cs

Sr

Integral Pu

Concentration of U,
mg/l

2000

2010

2000

2010

2000

2010

2000

2010

6.2 ·1010
1.2 ·1010
5.2 ·109
1.3 ·1011

3.6· 1010
3.9 1010
8.1 ·109
3.9 1010

1.0 1010
4.2· 1085.4 ·109
1.1 1010
1.7 109
5.5 1010

4.0 106
2.7 1053.6· 105
4.2 106

6.8 106
1.3 107
2.7 106
1.6 107

48
1.7
2.7
109

31
42
8.1
34

2.2 1010

Source: Krasnov et al. (2016).

Radionuclide releases with water can be associated with dissolved compounds andsuspensions of solid particles. Content of solid particles is at the level of (0.02 ± 0.6) g/l. The
dynamics of groundwater contamination by Cs and Sr measured in a borehole close to theshelter is shown in Figure 2.30. There is a clear tendency towards the reduction of Cs and Srconcentrations in groundwater.

Figure 2.30. Cs and Sr specific activities in groundwater (Bq/m3) near the shelter

Source: Krasnov et al. (2016)

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STATE OF THE ART OF LONG-TERM MANAGEMENT AND ACTIONSCountermeasures related to the sarcophagus radiation safety
Surveillance means were implemented for radioactive aerosol releases out of the sarcophagus
and for water inside the sarcophagus before implementation of the new confinement.
A dust-suppressing installation using water solutions sprays was implemented in the central
hall to reduce radioactive aerosols concentrations. This implementation was very effective in
reducing airborne aerosols concentrations but water was sprayed and added inside thesarcophagus. Many other technical measures reported earlier enabled significant suppression of
the aerosol source term (Figure 2.31) for the time before the new confinement installation.

Figure 2.31. Dynamics of daily (a) and annual (b) releases of β-active
aerosols with air leakages from 1998 to 2015

Source: Krasnov et al. (2016).

Tritium specific activity in the groundwater at the site during 1997-2000 increased from 2 to
5 000 Bq/l. Spatial distribution of tritium in the groundwater and other features, such as similar
distributions of pH index, concentrations of sulphates, chlorides, phosphates and nitrates
indicate leakage of contaminated water from the unit 4 into the groundwater (Shestopalov etal., 2000) and soluble radionuclide releases with groundwater into the geological environment.
Very limited quantitative data about tritium source term can be found in literature.
The new confinement moved on place in November 2016 has the objective to control
radioactive releases of aerosols and contaminated water, even if the original sarcophagus
collapses and re-criticality events occur.

Long-term management of damaged fuel and wastes

Plans for fuel and lavas recovery and removal from the shelter have not been finalised yet.
According to the most optimistic prognosis these activities will start after 2030. It is very
important to remove debris and waste during the lifetime of the new confinement.

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The amounts of fuel and lavas to be removed are shown in Table 2.2 but the masses of highly
contaminated solid radioactive wastes, including the wastes of units 1-3 are much higher.
It is not clear, at the moment, if the removed corium and fuel will be reprocessed and how
the final disposal will be realised.

Long-term water and solid waste management

Effective waste management in Ukraine would need the development of a national programme
for radioactive waste disposal, covering: (a) facilities for reprocessing and temporary storage,and (b) building a final repository for the high-level wastes. Chernobyl units 1 to 4 are not the
only sources of the wastes. Presently in Ukraine there are:•

2 917 960 m3 of wastes accumulated at the temporary storages;•

3 720 m3 of solid wastes and approximately 800 m3 of liquid wastes at 800 places inside
the 30 km zone around the plant.

The presently available technology of liquid waste reprocessing is based on water
evaporation. This technology cannot be used for contaminated water from the shelter and the
3rd unit because of very high concentrations of Pu and other transuranium elements but alsoorganic substances, such as oil and its products, surfactants and film forming compounds used
for suppression of dust, aerosols and fission product release from corium and fuel debris. Thereis clearly a large space for development of new decontamination and reprocessing technologies
and new practices for the cleaning of the NPP and the 30 km zone and finally its rehabilitation.
Comprehensive recommendations on management of radioactive wastes produced by the
Chernobyl accident but also state-of-the-art review of dedicated techniques and experiences
are formulated by NEA Radioactive Waste Management Committee (RWMC) and its Expert
Group on Fukushima Waste Management and Decommissioning (EGFWMD) established in 2014.
The work covers physical and chemical nature of the wastes, radiological characterisation,
waste classification/categorisation, conditioning, reprocessing and disposal.
According to the recently published NEA report (NEA, 2016):… management of FCM is identified as the most serious problem in the
shelter object, as these materials are in an uncontrolled state withassociated nuclear and radiation risks. FCM are the main source ofenvironmental releases of radionuclides from the shelter object. FCM werestudied carefully during the first years after the accident; after that there
were no detailed investigations of their physical and chemical
characteristics because of high dose exposure rates and costs. Physicaldegradation of some of these lava-like FCM is of concern because of thepotential to produce radioactive aerosols, which would be highly mobile.
The Ukrainian programme recognises the need to implement permanent
neutron activity and temperature monitoring of FCM, to assess thelikelihood (albeit low) of criticality, as well as monitoring of physical andchemical characteristics to study the dynamics of FCM destruction
processes. In addition, the need to retrieve FCM during the lifetime of the
new safe confinement, and process them to waste forms suitable for longterm management, has been recognised.
The report concludes on waste storage/disposal that: Waste generated on the site ofChernobyl NPP resulted from the operational activity of the ChNPP units and waste resulting
from clean-up after the accident, and mainly from decontamination activity. These wastes are
stored in existing (“old”) temporary storage facilities for solid and liquid waste. A set of new
facilities were constructed on the ChNPP site to start retrieval, characterisation, sorting,treatment and conditioning of stored waste. As a final product, it is expected to produce drums
and containers with solidified (cemented) waste to be acceptable for disposal in the near-surface
disposal facility.

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The first experience of licensing such a facility – the Engineered Near-surface Disposal
Facility (ENSDF), – has highlighted the lack of data on radiological characteristics of conditioned
waste, because of a lack of information about waste to be retrieved from the “old” storage
facilities. This has meant that cautious assumptions are made about the waste to be disposedof at the facility, which in turn means that the facility may not be used to its full environmental
capacity. It is recognised that operators of treatment plants should now provide betterradiological characterisation of the waste forms being produced.
Moreover, the requirements of the Ukrainian legislation for radioactive waste disposal havebeen found to be very conservative for some parameters compared with IAEA Safety Standards.
One lesson learnt is that Ukrainian legislation should be updated to take account of IAEA Safety
Standards, at the same time as radioactive waste classification is updated and general
acceptance criteria for future disposal facilities are developed.
Finally, the benefits from international co-operation for different decommissioning activities
are discussed in the report.

Summary of long-term management of the Chernobyl accident
Leaving aside the causes that lead to the Chernobyl-4 accident and the possible lessons learnt
from the early phase of the accident for reactor safety and Bolchoi Mochnosti Kanalnyi (highpower channel-type reactor) reactor design improvements, the following may be discussed in
the context of the long-term management of this accident.
It should be noted initially that the time to reach the stabilised controlled state of the unit 4,
as defined in Section 3.1, was very long. The reactor was in an “extreme-damage” state, since thereactor core and building were almost completely destroyed. Although the source term and theradiation levels were reduced significantly after about ten days when the core graphite fire wasextinguished, there was still considerable release of radioactive aerosols and very high dose ratefrom local contaminations at the site. It took six months to construct the sarcophagus (or shelter)
which isolated radioactive materials and equipment from the environment. This duration of six
months for the shelter construction project is in fact extremely short, considering the radiationlevels, destroyed systems and buildings, the size of the shelter and the functional requirements
for the shelter. Shelter construction was extremely costly in terms of doses for “liquidators” – thestaff that was sent to do the accident clean-up.
Several complicated, expensive and urgent technical measures, such as dose rate mapping,primary decontamination of turbine hall roof and of the plant site in the vicinity of the otherunits, isolation of destroyed unit 4 from unit 3, allowed restart of the first two units and restartof unit 3 later – in November of 1987. Unit 2 was shut down in 1991, unit 1 in 1996 and unit 3 in
2000. The decommissioning of units 1-3 is being carried out separately from that of unit 4.
Much more time was necessary to diagnose locations of lavas containing the melted fueland establish control of temperatures and criticality of the debris as well as to suppress
formation of dusts at its surfaces. Status of spent fuel was checked only in 1990, i.e. almost four
years after the accident. However, the possible damage to the spent fuel and prognosis of its
behaviour in dry conditions are still not evident.
The on-site decontamination and protection of rivers and lakes also took several years.
However, some uncontrollable leakages of contaminated water from the shelter anduncontrollable releases into the hydrosphere can have happened. Further, solid wastes
produced during decontamination and equipment contaminated during these operations wereplaced in temporary surface storages of trench type.
The liquidators worked in extreme conditions and many decisions with respect to thenecessary prompt measures to stabilise the damaged reactor unit were made with no chance of
having the necessary measurements, diagnostics and modelling support in advance. Even now,
after 30 years, there are continuous discussions about the effectiveness of different measures
undertaken directly after the accident, such us building a concrete slab below the reactor,
dropping different materials from helicopters to stop the fire, to form a filtration pad on the topof the destroyed reactor, pouring concrete from the top, building protective wall along the riverand several other measures, which were taken at that time.

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A relatively stable state of the debris in the shelter was reached in the early 1990s but withthe very high price of the high doses to the liquidators. Following that, various long-term
management actions contributed to maintain this stable state for approximately 20 years inparallel to the operation of other units on the site and following their shutdown. The mainefforts undertaken for the long-term management at the site were as follows:•

Monitoring of radiation inside the shelter and at the site including measurements ofradioactive releases from the shelter in the form of radioactive aerosols and contaminatedwater.

•

Detection of uncontrolled air and water leakages and sealing of the shelter building.

•

Reduction of radioactive dust formation by sprays of solutions forming a protective
polymer film on the surfaces of lavas and internal shelter structures/equipment.

•

Keeping optimal humidity inside the shelter to minimise both dust formation andradioactive water accumulation.

•

Diagnostics and monitoring of corium containing lavas at their locations. Control ofdebris temperature and reduction of criticality risk using spray system operated withwater but also with boron and gadolinium solutions.

•

Further on-site decontamination and building of radiation protective structures for
operating team as well as radioactive waste storage facilities.

•

Extensive research programmes necessary to understand accident progression, inparticular temperatures reached in the course of the accident, ex-vessel corium behaviourand MCCI, to complete database representing corium locations, compositions andproperties in order to support long-term management and engineering at the site but also
to develop prognosis about long-term evolution of Chernobyl lavas, environmental
releases, accumulation of radioactive wastes and corresponding risks.

•

Monitoring of mechanical and structural stability of various structures and components.
Three main parts were detected in 1989 as having critical deformations and mechanicalruptures: i) in the upper zone of deaerator support stack, ii) Ceiling of south room of main
recirculation pumps (#402/3) and iii) Ceiling of south room of ventilation pipes (#805/3).
These structures were renovated and strengthened in priority. Many other structures wereserviced, renovated with new reinforcement or replacement later on, e.g. main ventilation
pipe (1990), south and west walls of shelter, its roof (2000-2008), and other structures later
on. In spite of many completed renovation and reinforcement measures, it has beenconcluded that the sarcophagus is mechanically unstable and not leak tight. Decision wasmade to build a new confinement around the old one.

•

Design and construction of the new shelter with a lifetime of 100 years. It has been
completed and installed in late 2016.

The listed efforts are not sufficient for completion of long-term management of Chernobyl
accident but they provide the starting conditions for the future phase of spent fuel and corium
debris removal and reprocessing. The detailed plan, which is expected to be completed within
the life time of the new confinement, and technical measures for debris removal are currently
in preparation.
The following studies of the Institute for Safety Problems of NPP (ISPNPP, Ukraine) can belisted among the currently ongoing research which can further contribute to the long-term
management of Chernobyl NPP:•

complex assessment of cumulative environmental impacts of radiation-hazardous
objects of Chernobyl evacuation zone:–

development of measures to minimise the risks of re-criticality incidents;–

study of radioactive aerosol behaviour during commissioning and operation of the
new confinement;

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research and development to minimise the risks associated with contaminatedwaters in the new confinement;–

statistical analysis of radiation monitoring results and ecological studies of the new
confinement and the plant site;–

development of system design for monitoring of new confinement internal spaceand data acquisition system.

•

study of Chernobyl lava in the conditions of new confinement and development ofmethods and technologies of lava reprocessing;•uncontrolled releases from shelter and aerosol behaviour during new confinement
construction;•assessment of evolution in the evaporation-condensation dynamics of moisture in the
shelter covered by the new confinement, behaviour of Chernobyl lavas and dust
formation;•

radio-hydrogeological monitoring in the site;•

development of characterisation methods for solid radioactive wastes and packages ofwaste reprocessing plant;•

assessment of protective polymer coating behaviour in the under-roof space of the
shelter.

Experimental study of fission product leaching from Chernobyl lava samples by water will
be carried out at Khlopin Radium Institute (KRI), Russia in 2018-2019. This study is supported in
the international call of the NEA TCOFF project on Thermodynamic Characterisation of Fuel
Debris and Fission Products based on Scenario Analysis of Severe Accident Progression at
Fukushima Daiichi.

Synthesis of relevant Fukushima Daiichi experience
Generalities on long-term management and recovery actions at Fukushima Daiichi
The safe long-term management (LTM) at the Fukushima Daiichi NPP is a highly challenging
task that requires the allocation of enormous resources as well as the development and use ofinnovative technologies to deal with very difficult activities.
After the accident, reaching a stabilised controlled state (sub-criticality and cooling of thedamaged fuel and limitation of radioactive releases) for the damaged 1, 2 and 3 units and unit 4
SFP was the priority. This was particularly challenging at Fukushima Daiichi because of thedamages to the infrastructures due to the seisms, the tsunami, the hydrogen explosions andthe impact of fires and because of the large on-site contamination that resulted from the loss ofconfinement in units 1 through 3. Contamination spread was also aggravated by hydrogen
explosions and fires. First actions after the accident concerned the re-establishment of groundtransports, clean-up of the site to enable access and implementation of systems, equipment
and instrumentation for securing and monitoring a stabilised controlled state and to reduceradiological dose rate.
On 17 April 2011, Tokyo Electric Power Company (TEPCO) established the “Roadmap towards
Settlement of the Accident at Fukushima Daiichi Nuclear Power Station, TEPCO”. This Roadmap
set two steps for stabilisation of the plants as targets: “Radiation dose is in steady decline” as“Step 1” and “Release of radioactive materials is under control and radiation dose is being
significantly held down” as “Step 2”.

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Step 1 was completed on 15 July 2011, and step 2 was completed on 16 December 2011. Bycompletion of step 2, the following were achieved and the plants were declared as in a “coldshutdown state” by the government (NDF, 2016):•

the reactor pressure vessel (RPV) bottom temperatures and temperatures inside theprimary containment vessel (PCV) are kept below approximately 100 degrees centigrade;•the steam generation from the PCVs is suppressed via controlling water injection which
also suppresses the release of radioactive materials with steam from the PCVs;•

the mid-term safety of the circulating water cooling system is secured. 1

What also made and is still making the situation challenging at Fukushima Daiichi is thatthree units have been damaged with different accident short-term evolutions and accident
management actions resulting in different unit damaged states.
It should be emphasised that LTM management actions, while aiming at maintaining andmonitoring a stabilised controlled state of the degraded fuel in the damaged units, are now
largely addressing radioactive waste management and the preparation of damaged fuelretrieval issues. The radioactive liquid wastes management is a critical issue at Fukushima
Daiichi notably due to the groundwater intake in the damaged facilities and the large volumesof contaminated water produced. Details of radioactive waste are summarised in Management ofRadioactive Waste after a Nuclear Power Plant Accident (NEA, 2016). Damaged fuel retrieval is a
critical step towards safe decommissioning.
The situation on-site has been largely improved since the accident. Indeed, several
important tasks have been accomplished:•re-establishment of ground transport which was a priority issue;•

clean-up of the site resulting in a reduction in radiological dose rate, eased access tocritical parts of the plant for safe LTM and for the preparation of further recovery actions;•

implementation of closed-loop damaged core cooling in units 1 through 3;•

completion of the removal of the fuel from unit-4 (done late 2014);•

expanded and improved contaminated water treatment system to reduce chlorine(added with seawater injection during emergency phases) and radioactivity contents(reduction in main contributors’ contents, i.e. 137Cs and 90Sr) in water;•

implementation of measures to reduce groundwater intake and contaminated water
wastes;•installation of new tanks and associated systems for contaminated water storage.

The detailed timeline of these operations is shown in Table 2.7.
It is to be expected that the situation at the site will remain very complex. A range ofchallenging issues remain, such as the persistent underground water ingress to main buildings
and the accumulation of contaminated water on-site, the long-term management of radioactive
waste as well as those related to the removal of nuclear fuel, damaged fuel and fuel debris.

1.

“Mid-term” refers to the period of time between completion of step 2 and the decommissioning of thereactor. Mid-term safety is ensured by provisions that were taken by TEPCO and reviewed by the Nuclear
and Industrial Safety Agency (NISA). Reliability of the components of the circulating water cooling systemis ensured by diversification and other measures. Means to detect anomalies and alternative measuresin the case of inoperability of the system have been provisioned. Also, provisions were set to mitigate
any significant radiological risk.

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Table 2.7. Timeline of events and operations at the Fukushima Daiichi NPP
following the 11 March accident up to July 20172011
All reactors
in cold
shutdown

Mar 11
Mar 11
Mar 11
Mar 12
Mar 14
Mar 14
Mar 15
Mar 24
Apr 2
Apr 4
Apr 11-14
May 11
May 31
Jun 17
Jun 27
Jun 30Jul 31
Aug 10Aug 182012
Start making
full-scalemeasuresagainstcontaminated
water

2013
Struggled inefforts to deal
with issues

2014Completion offuel removalfrom Unit 4
spent fuel poolOct 28Dec 16Feb 3
Apr 5
Apr 19Apr 25
May 11
Sep 22
Sep 23
Oct 2
Mar 11
Mar 18
Mar 30
Apr 5
Apr 12
Jun 19Jun 30Jul 20Aug 12
Aug 19Jan 31
Mar 8
Mar 28Apr 1
Apr 9
May 15
May 21
May 27
Jul 29Sep 17
Oct 18Dec 22
Dec 2662

The 2011 Great East Japan Earthquake struck
Second wave of Tsunami arrived, main buildings of units 1 through 6 flooded
Station blackouts at units 1 through 5
A hydrogen explosion took place in unit 1 reactor building (R/B)
A hydrogen explosion took place in unit 3 R/B
A blowout panel was forced open to release pressure inside unit 2 R/B
A hydrogen explosion took place in unit 4 R/B
Three workers exposed to high radiation in unit 3 turbine building (T/B) basement
Highly contaminated water leaked into the sea near unit 2 water intake
About 10 000 m3 of low-level contaminated water discharged into the sea
Completion of silt fence installation near units 1-4 water intakes
Highly contaminated water near unit 3 water intake leaked into the sea
Full-scale operations of alternative cooling equipment in unit 2 SFP started
Operations of water treatment facilities to mainly remove caesium started
Full-scale operations of circulating cooling water in units 1 through 3 started
Full-scale operations of alternative cooling equipment in unit 3 SFP started
Full-scale operations of alternative cooling equipment in unit 4 SFP started
Full-scale operations of alternative cooling equipment in unit 1 SFP started
Operations of water treatment facility Simplified Active Water Retrieval and Recovery
System (SARRY) to mainly remove caesium started
Completion of building cover installation for unit 1 R/B
All of the reactors in cold shutdown (below 100℃) declared by the government
About 10 L of highly contaminated water leaked from a storage tank
About 12 m3 of contaminated water leaked from water treatment facility
Permanent shutdown of units 1 through 4 declared
Installation work started for seaside impermeable wall
Completion of seabed paving work at units 1 through 4 water intakes
A steel frame fell into unit 3 SFP during rubble removal work
Completion of Multi-Nuclide Removal Facility (ALPS) construction
Start construction of groundwater bypass pumping system
Unit 2 blowout panel closed to prevent further spread of radioactive materials
Power outage occurred and caused temporary suspension of SFP cooling system
Started test operations of Multi-Nuclide Removal Facility (ALPS)
Confirmed contaminated water leakage from underground reservoir
Start of dry cask operations to store spent fuel in the shared pool
A high level of radioactivity detected in monitoring wells at bank protection
Start of Entrance Control Building operations
Completion of building cover construction at unit 4 for spent fuel removal
Ten workers exposed to radiation from dust scattered during unit 3 rubble removal
About 300 tonnes of contaminated water leaked from bolted-type storage tanks
Permanent shutdown of units 5 through 6 declared
Completion of rubble removal inside the unit 4 SFP
A fatal accident at solid waste storage facility: a worker died in a mudslide
Fukushima Daiichi Decontamination and Decommissioning Engineering Company set up
Start of groundwater bypass pumping operations
Accumulated water leaks found around expansion joints of unit 3 main steam pipes
Started discharging groundwater pumped up by Groundwater Bypass system
Accumulated water leakage found in unit 1 Pressure Suppression Chamber
Completion of drainage B and C channels reconfiguration to discharge to the port
Test operations of Advanced Multi-Nuclide Removal Facility (ALPS) started
Test operations of High-Performance Multi-Nuclide Removal Facility (ALPS) started
Completion of fuel removal from unit 4 SFP
Water treatment facilities SARRY and KURION (named after the US waste treatmentcompany) improved to remove Strontium

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Table 2.7. Timeline of events and operations at the Fukushima Daiichi NPP
following the 11 March accident up to July 2017 (cont’d)
Jan 10
Jan 19Feb 12
Feb 24
Apr 10-20
Apr 17
Apr 23
May 272015
May 31
Completion of Jun 30
contaminatedJul 30water
Aug 2
treatment
Aug 8
Sep 3
Sep 14
Oct 5
Oct 20, 22
Oct 26
Nov 26
Dec 8
Dec 11
Feb 9
Mar 8
Mar18
Mar 22
2016
Mar 28Start of fullMar 31
scale
investigation April 12
for fuel debris Jun 20
removalSep 21
Sep 30
Nov 10
Dec 5
Jan 172017
Feb 16
Fuel debris
Mar 18
removalmethods for
Jul 19each units to be Jul 31
determined
Aug 22

Reverse-osmosis concentrated water treatment operations to mainly remove Strontium
startedA fatal accident: a worker fell from the top plate of a water storage tank
Investigation inside unit 1 PCV carried out using muon technology
Tainted rainwater leakage found from drainage into the sea
Investigation inside unit 1 PCV carried out using a robot
Start transferring drainage K water into the port with temporary pumps
Completion of seabed paving work for the entire port to seal in floating mud
Highly contaminated water treatment complete, stored in tanks after the accident
Large Rest House operations started
Highly contaminated water removal inside unit 2 seawater piping trench complete
Highly contaminated water removal inside unit 3 seawater piping trench completed
Completion of fuel handling machine removal in unit 3 SFP
A fatal accident: a worker died in the middle of cleaning a construction vehicle
Start of groundwater pumping by Sub-drain System installed around buildings
Start of treated groundwater discharge to the port, pumped up by Sub-drain System
Completion of roof panel removal of unit 1 building cover
Investigation inside unit 3 primary containment vessel (PCV) carried out using cameras
Completion of Seaside Impermeable Wall closure
Investigation around equipment hatch of unit 3 PCV carried out using small robot
Expansion of the areas where workers can move around in normal work uniforms
Highly contaminated water removal inside unit 4 seawater piping trench completed
Completion of landside impermeable wall installation
Adjustment of radiation protection areas and appropriate worker's outfits
Operation of miscellaneous solid waste incineration facility started
Investigation inside unit 2 reactor carried out using muon technology
Reconfiguration of drainage K outlet to the port completed
Freezing of landside impermeable wall started
Shielding materials started to be installed on the top floor of unit 3 R/B
New drainage channels constructed on the premises
Groundwater levels increased in the area 4 m above sea level due to extended rain
Construction of new main administration building completed
Completion of water panel removal (18 panels) from unit 1 R/B cover
Unit 3 core injection stopped temporarily but resumed with an alternative pump
New roof installation began on unit 3 R/B for spent fuel removal
Unit 2 PCV investigation conducted with a robot
Unit 1 PCV investigation conducted with a robot
Unit 3 PCV investigation conducted with a robot
Installation of unit 3 spent fuel removal cover dome started
Closure of last part of landside impermeable wall started

Present knowledge of PCV and RPV status
The knowledge of PCV leak paths and of the location and status of the fuel debris is of prime
importance to progressing safely in the fuel recovery preparation and plan the corresponding
actions. The investigations on the conditions of the PCVs and RPVs are progressing based on theobservations reported by TEPCO (from visual inspections, dose rate measurements,
identification of leakage paths etc. using robotics means) and with the support of accident
analyses (NDA, 2011). However, to this date, the access to both PCV and RPV is still challenging
due to high dose rates and the existing rubble. Further examinations and recovery actions areengaged and planned to gain further access and insights to PCV and RPV status. Of particularinterest for recovery actions are the PCV leak paths and the location and status of the damaged
fuel (usually called “debris” in discussed fuel retrieval plans).

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Temperature and dose rate in PCV have shown stable decrease and settling during the past
five years since the accidents (Figures 2.32 and 2.33). Based on inspections conducted so far,possible water leak paths of PCV have been located for each unit (Figure 2.34). For the
suppression chamber in unit 2, further efforts for identifying leak paths under the water surface
level have been made as described later.
Progressions towards decontaminating reactor buildings and removal of debris are briefly
described as follows:•

For unit 1, structures covering the damaged reactor building were removed and after theremoval of the cover, removal of rubble on the refuelling floor will be started. Robots
have been used to identify leakage paths from the PCV.

•For unit 2, the reactor building was not damaged and robots were used for examinationsin the refuelling floor zone (retrieving concrete samples) and in the PCV.

•For unit 3, rubble removal above the refuelling floor from the damaged reactor building
using remote controlled heavy machinery was completed. Rubble was also removed
from the SFP to prepare for fuel removal. Spent fuel removal equipment is being installed
after shielding installation.

Figure 2.32. Temperature trends during five years after accidents
in Fukushima Daiichi units 1, 2 and 3

Source: TEPCO (2018).

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Figure 2.33. Dose rate trends during five years after accidents
in Fukushima Daiichi units 1, 2 and 3

Source: TEPCO (2018).

Figure 2.34. Possible leak paths in Fukushima Daiichi units 1, 2 and 3

Source: TEPCO (2018).

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STATE OF THE ART OF LONG-TERM MANAGEMENT AND ACTIONSContainment overview and identified major leakage paths
Unit 1: A large amount of water and steam during the accident has likely been discharged into
the torus room through possible failure portions of the containment and suppression chamber(S/C), where leaking water flow indications were found by visual inspections on the PCV (suchas vacuum breaker lines on the S/C, and sand cushion drain pipes between the dry well and thetorus room). There seems to be no leakage from S/C lower parts below water level in the torus
room, because the water level inside S/C is maintained high. S/C itself is assumed leak tight,
except the leaking portion at the expansion joint cover of the pipes on the upper part of the S/C.
On the other hand, leakage holes in the PCV dry well are considered small, as water level insideof the PCV can be maintained at 2.5 m above the dry well bottom floor.
Unit 2: Water is leaking from the lower part of the S/C. The PCV dry well water level stays
relatively low since injected water flows into the S/C and is assuredly leaking from the S/C.
A remote controlled device with an ultrasonic water level sensor was sent from the torus room to
the S/C and was moved along its outer peripheral surface. It was identified that there is adifference in water level between inside and outside of S/C. The water level of S/C depends on PCV
pressure increase caused by nitrogen injection. It is interpreted that a moderate change in thiswater level (about 10 mm to 30 mm) is caused by submerged leak path(s) in S/C (Figure 2.35).

Figure 2.35. S/C water level measurement in Fukushima Daiichi unit 2

Source: TEPCO (2018).

Unit 3: Water is leaking through the PCV dry well from the expansion joint of the main steamline (MSL). No leakage is occurring from the PCV dry well below the MSL and no abnormality isreported about S/C. The water level can be maintained at 6.3 m above the dry well floor (up to
the level of the MSL penetration).

RPV status and estimated degraded fuel location
Analyses of cooling water temperature evolution at different RPV locations (heat balancemethod) provides indications that a large part of fuel debris was released from unit 1 RPV
whereas some degraded fuel is still present in the unit 2 and 3 RPVs. Taking into account theseobservation results and Modular Accident Analysis Programme (MAAP) calculation results
reported in the NEA Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power
Station (NDA, 2011), the present status of debris in RPV and PCV are estimated for units 1, 2
and 3 as depicted in Figure 2.36.

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Figure 2.36. Current status in the three damaged units of Fukushima Daiichi units 1, 2 and 3

Source: TEPCO (2018).

A summary of the investigations performed in damaged units 1, 2 and 3 is provided inTable 2.8.
An advanced radiographic measurement based on the muon tomography technology has
been implemented in units 1 and 2. It was observed that major structural parts can be visualised
as light and shade pixel maps. In unit 2, it was found that high-density materials reside in thelower head of RPV (Figure 2.37). In comparison with the calculation results shown in Figure 2.36,
this observation indicates that a significant part of debris is located in the lower head. It wasalso suggested that a substantial amount of fuel debris may remain in the lower and peripheral
parts of the core although images of these parts present uncertainties caused by internal
structures.
Investigations using self-propelling robots were conducted for units 1 and 2 in January to
March 2017. Results are summarised hereafter.
Unit 1: The Hitachi-developed Primary Containment Vessel Internal Survey Equipment(“PMORPH”) robot made several investigations into the PCV in five days and completed its
mission on 22 March. Equipped with a dosimeter and waterproof camera, it took dose
measurements (Figure 2.38a) and digital images (Figure 38b) at ten different points. Deposits
were found at the bottom of the PCV and on piping. The deposits will be analysed after taking
samples. Radiation dose levels decrease upon submerging into the water and rise again whennearing the floor (Figure 2.38c). There has been little change in dose levels above the gratingcompared to the previous survey in April 2015 and no significant damage was found to theexisting structures during this period. It will take more time to conclude whether fuel debris
exists at the measured points or how far it spread over the PCV basement.

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Figure 2.37. Muon tomography visualisation in Fukushima Daiichi unit 2

Source: TEPCO, www.tepco.co.jp/en/nu/fukushima-np/handouts/2016/images/handouts_160728_01-e.pdf, see also www.irid.or.jp/en.

Figure 2.38a. “PMORPH” robot dose rate measurements in unit 1 PCV

Source: TEPCO, www.tepco.co.jp/en/nu/fukushima-np/handouts/2017/images/handouts_170327_01-e.pdf, see also www.irid.or.jp/en.

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Figure 2.38b. “PMORPH” digital images in unit 1 PCV

Source: TEPCO, www.tepco.co.jp/en/nu/fukushima-np/handouts/2017/images/handouts_170327_01-e.pdf, see also www.irid.or.jp/en.

Figure 2.38c. “PMORPH” radiation dose levels in unit 1 PCV

Source: TEPCO, www.tepco.co.jp/en/nu/fukushima-np/handouts/2017/images/handouts_170327_01-e.pdf, see also www.irid.or.jp/en.

Unit 2: In order to investigate the PCV and clarify the conditions of debris and surrounding
structures, Toshiba’s self-propelled robot was deployed from January to February 2017 and theresults displayed in Figures 2.39a-c were obtained.

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Figure 2.39a. Robot inspection of the unit 2 control rod drive rail area

Source: TEPCO, www.tepco.co.jp/en/nu/fukushima-np/handouts/2017/images/handouts_170215_01-e.pdf, see also www.irid.or.jp/en.

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Figure 2.39b. Robot inspection of the unit 2 entrance of pedestal area

Source: TEPCO, www.tepco.co.jp/en/nu/fukushima-np/handouts/2017/images/handouts_170215_01-e.pdf, see also www.irid.or.jp/en.

Figure 2.39c. Robot inspection of the unit 2 pedestal area

Source: TEPCO, www.tepco.co.jp/en/nu/fukushima-np/handouts/2017/images/handouts_170215_01-e.pdf, see also www.irid.or.jp/en.

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STATE OF THE ART OF LONG-TERM MANAGEMENT AND ACTIONSUnit 3: TOSHIBA’s submersible remote operated vehicle was launched for an exploration of
the inside of the unit 3 PCV in July 2017 and the results displayed in Figures 2.40a-c wereobtained. The exploration revealed damage to multiple structures inside the pedestal. Also,
likely melted materials that are consolidated and some fallen substances such as grating andsediments were confirmed. The conditions inside the pedestal will be examined precisely basedon the image data obtained through the consecutive explorations.

Figure 2.40a. Unit 3 internal investigation using underwater remote operated vehicle

Source: TEPCO, www.tepco.co.jp/en/nu/fukushima-np/handouts/2017/images/handouts_170713_02-e.pdf, see also www.irid.or.jp/en.

Figure 2.40b. Unit 3 PCV internal investigation results

Source: TEPCO, www.tepco.co.jp/en/nu/fukushima-np/handouts/2017/images/handouts_170722_01-e.pdf, see also www.irid.or.jp/en.

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Figure 2.40c. Unit 3 PCV internal investigation results

Source: TEPCO, www.tepco.co.jp/en/nu/fukushima-np/handouts/2017/images/handouts_170721_01-e.pdf, see also www.irid.or.jp/en.

Table 2.8. Summary of investigations performed in units 1, 2 and 3
Unit

Place of investigation
1st internal investigation(October 2012)

2nd internal investigation:
PCV internal Investigation outside theinvestigation pedestal (April 2015)
3rd internal investigation:
Investigation on the lower
outer part of the pedestal(scheduled in March 2017)

Investigation results• Confirmation of no major damage to various facilities inside
PCV (air conditioning equipment, recirculation pump andpiping, pedestal wall, etc.).
• Confirmation of temperature, radiation levels, and water levelin PCV: atmospheric temperature: about 18~21°C, radiationlevels: about 5~10 Sv/h, water level: approximately 2.8 m fromthe D/W basement floor.
• Collection of accumulated water.
• Installation of permanent monitoring instrument.
• Measurement of atmospheric temperature, radiation levels,
underwater radiation levels, sediment collection, replacementof permanent monitoring instruments.

Investigation around the lower • Confirmation of leakage from sand cushion drain line as avent pipe (November 2013)leakage point from PCV.

Unit 1

Investigation of the pressure
suppression chamber upper
part (May 2014)
Investigation TIP room investigation
inside the(September to October 2015)
building
Investigation of the main
steam isolation valve roomand the air lock chamber(November to December
2015)
OthersMuon measurement(February to May 2015)• Confirmation of leakage from the expansion joint of the PCVvent pipe vacuum break line as a leakage point from PCV.
• Instrumentation penetration had high radiation levels of
100 mSv/h or more and others had low radiation levels.
• In the main steam line isolation valves room, the status of thebellows section could not be confirmed due to the inference ofthe equipment.
• The air lock chamber has high radiation levels of 7 Sv/h or moreat the lower part of the PCV penetration section, and the inside of
the bellows cover is estimated to be a source of contamination.
• Confirmation of not much fuel in the core area.

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Table 2.8. Summary of investigations performed in units 1, 2 and 3 (cont’d)
Unit

Place of investigation
1st internal investigation(January 2012)
2nd internal investigation(March 2012)
3rd internal investigation:
Investigation inside the pedestal(August 2013)

PCV internal
investigation
4th internal investigation:
Investigation inside the pedestal(January to February, 2017)

Unit 2

Investigation results• Confirmation of sediments and fall-off and deformation ofgratings inside the pedestal.
• Confirmation of no serious damage to the control rod drive
housing support near the pedestal entrance.
• Confirmation of temperature, radiation levels, water level
inside PCV.
1st and 2nd time:
Atmospheric temperature:
about 43℃~46℃
Radiation levels: about 31~73 Sv/hWater level measurement: about 0.3 m
3rd time: Atmospheric radiation levels:
about 24 Sv/h~36 Sv/h
Collection of accumulated water,
Installation of permanent monitoring
instruments
4th time: Atmospheric temperature: 16.5℃
Radiation levels: about 210 Sv/h* on the control rod drive rail
*As a result of the validation implemented after the
4th investigation, this radiation levels are amended;
210 Sv/h to 70 Gy/h.
• Sediment collection.

Investigation of the torus room
Investigation wall (July 2014)
inside theInvestigation of the pressurebuilding
suppression chamber lower part

• No leak point from the PCV was found.

(September 2014)

OthersMuon measurement(March to July 2016)

PCV internal 1st internal investigation
investigation (October to December, 2015)

Unit 3

Investigation of the PCV
equipment hatch (September
and November, 2015)
Investigation of the main steam
Investigation
isolation valve room (May, 2014)
inside thebuilding• Confirmation on the presence of high-density substancesconsidered to be fuel debris at the bottom of the pressure
vessel, in the lower part of the core and in the outer peripheryof the core.
• Damage to the structures or wall inside the PCV was notfound, and internal radiation levels were confirmed to belower than those of units 1 and 2.
• Confirmation of temperature, radiation levels, water level
inside the PCV.
Atmospheric temperature: about 26~27℃
Radiation levels: about 1 Sv/hWater level: about 6.3 m
Collection of accumulated water
Installation of permanent monitoring instruments• Confirmation of contamination and rust around theequipment hatch in lower level areas due to the water levelin the PCV.
• Confirmation of leakage from the main steam pipe
expansion joint as a leakage point from the PCV.

• Confirmation of multiple damaged substances and somepart of the control rod drive housing supporting clamp
Investigation of the inside PCV byinside of the pedestal.
submersible remote operated
vehicle (July,2017)• Confirmation of likely melted materials that are solidified,some fallen substances such as grating and sediments.

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STATE OF THE ART OF LONG-TERM MANAGEMENT AND ACTIONSRegulatory framework and implementing long-term management plans
After the tsunami had induced the accident at the Fukushima Daiichi NPP, emergency measures
were taken to cope with the resultant hazardous situation based on Article 64, paragraph (1) of
the Reactor Regulation Act. The reason why it was necessary to revise the said Act adding
Article 64-2, 64-3, and 64-4 was that the accident was beyond the framework of pre-existing
Article 64 and that it should take quite long time for recovery or decommissioning.
Based on Article 64-2, paragraph (1), the Nuclear Regulation Authority (NRA) of Japan
designated the Fukushima Daiichi NPP as specified nuclear power facilities, which are definedby this provision as facilities to be placed under special management, on 7 November 2012.
Based on paragraph (2) of the said Article, the NRA required the relevant licensee to submit aplan to implement measures for the operational safety of the facilities (implementation plan),
indicating “matters for which the measures should be taken” and the time limit therefor. The
matters consist of those regarding the entire process and risk assessment, infrastructure,
security, protection of nuclear material, fuel debris removal and decommissioning, planning of
implementation, implementation, and inspection.
On 7 December 2012, the NRA received the implementation plan prepared by TEPCO basedon Article 64-3, paragraph (1) of the Reactor Regulation Act. Subsequently, the NRA establishedthe Supervision and Evaluation Committee for the Specified Nuclear Power Facilities to help
investigate validity of the implementation plan and the status of the implemented measuresthrough discussion with external experts. The Committee examined whether each facility
described in the implementation plan, and each countermeasure, conforms to “the matters forwhich measures should be taken,” or safety requirements, and other points using results of site
inspection for reference. After a series of discussions of the Committee, it was confirmed thatthe submitted implementation plan conformed to the required goal. The NRA acknowledgedthat the implementation plan was sufficient to provide protection from nuclear fuel materials
or contaminated objects as well as to prevent reactor-related disasters and protect the specified
nuclear fuel materials, and approved the plan on 14 August 2013. The NRA also indicatedprecautions to be observed when implementing the plan, with regard to risk evaluation,
monitoring reactors and equipment, fuel removal, storage of radioactive waste,countermeasuresagainstcontaminatedwater,
radiation
protection,
emergency
countermeasures, response to tsunamis, organisation structures, removal of fuel debris, andpromoting public understanding to implement the plan.

Figure 2.41. Implementation actions for specified nuclear facilities and progress

Source: NRA (2013), www.nsr.go.jp/data/000067054.pdf.

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Based on Article 64-4 of the Reactor Regulation Act, only part of the provisions of the said
Act was decided to be applied to the Fukushima Daiichi NPP, as specified by the Cabinet Orderestablished and enforced on 8 March 2013, as long as measures for operational safety are
implemented in accordance with the implementation plan (Figure 2.41).
The provisions of the Reactor Regulation Act in English are available at www.nsr.go.jp/data/000067232.pdf.

Approaches to risk for long-term management

Risk mapping

In the “matter for which the measures should be taken” indicated by the NRA regarding theentire process and risk assessment, it is required to reduce and optimise risks of the specified
nuclear power facilities as a whole, clarifying the entire process and evaluating each operationand stage for the entire process of measures towards completing decommissioning including
removal and storage of melted or damaged fuel rods from units 1 to 4, and for the entire process
to maintain cold shutdown stably for units 5 through 6. It is also required that risk assessment
is applied to a large extent assessing the environmental impact in broad areas outside the plant,
thereby ensuring that the reduction and optimisation of risk is sufficient for safety inside andoutside the premises. The Supervision and Evaluation Committee for the Specified Nuclear
Power Facilities has examined the plant status reported by the licensee regarding thecompleted/ongoing/planned measures against risks and the current situation. The committee
has also asked the licensee to extract and rank risks in the whole facilities and to report on
assessment of risk reduction.
TEPCO has complied with this request conducting risk assessment to clarify objectives,
identify risk matters, supposing accident scenarios or causing events, designing measuresagainst major scenarios, validating the measures, and designing unimplemented measuresagainst residual risk matters in the short and mid- and long term. The unimplemented measures
were classified into five categories: (1) those for external events (e.g. earthquake/ tsunami or fire),(2) those for situations worsening in the near future (e.g. storage capacity for contaminated waterand solid wastes), (3) those for situations triggered by implementing measures (e.g. groundwaterlevels or radioactive dust), (4) those for mid- and long-term management (e.g. spent fuel orradiation exposure), and (5) those for reliability of infrastructure (e.g. prevention or monitoringof leakage). TEPCO realised that the measures belonging to categories (1) to (3) should be
implemented as soon as possible. The NRA accepted TEPCO’s risk assessment requiring
sustained efforts to reduce risks in varying situations, especially risks of contaminated waterleakage.
Besides the mentioned risk assessment, the NRA suggested to compile a kind of risk map tovisualise risk issues of the Fukushima Daiichi NPP affecting the environment. Such risk map
should be designed to relate progress of safety operation to risk reduction clarifying completed
measures rather than to quantify each risk source. Accordingly, the Secretariat of the NRA
proposed a template for a map of measures for mid-term risk reduction indicating about fiveyear perspective regarding seven issues, five main issues plus two related ones: contaminatedwater, radioactive wastes, spent fuel pool, earthquakes and tsunamis, effective dose at the siteboundary (estimated), work environment, and examining the inside of the facilities. The risk
map, “Measures for Mid-term Risk Reduction at TEPCO’s Fukushima Daiichi NPS”, has been
compiled with the following objectives:•to present key priorities for safety identified by the NRA among the various measures
undertaken by TEPCO for the decommissioning of the Fukushima Daiichi NPP;•

to clearly distinguish completed measures from ongoing and planned measures.

The progress made for each objective will be measured by the risk map, while this document
will be regularly reviewed to reflect the state of risk reduction. The NRA has released the risk
map five times until 12 July 2017, adding one issue on measures for preventing scattering of
dust (Figure 2.42). The past editions of the risk map in English are available atwww.nsr.go.jp/english/library/nraplans_01.html.

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STATE OF THE ART OF LONG-TERM MANAGEMENT AND ACTIONS

Figure 2.42. Measures for mid-term risk reduction at TEPCO’s
Fukushima Daiichi NPP as of July 2017

Source: NRA (2017), www.nsr.go.jp/data/000201934.pdf.

Risk assessment

The Decommissioning Office of the Nuclear Damage Compensation and Decommissioning
Facilitation Corporation (NDF) is the organisation that was created to facilitate Fukushima Daiichi
decommissioning. The strategic plan developed by NDF includes many aspects related to safe andprompt decommissioning of the plants: risk-driven decommissioning strategy, fuel debris
retrieval strategy considering various options, radioactive wastes management strategy, researchand development (R&D) plan for decommissioning, and organisational relationship between
various actors (government, NDF, research institutes, TEPCO who is performing the on-site
decommissioning).
NDF runs a fully comprehensive risk assessment, which can substantially be divided into
two different approaches (NDF, 2016):
1. a risk reduction strategy aimed at prioritising LTM activities upon the level of riskpresented by each type of radiation source;
2. a risk analytical approach based on splitting the LTM tasks into fuel debris or wasterelated activities.



Risk reduction strategy

The first approach stems from adapting the UK Nuclear Decommissioning Authority’s (NDA)
so-called safety and environmental detriment (SED) score which is one of the scores used toprioritise activities at the 17 sites ruled by NDA (2011). Such score takes into account theradiation source in terms of stored radioactivity, how easily such radiation can be released in
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case of loss of containment function which depends on the physical phase, and the recovery
time under generic threatening events (such as loss of cooling, loss of inerting, etc.). These threefactors represent the consequence side of risk according to the following formula:

𝑅𝑅𝑅𝑅𝑅𝑅 =𝐼𝐼𝐼𝐼𝐼𝐼𝐼𝐼𝐼𝐼𝐼𝐼𝐼𝐼𝐼𝐼𝐼𝐼 ∙ 𝐹𝐹𝐹𝐹𝐹𝐹𝐹𝐹 𝐹𝐹𝐹𝐹𝐹𝐹𝐹𝐹𝐹𝐹𝐹𝐹
𝐶𝐶𝐶𝐶𝐶𝐶𝐶𝐶𝐶𝐶𝐶𝐶𝐶𝐶 𝐹𝐹𝐹𝐹𝐹𝐹𝐹𝐹𝐹𝐹𝐹𝐹

where RHP stands for radiological hazard potential and it is a measure of risk through theinventory variable which is proportional to the stored radioactivity by each source; Form Factor
which depends on whether the radiation source is stored as a gas, liquid or types of solid; andControl Factor which attempts to give a measure of the available recovery time from a degradedsituation.
On the likelihood side of risk, two factors help quantify the priority with regard to
management to maintain safety: a first, Facility Descriptor or FD index categorises the suitabilityof containment by focusing on the status of barriers between the radioactive source and theenvironment. According to the status of the containment, a specific FD is assigned. A second,
Waste Uncertainty Descriptor or WUD determines whether the radioactive source is degrading or
chemically reactive and how monitored, managed and packaged it is in order to categorise thelikelihood of deterioration of the source material such that future removal will be more onerous,
ending up in the following SED formula:
𝑆𝑆𝑆𝑆𝑆𝑆 = 𝑅𝑅𝑅𝑅𝑅𝑅 ∙ (𝐹𝐹𝐹𝐹 ∙ 𝑊𝑊𝑊𝑊𝑊𝑊)4

where Chemical Hazard Potential similar to RHP is disregarded. Identified radioactive
sources comprise the fuel debris inside the PCV of the three units, plus the fuel elements stored
in all the SFPs including the common pool and dry casks, the contaminated water distinguishing
the source upon a different location is identified, and all types of secondary and solid wastes.
A SED score is then calculated for each of these sources for a series of major radionuclides andthen compared in terms of the two main risk components of consequence or hazard RHP, andlikelihood or safety management (𝐹𝐹𝐹𝐹 ∙ 𝑊𝑊𝑊𝑊𝑊𝑊)4 according to Figure 2.43.

Figure 2.43. Ranking of hazard and likelihood risk components at Fukushima Daiichi

Source: NDF (2016), www.dd.ndf.go.jp/en/strategic-plan/book/20170322_SP2016eFT.pdf.

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STATE OF THE ART OF LONG-TERM MANAGEMENT AND ACTIONSTherefore, this comprehensive approach to risk allows facing different risk sources at onceand revealing where higher risks are stored with the scope of prioritising LTM actions. In thissense, such approach does not yet enable the user identifying and quantifying the specific sourcesof risks specific of every risk source. This is why a second different strategy is necessary.

Analytical approach to risk

The proposed analytical approach consists of identifying and categorising the different actions
addressing the LTM within the two main broad types of issues of fuel debris and waste
management. Such analytical task will be performed through a three-step approach:
1. A preliminary step based on analysing every action and identifying how to specificallyaddress it at best.
2. Development of a so-called Logic Tree as a chart tool that easily allows identifying andarranging the needed tasks under a well-ordered structure.
3. Application of a generic risk method tool such as failure mode and effects analysis or
IAEA-developed DRiMa (Decommissioning Risk Management) devoted specifically to
decommissioning issues.
When making such analysis, it is essential to take into account all types of risk factors
during the decommissioning activities not only dealing with risk reduction or safety, but alsowith workers’ health, environment, security or financial affairs, as recommended by DRiMa or
the NDA Value Framework.

Long-term management for risks reduction in relation to plant damaged states

This section describes the progress of LTM implemented at the Fukushima Daiichi damaged site.
In the early stage after the accident, the focus was primarily on recovery of critical safetyfunctions and necessary systems and components. Further actions were then implemented toenhance robustness and monitoring of these critical safety functions, manage produced liquidand solid wastes, remove stored fuels from SFPs, implement countermeasures against external
hazards, improve working environments, reduce dust dispersion and the site boundary doserate, perform site inspections and prepare for decontamination and defuelling of the damaged
RPVs and PCVs. LTM planning was driven by reductions in highly identified risks depending on
the plant damaged state. Progresses in recovery actions have changed the plant damaged state
by reduction of some risks and thus the LTM focus with time.

Monitoring and strengthening of safety systems for long-term management

LTM management actions after reaching a stabilised controlled state are primarily designed tofulfil the three following “critical” safety functions:•maintain the sub-criticality;•

maintain the cooling in the RPV and in the PCV;•

maintain the controlled release.

The sub-criticality in the three units containing degraded fuel is monitored every hour by
Xe detection in the PCV gas control system (Figure 2.44), by temperature evolution
measurements of the RPV lower head and by air dose rate measurements by monitoring posts.
The criteria have been defined for checking sub-criticality. If the observed value deviates fromthe criteria, measures would be implemented to provide borate injection through the liquid
control system and eventually, once the liquid control system is exhausted, by seawater
injection. Up to this date, no deviation from sub-criticality has been detected.
135

The cooling of the RPV and the PCV in the three units containing degraded fuel is ensured
by a closed loop, injecting water in the RPV, using either the core spray safety system or the feedwater system. Out-coming water is treated and re-injected in the core to limit volumes of water
wastes (see the next section). Cooling is monitored by temperatures of the RPV lower head andthe PCV atmosphere, which should stay below 80°C. One issue in monitoring was the reliability

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of thermocouples after they had experienced severe accident conditions. Selection of reliable
thermocouples had to be made based on analysis of their electric characteristics and response
evolution. Monthly reliability analyses of thermocouples are performed. Temperatures andinjected flowrates are checked every day. Flowrates for each reactor were reduced from 4.5 m3/h
to 3.0 m3/h until early 2017.
The confinement of radioactivity remains a critical aspect as the containment leak tightness
was lost during the accident in the three damaged units. As reported earlier, identifying all
containment leakage paths is challenging. Atmospheric releases from the PCV are however
reduced by high-efficiency particulate air (HEPA) filtration and monitored by γ-counting in the
PCV gas monitoring system (Figure 2.44). N2 injection to the PCV is done to prevent an H2
explosion in the PCV; H2 being produced by water radiolysis. H2 and O2 concentrations aremeasured in the gas monitoring system and it is checked that H2 concentrations remain below
limits that would permit combustion. N2 flow and H2, N2 and O2 contents are checked every day.
Concerning waters used for RPV and PCV cooling, their volume is reduced as much as possible
to reduce the amount of contaminated water to be treated (see the next section).

Figure 2.44. PCV gas monitoring system at Fukushima Daiichi reactors

Source: TEPCO (2018).

All units have been maintained in a cold shutdown state, with the cooling being maintainedby a 0.8 km long closed water loop. SFP cooling is ensured via existing and newly installedfacility.
As of April 2017, the temperatures are maintained between 15 and 35°C. Concentrations ofradioactive gas (135Xe) are not showing any significant changes and are close or below detection
limits of the measuring devices. No sign of abnormality in cooling nor criticality was observed.
These observations confirm that all three units are in a stable cold shutdown state.

Contaminated water management actions

The Fukushima Daiichi NPP is located downstream from abundant groundwater flowing
towards the Pacific Ocean from mountainous areas on the west side of the power stations.
Before the tsunami and the subsequent severe accident occurred, approximately 730 m3 ofgroundwater had been pumped up every day through the sub-drain system, a group of wells, inorder to control levels of groundwater around the reactor buildings and the turbine buildings of80

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units 1 through 4. Since the sub-drain, this pumping equipment and the buildings weredamaged, a large quantity of groundwater, which is combined with rainwater in rainy
conditions, has entered the buildings and been subjected to contamination. Both thecontaminated water and cooling water injected into the RPVs of units 1 through 4 are collected
as a whole. About half of the contaminated water is injected back into the RPVs to cool the fueldebris, while the rest is stored in water storage tanks after getting treated to decrease
concentrations of radioactive materials (the amount of added water was approximately
400 m3/d as of January 2014; it is reduced to 220 m3/d as of April 2017).
On 19 June 2013, TEPCO reported to the NRA that the concentration of tritium ingroundwater near the revetment between seawater inlets of units 1 and 2 at the east side of theturbine buildings had increased to 500 kBq/L, a level ten times higher than that measured in
December 2012, while that of strontium also had increased to 1 kBq/L which is much greaterthan the limit stipulated in the laws and regulations. TEPCO believed that previous fallouts wereunlikely to have affected these concentrations and that the highly contaminated water in the
trenches on the seaward side, which are underground tunnels designed to store pipes andcables, had leaked underground and mixed with the groundwater. The NRA established theWorking Group on Contaminated Water Countermeasures under the Supervision and
Evaluation Committee for the Specified Nuclear Power Facilities to examine the state of leakage
at the site as well as TEPCO’s implementation plan of countermeasures against leakage. The
NRA instructed TEPCO to strengthen its monitoring of contaminated water, immediately
implement countermeasures to prevent the contaminated water from flowing into the marine
environment, immediately implement countermeasures to prevent leakage from the trenches
on the seaward side containing highly contaminated water, which may have leaked
underground, and implement any other actions deemed necessary.
TEPCO has three fundamental principles on contaminated water management: i) to remove
sources of contamination, ii) to redirect fresh water far from contaminated areas, and iii) toretain contaminated water away from leakage to the environment. Several major management
measures have been implemented as follows.
Multiple facilities were put into operation for sequential treatment processes of contaminatedwater accumulated in the premises (Figure 2.45). First, several removal systems including KURION,SARRY and mobile devices reduced caesium and strontium in contaminated water. After
subsequent treatment for desalination, about half of the caesium/strontium treated water is
injected back into the RPVs. The rest is treated to remove most of the radioactive materials except
tritium at a multi-nuclide removal facility with an advanced liquid processing system (ALPS). The
ALPS treated water is finally stored in storage tanks. ALPS was confirmed to have the capability to
reduce concentrations of 62 nuclides under the designated concentrations limits. On 27 May 2015,
treatment of most of the highly contaminated salt water stored in storage tanks was completed.
From this point, ALPS treated water and caesium/strontium treated water have been stored in thestation and they have been accumulated up to approximately 1 020 000 m3 as of October 2017.
TEPCO has been expanding the storage capacity replacing existing flange type tanks with welded
type ones to realise higher reliability.
In December 2015, removal of highly contaminated water accumulated inside the seawater
piping trench connected to the seaside of the turbine buildings of units 2 through 4 wascompleted. Considering the quality and quantity of water contained in the tunnels or verticalshafts of the trench, contaminated water was transferred to the turbine buildings. The inside of
the trench was also filled with grout materials. The operation was completed in March 2017.
TEPCO has implemented countermeasures against contamination of the marine
environment: groundwater bypass, sub-drain and groundwater drain system, and seaside
impermeable wall (Figure 2.46).
The groundwater bypass was implemented in order to direct clean groundwater to flow
downhill towards the sea, bypassing the facilities. The water quality is monitored regularly toconfirm satisfaction of the discharge criteria (less than 1 Bq L–1, 1 Bq L–1, 5 Bq L–1, and 1 500 Bq
L–1 for 134Cs, 137Cs, gross beta, and 3H, respectively). The groundwater bypass has been in
operation since May 2014, and was estimated to reduce groundwater flowing into the building
basements by up to 100 m3/d.

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Figure 2.45. Flow diagram of the contaminated water treatment

Source: TEPCO (2018)..

Figure 2.46. Countermeasures against contamination of the marine environment

Source: TEPCO (2018)..

About a half of the damaged pre-existing pits had been restored and new ones had been
added to the sub-drain system around the buildings of units 1 through 4. Groundwater has been
pumped up since September 2015 in order to reduce the flow into both the buildings and thesea. The amount of groundwater pumped up, which is approximately 510 m3/d through 42 pits
as of 2017, has been decided considering the treatment capacity and keeping levels of the highly
contaminated water inside of the buildings lower than those of surrounding groundwater,
aiming to prevent the former from flowing into the latter. The collected groundwater has been
purified at a treatment facility specifically designed for low-level contaminated water andstored in temporal storage tanks. After checking quality, the purified water has been discharged
to the port area based on agreement with the stakeholders.

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STATE OF THE ART OF LONG-TERM MANAGEMENT AND ACTIONSThe seaside impermeable wall had been constructed along the coastline on 26 October 2015,
aiming to reduce risks of contaminated water flowing into the ocean if any contaminated waterleakage would occur in the premises. The wall is composed of 594 steel pipe sheet-piles driven
into the ground deeper than the lower permeable layer across a length of about 780 m alongsideunits 1 through 4, and is expected to hold groundwater flowing into the port area from thepremises. The accumulated water behind the wall has been pumped up through thegroundwater drain, which consists of five pits, preventing overflow into the port, and treated
together with the water collected through the sub-drain system.
A landside impermeable wall (ice wall) started operation in March 2016. The total length of
the wall is about 1 500 m and the amount of frozen soil is about 70 000 m3. So far, the amountof underground water including rainwater flowing into the buildings has been reduced toapproximately 100-200 m3/d.

Solid waste management actions

A huge amount of waste was generated by recovering works from the emergency situation in ashort time. Due to the limited capacity in waste storage buildings, a large part of waste wasstored outside temporarily. The waste stored in the temporary storage will have its volume
reduced as much as possible before moving to the solid waste building.
Solid waste generated in the course of decommissioning work is classified based on thesurface dose rate. Rubble exceeding 30 mSv/h is stored in the waste storage buildings. Fallentrees, used protective clothing and rubble below 30 mSv/h are stored in temporary areas. The
secondary wastes from the water treatment facility such as high integrity containers from themulti-nuclide removal facility (ALPS) are classified into categories like adsorption towers, sludge
and concentrated waste liquid then temporarily stored. The solid waste incinerator started
operation in March 2016 (300 kg/h x 2 trains) and used protective clothing is being incinerated.
The solid waste storage management plan was developed in March 2016 (revised in June2017) as scheduled in the mid-long-term roadmap. The amount of wastes in the coming ten
years is estimated at about 754 000 m3 and it turned out that the estimated amount would
exceed the current storage capacity. Construction of an additional solid wastes incinerator,
volume reduction facility and storage building are planned. With these enhanced capacities,
rubble volumes will be reduced as small as possible then it will be stored in storage buildings sothat temporary storage will no longer be needed.
The secondary wastes from water treatment are temporarily stored outdoors now. It isplanned to shift to storage in the buildings and decrease the temporary outdoor storage in the
future. The studies for the stable storage will be also planned. Considering future changes inthe decommissioning work plan which is the basis of quantity estimation of wastes, the solid
waste storage plan will be revised once a year.
Details of radioactive waste are summarised in Management of Radioactive Waste after a
Nuclear Power Plant Accident (NEA, 2016).

Management of spent fuel pools

Spent fuel pools at units 1 to 4 are maintained under desalination and recirculation cooling.
Water temperature is maintained stably and reliability of the cooling system has been enhanced.
Currently, SFP cooling systems are installed at each unit but a plan has been developed tointegrate them into a single system that will start operation in 2018.

Unit 4
Spent fuel removal at unit 4 was started in November 2013 and all the 1 533 fuel bundles were
removed from the SFP in December 2014 (2 new fuel assemblies were earlier removed in 2012).
1 331 spent fuel bundles were transferred to the common pool and 202 new fuel bundles weretransferred to SFP at unit 6. These removal operations were completed safely.

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STATE OF THE ART OF LONG-TERM MANAGEMENT AND ACTIONSUnit 3The unit 3 reactor building was damaged due to a hydrogen explosion. The fuel handlingmachine had fallen in the SFP and other equipment was scattered on the refuelling floor.
The dose rate on the refuelling floor was as high as several hundred mSv/h because of
scattered rubble. Then, the rubble including the fuel handling machine and the other equipment
were removed and decontamination was conducted. In April 2016, the dose rate has been
reduced to less than 5 mSv/h in the majority of the refuelling floor by shields installation.
A fuel removal cover for the SFP was pre-assembled into several large components at
Onahama, located about 60 km south of the Fukushima Daiichi NPP in order to reduce theon-site exposure dose by the construction work. These components were transported to the site
and are being assembled on-site as of January 2018. The fuel removal work is planned startingfrom mid-2018.

Unit 1
The reactor building cover was dismantled to prepare for rubble removal on the refuelling floor.
The dismantling work started in May 2015 with spraying of an anti-dust agent. Roof panels were
removed and investigation was conducted on the refuelling floor. In April 2016, the water sprayequipment was installed to prevent possible radioactive dust dispersion.
Investigation on the refuelling floor is being conducted. The status of fallen ceilings, thereactor well plug and the overhead crane has been established. This information will be used inplanning future rubble removal works on the refuelling floor.

Unit 2
To facilitate removal of fuel assemblies and debris from the spent fuel pool, it was decided todismantle the whole rooftop above the highest floor of the reactor building. The objective is toremove the fuel as fast as possible to reduce radioactive release risks during the removal process.

Countermeasures against external hazards

External hazards considered at the Fukushima Daiichi NPP are earthquakes, tsunamis, fires,severe rain, typhoons and tornadoes. It is important to note that the decay heat of spent fuels
and fuel debris has decreased significantly and the risk associated with radioactive material
release to the environment has also decreased. In particular, most of the volatile radioactive
gases and iodine have already decayed. The time margin in the case of suspended cooling has
significantly increased.
In the process of decommissioning, risk sources will be removed and reduced. On the other
hand, some risks associated with those factors such as spent fuels, fuel debris, waste storage
facilities, contaminated water stored in tanks and buildings will remain on-site.

Earthquakes and tsunamis
There are two sets of seismic accelerations and tsunami heights which were revised after theaccident. One is the design basis earthquake and tsunami: 600 gal and 15 m. The other is the
beyond design basis earthquake and tsunami: 900 gal and 26.3 m.
Structural integrity of the reactor buildings containing spent fuels in SFP and fuel debris isevaluated to be maintained against the design basis earthquake and the beyond design basis
earthquake (600 gal and 900 gal). At the same time, reliability of flexible emergency measureswith portable equipment needs to be enhanced.
Outflow of contaminated water accumulated in the buildings needs to be addressed in caseof the design basis tsunami of 15 m based on the size of opening and inventory withconsideration on the progress of water treatment and relative risks to the environment. Since
measures against the tsunami of 26.3 m will take long time, contaminated water treatment andstudies on safe transferring/treatment of radioactive sludge are underway in order to further
reduce risks.

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Fires
In order to protect the site from external fire events, a firebreak is established in the vicinity ofimportant facilities. The firebreak is 30 m in width taking into account the results of “Impact
evaluation on forest fire to Fukushima Daiichi NPS” (JNES-RC-2012-0002). Also, the on-site firebrigade will battle the fire on the site using fire trucks and sprinkler trucks as an initial response
in case it is anticipated that the fire may spread into the site.

Severe rain and typhoons
High-level radioactive contaminated water is contained inside reactor buildings, turbine
buildings, a rad waste building, a process main building, control buildings, a high temperatureincineration building, an incinerator and machine shop building and spent fuel pool building.
These buildings are designed in accordance with the Building Standards Act which sets
conservative assumptions on loads acting on the building from the past severe wind records.
Although it is unlikely that the buildings would lose their function due to severe rains andtyphoons, level-raising of accumulated contaminated water at the underground of thesebuildings was evaluated as follows:•

Observation records near the Fukushima Daiichi NPP show the maximum annual
precipitation is below 2 400 mm, maximum daily precipitation is 285 mm and maximum
monthly precipitation is 634 mm at Namie town.

•

Assuming conservatively that daily precipitation would be 1 000 mm and the water levelin the units 1 through 4 buildings is evaluated to raise at T.P. 2 311 mm that remains
below the level (T.P. 2 564 mm) at which the water might leak outside.

In conclusion, even under hypothetically conservative maximum precipitation, it isconfirmed that the contaminated water in the buildings would not leak out.

Tornadoes
Since the buildings containing contaminated water are constructed with reinforced concrete
structures, direct damage to the buildings caused by tornadoes is unlikely.
Pumps for water injection into the RPV and the PCV are dispersedly located on the high land,in the turbine buildings and the high temperature incineration building. A risk to lose functionsof all these pumps by a single tornado is considered to be small. However, if the time to restore
water injection is judged to be long, fire trucks are to be deployed to inject water. Fire trucks
and materials necessary for temporary water injection are also dispersedly located so that they
will not be lost at one occasion.
As for the SFP, it will be protected by practical measures such as installing a cover on thesurface. In the case of pool water leakage, mitigation measures will be taken with emergency
motor driven pumps, fire trucks and concrete pump trucks.
Water treatment equipment and electrical equipment are located in the reinforced concretebuildings and the risk of direct damage caused by a single tornado is also unlikely. If all powerbuses become unavailable, dedicated generators installed near each facility will provide
necessary power. If power trucks are available, they will be deployed to supply power toimportant equipment.

Site boundary dose rate, mitigation of dust dispersion, improving working environments

Monitoring results
TEPCO has enhanced monitoring of radioactive concentration in the sea water in and out of theport since March 2011. Although the reported concentration exceeded legal limits just after theaccident, it has decreased over time and currently it is maintained below the legal limits eveninside the port. The concentration level monitored in the sea water intake in front of units 1
through 4 showed significant decrease after completion of the seaside impermeable wall.

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STATE OF THE ART OF LONG-TERM MANAGEMENT AND ACTIONSDose evaluation at the site boundary
TEPCO reported that the effective dose attributed to radiation from rubble stored on-site,
contaminated water in tanks and radioactive materials released to the environment (liquid andgas) decreased below 1 mSv/y at the site boundary in March 2016.
At the end of March 2016, TEPCO estimated the maximum total site boundary dose rate as
0.96 mSv/y. The breakdown is as follows: radioactive gases contributed 0.03 mSv/y, direct andsky shine from various on-site facilities contributed 0.65 mSv/y, liquid waste discharge
contributed 0.22 mSv/y, and water spray using treated rain water from dyke surrounding tankareas contributed 0.066 mSv/y.

Mitigation of dust dispersion
Multiple measures were in place to protect the environment and worker health for the unit 1building cover removal work by preventing the scattering of contaminated dust. Anti-scattering
agents were used to keep dust down, small pieces of rubble that can create dust were vacuumed,
and mist sprinklers were also used.

Work environment improvement
Since July 2014, the average number of daily workers ranges from about 5 500 to 7 000. As of
March 2017, the local hiring ratio is about 55%. It was reported that the worker dose (including
TEPCO and contractors) was high just after the accident but it decreased rapidly. It was reported
that the monthly average dose was 0.30 mSv in August 2017 that is significantly lower than themonthly limit of 1.7 mSv (equivalent to the regulatory requirement of 20 mSv per year averaged
over five years).
After removal of high dose rubble, surface soil and paving, as for the on-site radiation level(except for units 1 through 4 area and waste storage areas), it was reported that the target of
5 μSv/h has been achieved in a majority of the on-site areas (it was about 3 mSv/h before theseworks).
TEPCO reorganised on-site zoning of radiation controlled areas and optimised personal
protective equipment as of 8 March 2016 (Figure 2.47). As a result, general wear or dedicatedwear for on-site works is adopted in about 90% of the site area. On-site work environment
improvement continues to further optimise personal protective equipment requirement.

Figure 2.47. Optimised personal protective equipment

Notes: *1. For works in buildings including water treatment facilities (multi-nuclide
removal equipment, etc.) (excluding site visits), wear a full face mask; *2. For works in tankareas containing concentrated salt water or Sr-treated water (excluding works nothandling concentrated salt water, etc., patrol, on-site investigation for work planning, andsite visits) and works related to tank transfer lines, wear a full face mask; *3. Specified lightworks (patrol, monitoring, delivery of goods brought from outside, etc.).
Source: TEPCO (2018).

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STATE OF THE ART OF LONG-TERM MANAGEMENT AND ACTIONSIn order for workers to easily check the radiation dose rate at the work field, 86 radiation
monitors were installed (as of March 2016) and real-time measurement can be observed on large
screen monitors in the seismic isolation building and access control building.
Welfare facilities are also being improved. Worker surveys are periodically conducted andimprovements have been made in reflection of needs from workers.
A large rest facility with a cafeteria with a capacity of 1 200 workers started operation in
May 2015, and a convenience store opened in March 2016. A meal service centre in Ohkuma
town started operation in June 2015 and provides foods to the on-site cafeteria. A new officebuilding opened in October 2016 and about 1 100 employees are working in this building.

Site inspection

Based on Article 64-3, paragraph (7) of the Reactor Regulation Act, TEPCO is required to undergo
inspections conducted by the NRA to confirm whether measures for the operational safety of
specified nuclear facilities or for the physical protection of specified nuclear materials are
implemented in compliance with the implementation plan. The NRA has often conducted
inspections with regard to starting-up of systems, checking of infrastructure, and security.
In addition to the said usual inspections, the NRA conducted special ones when necessary.
On 19 August 2013, TEPCO reported to the Secretariat of the NRA that contaminated water,which was estimated to be up to 300 m3, had leaked out of the weirs installed around a storagetank area for contaminated water, and that the contaminated water seemed to be released froma flange type water tank but the exact leakage path could not be identified. The NRA carried out
a site inspection on 23 August to observe the situation as well as TEPCO’s response to theincident. Based on the results, the NRA instructed TEPCO to assess the contamination outside
the weirs, to identify the leakage paths as quickly as possible, to strengthen monitoring of theperimeter of the said tank area, to remove the soil contaminated by the leaked water, and toinvestigate outflow of the contaminated water into the ocean.
TEPCO subsequently removed the contaminated soil and undertook an environmental
assessment through close monitoring, and also implemented countermeasures to prevent
spread of leakage, such as replacing the old tanks with welded tanks, storing water flooded
inside the weirs including rainwater in temporal storage tanks, raising the weirs, and coating
the concrete floors under storage tanks. Implementation of these countermeasures was
confirmed by the Fukushima Daiichi nuclear safety inspector’s office and by the Working Groupon Contaminated Water Countermeasures.

Preparation of defuelling actions

Many efforts and resources are now placed on the preparation of defuelling actions with many
actors and initiatives in Japan.
In order to address the profound challenges arising from the accident, tasks involving
interconnected, diverse and evolving conditions were embodied in waste management anddecommissioning and a new branch of TEPCO, called the Fukushima Daiichi Decontamination
and Decommissioning Engineering Company (FDEC), has been created; this is focusing on
contaminated water countermeasures, compilation of international knowledge on debris andfuel removal, and progression towards long-term resources for the decommissioning. Much
information is being published on a ministry website (see www.meti.go.jp/english/earthquake/nuclear/decommissioning/index.html).
The establishment of the NDF strategic plan faced challenges as the access to contaminatedareas and in particular PCV and RPV is difficult and information on the PCV and RPV situationis still too uncertain. A few more years will probably elapse before fuel debris retrieval will start,
with significant R&D on debris characterisation and on retrieval methods and tools in progress.
Figure 2.48 summarises, for instance, the present strategy proposed by NDF for investigation of
degraded fuel in-vessel and how it interacts with the decision and planning of retrieval
operations.

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Two main options facing different challenges have been discussed for fuel debris retrieval:•

Degraded fuel retrieval under water: this would ease fuel cooling, prevent dispersion ofradioactive material during retrieval operations and provide shielding against high dose
rate but would face major challenges such as repairing the PCV leaks for submersion,
maintaining sub-criticality during submersion, ensuring the submerged structures
mechanical resistance (in particular to seism), and managing the contaminated waters
after flooding.

•

Degraded fuel retrieval in partial submersion conditions: the approach would face major
challenges such as less-shielded high dose rates, risks of radioactive dust dispersion at
the time of degraded fuel retrieval, and device survivability to high dose rates.

A risk-informed approach is used for safe fuel debris retrieval incorporating in-coreinspection results and up-to-date R&D results as illustrated in Figure 2.49. Current knowledge
and technology may recommend that the efforts be focused on the partial submersion method
which is more feasible than the submersion method due to difficulty in repair of the top of the
PCV.
The Supervision and Evaluation Committee for the Specified Nuclear Power Facilities
established by the NRA has also investigated precautions for safety involved in defuelling
actions examining TEPCO’s specific implementation plans and assessing required safetymeasures.

Figure 2.48. Strategy for debris characterisation

Source: NDF (2016), www.dd.ndf.go.jp/en/strategic-plan/book/20170322_SP2016eFT.pdf.

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Figure 2.49. Logic tree diagram for reduction of risks related to fuel debris retrieval

Source: NDF (2016), www.dd.ndf.go.jp/en/strategic-plan/book/20170322_SP2016eFT.pdf.

Summary of main implemented countermeasures

Reactor coolingMultiple and diverse cooling facilities were installed shortly after the earthquake and thefacilities have been upgraded since then so that stable cooling conditions could be maintained.
The decay heat in the reactor fell down to 1/100 or less of that at the earthquake time andit is evaluated that even if the cooling facilities are damaged due to a large-scale earthquake ortsunami and re-cooling requires 12 hours, the effective dose at the site boundary would notexceed 1.5 x 10-6 mSv/y. This dose is 500 000 times less than that measured six months after theearthquake.
The cooling status of the spent fuel pools (SFPs) has been maintained in a stable manner
through a long period of time, and it is evaluated that even assuming that the cooling facilities
are temporarily lost, the time delay until the water temperature of the spent fuel pool would
reach 100℃ is more than 200 hours, even in the most severe case of the common pool.

Spent fuel removal
Spent fuels from unit 4 were removed in December 2014. Currently, equipment operations arebeing carried out for the cover and fuel handling system for removal of unit 3 spent fuel. A planis being carried forward so that the fuel can be removed safely and steadily for units 1 and 2.

Contaminated water treatment
In the early stages after the earthquake, measures were taken against contaminated water
occurring due to the inflow of groundwater, but multiple contaminated water treatment
facilities were put in service after the autumn of 2014, and in May 2015, the treatment of allconcentrated seawater was completed. Furthermore, for the highly-concentrated contaminatedwater remaining in the seawater piping trench of units 2 and 3, the accumulated water wastransferred by August 2015, and the trench filling work was finished.

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During contaminated water treatment, various measures were taken based on the threepolicies: i) to remove sources of contamination, ii) to redirect fresh water far from contaminatedareas, and iii) to retain contaminated water away from leakage to the environment, and it was
confirmed that results, such as reduction in the amount of contaminated water generated
everyday due to the groundwater, were steadily achieved.

Radioactive waste
The waste, such as scattered rubble, generated after the earthquake and tsunami, is temporarily
stored in accordance with the dose rate, and the secondary waste generated duringcontaminated water treatment, such as sludge, slurry and adsorption towers, are stored inappropriate storage facilities taking into account their characteristics.
The amount of waste is reduced by incineration and volume reduction and is consolidated
by storage inside the building.
A “Waste Storage Management Plan” was established in March 2016 (revised in June 2017)
in which the amount of solid waste that would be generated in the next ten years was estimatedand the necessary waste-related facilities were studied. Hereafter, the progress of work will beincorporated into the plan, and the estimation of the generated amount of solid waste will be
reviewed and updated once a year. More details on waste management planning are provided
in Management of Radioactive Waste after a Nuclear Power Plant Accident (NEA, 2016).

Radiation protection
By implementing various measures, effective dose on the site, except for fallout (radioactive
materials that had been released at the time of the Fukushima Daiichi accident and that are stillin the environment), has reached less than 1 mSv/y at the end of 2015. After that, the dose has
been maintained below 1 mSv/y.
Inside the port, the radioactive concentration has been reduced with time since theearthquake, and after the installation of the seaside impermeable wall in October 2015, a further
reduction has been measured.
To control the dispersion of radioactive dust outside the site, measures such as spraying ofanti-scattering agents in work areas are being taken and in addition to the existing monitoring
posts, monitoring has been enhanced by the installation of continuous dust monitors on thesite and on the refuelling floor where there is a possibility of dust dispersion.

Earthquake and tsunami countermeasures
Measures including the application of flexible response to the design basis earthquake groundmotion 600 Gal, and earthquake ground motion for study 900 Gal, are taken.
For outer rise tsunamis, protection was implemented through a temporary seawall.
Measures such as aperture sealing are being taken for the highest-recorded 15.5 m-leveltsunami.
Regarding the beyond design basis tsunami (26.3 m), measures for the treatment of water
accumulated in the buildings will be undertaken in priority.

Improvement in work environment
With the beginning of usage of the main building and contractors building, the working
environment of the employees and the contractor workers has been improved.
The on-site rest areas are also being expanded and added in a planned manner.
With the decontamination on the site progressing in a planned manner, TEPCO’s monitoringdata indicated that the target dose rate (5 μSv/h) was achieved in a wide area on-site except for
the vicinity of units 1 through 4. The controlled areas have been classified according to the statusof contamination and the radiation protection equipment has been optimised.

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STATE OF THE ART OF LONG-TERM MANAGEMENT AND ACTIONS

Summary of main issues and challenges at Fukushima Daiichi

The accident that occurred at TEPCO’s Fukushima Daiichi NPP on 11 March 2011, is different
from TMI-2 and Chernobyl as it evolved into multiple units.

Regulatory framework
After the accident, the regulatory framework was revised to regulate facilities which were
specified as “specified nuclear power facilities” and needed special management towards
recovery or decommissioning. With regard to risk reduction and operational safety, the NRA
confirms the validity of implementation plans and the status of the implemented and ongoingmeasures.

Approaches to risk
The NRA has constructed a risk map, “Measures for Mid-term Risk Reduction at TEPCO’s
Fukushima Daiichi NPS”, to present key issues for risk reduction identified by the NRA among
the various measures being undertaken by TEPCO, and to indicate the progress of measures forrisk reduction.
NDF has run a comprehensive risk assessment composed of a risk reduction strategy and arisk analysis approach.

Management of contaminated water
In order to continuously reduce the risk of water leakage from the buildings and the storage
tanks to the environment caused by some future potential event such as an earthquake,tsunami or ageing, some countermeasures are necessary.
It is necessary to make untiring efforts for risk reduction measures such as replacing
flanged-type water tanks with welded type (high reliable type) ones and suppressing
groundwater ingress into the buildings by a combination of various water management
measures such as groundwater bypass, the sub-drain system, and impermeable walls.

Fuel removal from spent fuel pools
The spent fuel removal work is planned by TEPCO in the order of unit 4, unit 3 and units 1 and 2
taking into account the risks and difficulties.
The spent fuel removal work at unit 4 was completed in December 2014. The knowhow
obtained through the work at unit 4 will be reflected in the safe work at unit 3 and the following
units.
At unit 3, rubble removal and decontamination/shielding installation works have been
performed by November 2017. The fuel removal building cover is now being constructed andthe preparation works for the removal operation and maintenance are ongoing in parallel.
As for unit 1, the building cover, which had been built in 2011 for preventing radiationmaterial from dispersing to the environment, was dismantled.
Rubble removal works are necessary before starting the fuel removal work.
Major challenges of the work at unit 1 are countermeasures to prevent dust dispersion
during the rubble removal works, careful handling of large rubble such as the destroyed
overhead crane over the pool and careful handling of spent fuel bundles.
As for unit 2, the integrity of the reactor building structure remains relatively intact because
no hydrogen explosion occurred there. However the radiation dose rate in the reactor building
including the refuelling floor is still high and the overhead crane and the fuel handling machine
are considered to be useless taking into account the conditions after the accident such as longterm exposure to high temperatures and humidity.
So, TEPCO plans to dismantle the upper floors of the reactor building and will study somebasic concept designs of the fuel removal cover in the future.

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Fuel debris retrieval
The fuel debris removal work mainly consists of the following steps: i) investigations of RPV/PCVinternal conditions, ii) installation of removal work equipment, iii) removal of obstacles toaccess, iv) debris cutting, v) transferring debris into containers and transportation, vi) analysis,
and vii) storage.
Even though various investigations in the PCVs have been conducted and roughdistributions of fuel debris were assumed so far, detailed locations and the conditions have notbeen confirmed as of January 2018.
So, the challenge for the time being is to survey and grasp more detailed and precise
distributions and the conditions of the fuel debris by carrying out further investigations.
In addition to the major challenges of the fuel debris retrieval works themselves, various
safety measures such as securing sub-criticality, cooling, shielding, anti-scattering measuresand appropriate storage will be required as well.

Solid waste management
It is evaluated that a huge amount of waste will be generated in the next decade. So, thecountermeasures such as waste volume reduction and increasing waste storage capacity are
the key issues for future waste management.
With regard to the volume reduction measures, construction of incineration facilities andvolume reduction facilities are planned. And additional waste storage buildings are also planned
for the increase in the storage capacity.

Work environments
The decommissioning work will continue for a long period of time and it is necessary to reduce
the workers’ inconveniences and anxieties through continuous improvements of the work
environments in order to secure a stable workforce in the future.

2.2. Status of long-term post-accident management and actions in NEA member
countries
This Section was established based on answers received from utilities and safety authorities ortheir technical support organisations through a questionnaire that was distributed by the workgroup (provided in Appendix A). The questionnaire objectives were to review planned,
envisaged or existing regulatory requirements, guidance and practices and to identify criticalissues with respect to long-term management and actions for a severe accident. This section
provides a summary of the collected information and is based on contributions from Belgium,
Canada, Czech Republic, Finland, France, Germany, Japan, Korea, Slovenia, Spain, Switzerland,and United States. The summary starts with the summary of the existing regulation concerning
the long-term management of the severe accidents in different countries. The following
subsections summarise the identified critical issues and knowledge related to the long-term
management.

Existing, envisaged or planned regulation for long-term management in NEA member
countries
It is generally considered that the requirements and guidance that apply to emergency
operating procedures/severe accident management guidelines (SAMGs), to emergency
preparedness, to NPP siting, and decommissioning should also apply to post-SAMG long-term
management phase. Following gives a short summary of the state of the regulation for LTM inthe different countries which provided response to the questionnaire.

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Belgium
There are no specific regulatory requirements for the long-term management of severe accidents,
and currently, no regulatory requirements or guidance from the safety authority specifically forlong-term management and actions for a severe accident are planned or anticipated. In thecontext of the action plan in response to the accident in Fukushima Daiichi, the licensee has, on
demand of the Belgian safety authority, developed several measures to manage the accident in
the long term, and to minimise the effects of the radioactivity. These measures include guidanceon radiation protection, setting up a logistic support centre with a mobile unit, confinement of
the radioactivity both as air and water releases, and management of water waste.

CanadaCanadian Nuclear Safety Commission (CNSC) has issued REGDOC-2.3.2 Version 2, AccidentManagement, in September 2015. In this regulatory document, although post-SAMG long-term
accident management is not defined, general requirements and guidance are stipulated without
a time frame. Implicitly, those requirements and guidance apply to the emergency operatingprocedure/SAMG phase as well as the post-SAMG long-term accident management phase. The
following specific goals of accident management shall be achieved for any accident
management phase including long-term accident management: i) minimise the release ofradioactive materials into the environment, ii) achieve a long-term safe stable state of thereactor core or spent fuel storage.

Czech Republic
There are no specific regulatory requirements for long-term accident management. Regulatory
requirements for long-term recovery post-accident phase follow from regulatory degrees onplant siting (SÚJB, 1997), on emergency preparedness (SÚJB, 2002), and from the degree on
decommissioning of a nuclear installation (SÚJB, 2003). Current severe accident management isaimed to recover the damaged unit into controlled safe state, to verify coolability and noncriticality of the degraded fuel and to assess the extent of the damage.

Finland
In the Finnish regulations, most of the severe accident management requirements are not given
for a certain time period, but are applied as long as the condition addressed by the requirement
persists. Monitoring of the plant status and the environmental releases, molten core material
cooling and containment integrity must be ensured under all conditions. As the overall safety
goal, the Nuclear Energy Decree (1988/161) stipulates that a severe accident shall not cause acuteharmful health effects to the population in the vicinity of the nuclear power plant, nor any longterm restrictions on the use of extensive areas of land and water. This requirement is in effectregardless of the accident duration, and it is satisfied if the probability of the total atmosphericrelease of 137Cs exceeding 100 tera-Becquerel (TBq) is extremely small. Regulations ofdecommissioning and handling of nuclear fuel and waste, and the radiation dose limits forthese activities, do not include separate requirements for decommissioning a damaged NPP
after a severe accident. Review of the regulations which has been performed after accidents at
Fukushima Daiichi has not resulted in new requirements for long-term management.

FranceArrêté INB (Nuclear Installation Decree), Article 3.07 states that the radioactive releases out of
the installation as a result of an incident or an accident should be evaluated in a pessimistic
way. To evaluate the radiological consequences of the radioactive releases, the demonstrationof nuclear safety addresses the consequences in the short, medium and long terms, byconsidering the various ways of transfer of radioactive substances to the environment andpeople. This is translated to an objective of maintaining the containment in the long term, andan objective of limiting the long-term effects (permanent relocation, crop losses). The French
Safety Authority (ASN) also imposed a set of requirements in response to the post-Fukushima
Daiichi stress tests with the ambitious objective to reach for generation II reactors the safety
objectives of generation III ones. LTM should benefit from important countermeasures

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implemented or in the process of being implemented by Electricité De France (EDF) in responseto these requirements, e.g. the fast action force (FARN) with its mobile materials or ultimate
electrical (bunkerised diesel generators) and cooling emergency systems.

Germany
There are no specific requirements with regard to the management of long-term severe accident
situation on-site, but there are general requirements in the latest German nuclear rules for the
management of severe accidents and there are documents for off-site management andprotection of public. Preparations and measures are indicated to assess the situation, to alert
the involved parties and to bring the plant back into a “safe and stable state”. In response to theaccidents in Fukushima Daiichi, improvements have been made with regard to on-site accident
management measures, mainly by additional mobile equipment and implementation of SAMG.
Both German safety commissions, the “reactor safety commission” and the “radiation
protection commission (SSK)” have issued in 2014 and 2015, respectively, two reports on
updated requirements on accident management, emergency preparedness, radiation protection,
protection of people in the vicinity of the plant, etc. but with no explicit requirements for longterm management.

Japan
Nuclear Reactor Regulation Law has a chapter about a special nuclear facility that needs special
treatment for nuclear safety (Chapters 64-2). The NRA of Japan published long-term
management requirements to the Fukushima Daiichi NPP based on the above law on
7 November 2012 including requirements for monitoring of RPV, residual heat removal,monitoring of containment atmosphere, keeping RPV and containment inert, defuelling, storageand management of fuels from SFP, securing power sources, designing measures for loss of
power, treatment, storage and management of solid, gas and liquid wastes, radiation protection
measures, emergency measures, design requirements for safety functions, and others.

Korea
There are no existing or planned regulatory requirements or guidance for the long-term
management and actions in Korea. The general requirements and guidance arecomprehensively and implicitly included within a framework of SAMG (implemented before the
Fukushima Daiichi accidents) and accident management plan (to be implemented after the
Fukushima Daiichi accidents) without a differentiation of time frame for a severe accident. The
overall safety goal, regardless of the accident duration, is that the 137Cs release of more than
100 TBq has a probability lower than 10-6/yr.

Slovenia
National Emergency Response Plan for Nuclear and Radiological Accidents is in the process ofrevision. The Slovenian regulations are given in SNSA, 2009a, 2009b, 2010, 2011, 2015, andGovernment of Slovenia, 2010 and are all derived from the WENRA RL (2008). The requirements
will be updated according to new WENRA RL (2014) that include design extended condition (DEC)A and B criteria. Currently, the Slovenian legislation does not distinguish between short-termand long-term management and actions for a severe accident except in the Act ZVISJV, where
the requirements for the mitigation of the consequences of an emergency are set (after the
emergency is brought under control).

Spain
Consejo de Seguridad Nuclear (CSN) has not issued any specific regulatory requirements for longterm management of severe accidents. After the accidents at Fukushima Daiichi, CSN requested
provisions for management of contaminated water. This was addressed by implementing
measures to store contaminated water and to safely transport the water from the site. In addition,
a number of requirements for radiological protection of the workers were requested. These
requirements are valid for any time after the occurrence of a severe accident. There are no
immediate plans to implement new guidance of requirements for long-term management.

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STATE OF THE ART OF LONG-TERM MANAGEMENT AND ACTIONSSwitzerland
The Nuclear Energy Act states that licensees must make provisions against undue releases ofradioactivity and radiation exposure of people. A regulatory requirement specifies that SAMG
needs to cover all phases of a severe accident. In addition, long-term management will benefit
from measures taken by the licensees after the accidents at Fukushima Daiichi: commissioningof an off-site storage facility and storage of additional severe accident management resources
on-site.

United States
Long-term management of accidents in commercial nuclear power plants is the responsibilityof plant owners and licensees. The scope of licensee actions include: i) bringing a plant to astable state; ii) performing site clean-up and decontamination operations; iii) performing
defuelling operation; iv) providing facilities and infrastructure for temporary and permanentstorage of radiological waste forms; and v) providing protection of health and safety of plant
personnel. The US Nuclear Regulatory Commission has the role of providing regulatory
oversight to ensure public health and safety and that of the environment. Administration of thisoversight function is codified in Title 10 of Code of Federal Regulations (10 CFR) including 10 CFRPart 20 (Standards for Protection against Radiation), 10 CFR Part 50 (Licensing power reactors),
10 CFR Part 50.47 (Emergency Plans), 10 CFR Part 51 (Environmental Protection Regulations for
Domestic Licensing and Related Regulatory Functions), 10 CFR Part 71 (Packaging andTransportation of Radioactive Material), and 10 CFR Part 73 (Physical Protection of Plants andMaterials). The lessons learnt from TMI-2 for long-term management have been factored in by
the industry in its operational practices and by the NRC in its reactor oversight programme (ROP).
In light of the Fukushima Daiichi accidents, the ROP is being examined for potential updates inareas related to mitigation of beyond design basis events.

Identified critical safety functions for long-term management
It was generally agreed among the contributors that the critical safety functions important tolong-term accident management are:•to control the reactivity;•

to remove heat from the degraded fuel, and to achieve and to maintain a long-term
stabilised controlled state of the reactor core or spent fuel storage;•

to confine the radioactivity by controlling the operational discharges, as well as limiting
accidental releases;•

to shield against radiation;•

to monitor the safety important parameters to guide operator actions.

Assessment of the plant damaged state and contamination after the event is necessary toallow planning of recovery actions. The main priority of recovery actions on-site is to maintaina controlled state for all units at the affected location during all LTM phases while minimising
the staff occupational dose and minimising additional releases possibly related to recovery
actions.

Identified measures and actions for long-term management
The measures and actions identified by the participants are divided into three subsections:procedures and guidelines, human and organisational resources, and technical resources.

Procedures and guidelines
The first procedural requirement is to ensure that accident management measures cover allmodes of reactor operation including the shutdown states, and that all events that could cause
damage to the fuel in a reactor core, in transport to storage, or stored in a spent fuel pool are
included in the management measures also in the long term.

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The foremost feature of the response by the participants was the recognition that asystematic identification of long-term accident management challenges and measures is
missing, and the necessity for long-term management measures has not been assessed.
Consequently, the following issues were identified:•

Most of the countries do not have any specific guidelines or procedures relevant to thelong-term management of severe accidents.

•

In most countries, such procedures are not foreseen to be implemented.

•

In the cases in which measures are required or foreseen for the long-term management,
the measures have not been properly analysed, and confirmed by verification, evaluation,and independent review.

•

In general, factors and criteria that should be considered for the design andimplementation of measures and actions applicable to long-term accident management
have not been reviewed.

•

The effect of the present SAMGs on the long-term management has not been evaluated.

•

The transition from the accident management activities to accident recovery is notalways defined. Some generic SAMGs include specific criteria to define a controlled
stable state after a severe accident and to leave the SAMGs. These criteria are usuallycore exit temperature, hydrogen content and pressure in containment, radiation level onplant area and in SFP, and the water level in SFP.

Even though there are not many requirements for LTM, recommendations were given forthe planning of LTM by the participants:•

LTM should be done in phases progressing from decontamination to defuelling and thendecommissioning.

•

Each phase needs detailed assessment of the plant damaged state. Data can be collectedin consecutive steps depending on accessibility which will impact planning of recovery
actions.

•

It is recommended to do periodic evaluation of the plant damaged state using monitored
data and numerical simulations of the accident scenario to support LTM planning.

•

Importance of visual observation means was highlighted.

•

Multi-unit arrangement of operated plants should be taken into consideration for longterm post-accident management.

•

Contamination monitoring protocols and locations during the recovery phase should be
defined, and a national strategy regarding solutions for post-accident contamination
should be prepared.

•

Criteria for returning to the evacuated area, and criteria for return to normal from theemergency state should be established.

Human and organisational resources
A long-term support to an NPP may be a real challenge for plant operators and stateorganisations, especially when there are aggravating factors such as multi-unit accidents, anddegraded regional infrastructure.
The following issues were mentioned by the participants to be considered regarding thehuman and organisational resources:

96•

The roles and responsibilities of the personnel responsible for accident management
should be clearly identified and communicated.

•

The personnel involved in managing an accident should have all the necessary
information, procedures and human and material resources to carry out accident
management actions.
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STATE OF THE ART OF LONG-TERM MANAGEMENT AND ACTIONS•Technical support centres for severe accident management and emergency crisis
management teams can be provided in addition to the human resources at the plant.

•

Solidarity between nuclear sites can provide additional personnel to the affected site.

•

External contractors may be employed for specific tasks, e.g. mobile radiologicalmeasuring units, medical units, and additional fire men.

•Training of the personnel for severe accident management should be conducted
regularly, and could also include elements of long-term management, such as the use ofmobile equipment.

External conditions, e.g. extreme heat, cold, high radiation level, pose an additional
challenge to the personnel at the affected site, and should be properly accounted for.

Equipment, infrastructure and instrumentation
After the degraded core becomes stabilised, the long-term management of the accident is
expected to be carried out using the existing long-term plant monitoring capability, restoredequipment, infrastructure and instrumentation, all the available and applicable procedures andguidelines, and available human and organisational resources. Measures required for the longterm accident management functions and goals are generally expected to exist in NPPs. These
measures include plans, procedures and guidelines, equipment and infrastructure, plantmonitoring instrumentation, and human and materiel resources.
All safety systems and components of the plant maybe useful for LTM and should bemaintained operable as much as possible. Accessibility of the plant should be ensured.
Measures and systems to monitor, control and maintain safety critical functions (sub-criticality,
degraded fuel cooling, radioactive release limitation) on the long term should be provided aswell as their support equipment.
Most countries mention the possibility of using specific equipment and measures developedfor emergency phases also in LTM either on-site or off-site, e.g. central protected storage
facilities for mobile equipment. Equipment such as large fire water pumps to supply coolingwater to more than one unit or SFP and diesel generators to supply power to main plant safetysystems are envisaged or already implemented in most countries.
Several systems of special importance to the long-term management were mentioned byparticipants:•

alternative water sources and injection capabilities, including several connections forthem, e.g. reactor, steam generators, containment spray, and containment sumps;•

a dedicated system to reflood the corium and evacuate the residual heat from thecontainment;•a core catcher, sacrificial material, or other system to arrest the concrete erosion by MCCI;•a filtered containment venting system for containment pressure reduction and methods
for long-term passive cooling of the containment;•

passive autocatalytic recombiners for hydrogen control;•

dedicated sampling systems for containment gas phase (dose rate, H2 and O2
concentrations) and sump coolant, and systems that can tackle high activity inventory
samples;•

provisions for safe handling of samples and capability of analysis of high activity
samples;•on-site monitoring of activity release and effluent sampling;•

the storage for highly contaminated liquid and solid wastes;•

proper radiation protection measures, staffing and technical support centres;

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infrastructure needed to support at least three shifts on-site;•

accessibility of the site for emergency material and deliveries from outside.

Some more advanced technologies which were mentioned and considered to be useful for
the LTM were given as:•

water treatment technology including decontamination and potentially desalination;•

remote control monitoring and investigation using robots;•

acoustic monitoring for passive detection of boiling behaviour;•muon tomography for detection of core debris location.

Approaches, methods and tools for long-term management and actions assessment
In general, no established and well-thought approach to deal with LTM issues has beendeveloped. The following tools were identified by the participants to be useful in the LTM:•

The existing integral severe accident codes, such as Accident Source Term Evaluation
Code (ASTEC), MELCOR, and MAAP, can be used to assess the status of the affected plant,
the possible damage to RPV and PCV, degraded fuel characteristics and location, as wellas to support preparation of the fuel recovery actions. The analysis results should becompared with the available plant data regularly.

•

Dedicated codes for certain phenomena, e.g. core-concrete interaction, may provide amore detailed assessment of the phenomenon under inspection than what is possible
with the integral codes.

•

A solid level 2 probabilistic safety assessment associated with detailed SAMG procedures
can provide helpful guidance and understanding appropriate for long-term management,
for instance, the identification of the most probable failure mode of RPV could provide
valuable insights regarding the time windows available to the operators to flood thecontainment.

•

Instrumentation and measurements may be applied to follow the evolution of theaccident on the site.

•

Monitoring radioactivity on-site and in the vicinity of the plant is important in the longterm management phase to detect possible leakage of radioactivity as a result of the coredegradation or as a consequence of the decontamination activities.

Some challenges were identified in application of the existing tools for LTM:•

No tools or methods exist which have been developed and validated for analysis of longterm behaviour of a damaged nuclear power plant.

•

The severe accident codes cannot predict the damage state of the plant in sufficientdetail:•98–

uncertainties in the prediction of the status of the damaged RPV and PCV, degradedfuel characteristics, and location for all conceivable accident scenarios;–

uncertainties in the long-term accident phases such as MCCI termination anddelayed releases of fission products;–

fission product and actinide distribution in the plant cannot be predicted in sufficientdetail;–uncertainties in the location of possible air or liquid leakages of radioactivity.

Absence of data and models to assess risks that are specific for the LTM phases,
e.g. safety system failure risks due to harsh conditions from the accident transient andfrom the long-term operation, risks related to degraded fuel leaching, combustion risks
due to effects of corrosion reactions, potential iodine release from liquids or filteredcontainment venting systems, risks related to fuel recovery actions and water waste
management.
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STATE OF THE ART OF LONG-TERM MANAGEMENT AND ACTIONS•Application of probabilistic safety assessment to all events and all modes of operation isincomplete.

•

Probabilistic safety assessment considered as not necessarily relevant to support longterm management and actions (LTMA) as the focus shifts from plant operation to thesafe performance of recovery and waste management actions.

It is recognised that risk-informed approaches should be used to select the safest approachfor LTMA incorporating observations and analyses results enlightening the plant damaged state.
The plant damaged state will be changing with time with the progress of recovery actions. The
uncertainty in the plant status decreases with time as more data from analysis and observations
become available to assess the accident evolution. The evolution has to be considered for LTMA
planning.

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Chapter 3. Approach to long-term post-accident
management and actions

The main goal of the present chapter is to describe fundamental aspects of long-term
management (LTM) of a severe accident and to provide useful tools to address it. The definitionand scope of LTM, as defined by the group, is given in Section 3.1. The main long-term controlled
state functions as well as their monitoring for a safe LTM are stated in Section 3.2.
Then, an approach is described to identify challenges, issues and risks for LTM, covering alarger diversity of accident scenarios (reactor or spent fuel pool (SFP) severe accidents) in order
not to limit the scope only to Three Mile Island unit 2 (TMI-2), Chernobyl and Fukushima Daiichi
accidents scenarios.
Section 3.3 provides a generic classification of accident scenarios as a function of accident
progression events (e.g. reactor pressure vessel (RPV) and/or containment failed or intact) andlevel and extent of radiological consequences in the plant which should largely influence the
LTM scope. LTM challenges and issues are then identified for each generic scenario.
Section 3.4 presents a structured approach aimed at identifying and categorising main issues
and risks for LTM. Within this approach developed to guide LTM, a risk-informed, plant-specificclassification method of the events, alternate to the one discussed in Section 3.3, is used.
Finally, Section 3.5 presents an action identification and ranking table (AIRT) exercise thatwas conducted to identify challenges, open issues and knowledge and technological gaps
related to main LTM actions: maintaining a coolable configuration and managing confinement,water waste, solid waste and effluent management, site clean-up and decontamination,
defuelling of damaged reactors, damaged fuel/fuel debris and radioactive waste disposal, andspent fuel removal.
It is worth noting that the plant damaged state (PDS) classification (Section 3.3) and AIRT
exercises (Section 3.5) were performed to support the formulation of generic recommendations(Chapter 5).

3.1. Definition and scope
In this report, long-term management 1 of severe accidents refers to accident management actions
implemented after a plant has reached a stabilised and controlled state following a degradedcore accident and up to and including fuel and debris retrieval from the damaged plant,
temporary on-site storage of the fuel and debris, and eventual transportation to off-sitepermanent storage, and removal of spent fuel from the pool and transportation to off-sitepermanent storage. The timing of a plant reaching a stabilised and controlled state depends on,among other things, the accident initiating event, any and all prevention, mitigation andemergency response measures taken to reach a stabilised controlled state and the resulting
plant damaged state and operating conditions.

1.

The definition of “long-term management” in the context of this project is not to be interpreted as the socalled long-term management action traditionally considered by the industry as part of the severeaccident management. The latter definition still relates to bringing a plant to a controlled state following
a degraded core accident, and involves, besides use or restoration of on-site equipment, implementationof mitigation measures employing mobile equipment, among other things, brought from off-site locations.

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Preparation of some long-term management actions may be initiated before reaching astabilised controlled state. However, in a severe accident situation, the priority should logicallygo to actions to terminate the emergency response phase and to stabilise the plant.
A plant is considered in a stabilised and controlled state when i) all parts of the degraded
core, either still in place and/or relocated in-vessel and/or ex-vessel, are coolable and subcritical,
and any stored spent fuel is also coolable and subcritical; ii) dispositions have been taken to
limit as far as possible any further radioactive products dispersion and release to theenvironment; iii) there is no apparent nor urgent risk of combustible gas explosions.
The degraded core, if retained in-vessel, is considered to have reached a coolable
configuration when there is no further hydrogen production from water-metal (clad andstructural materials) interaction; the release of fission products from the degraded core isexceedingly small; there is no risk of corium (molten composition of core materials, structural
materials, and other control materials) rupturing the vessel because of thermal attack (thoughother risks, e.g. loss of integrity because of fragility, can keep being developed further); and thedegraded core remains subcritical. Similarly, the degraded core, if ex-vessel, is considered tohave reached a coolable configuration when there is no further non-condensable gas generation
either from metal oxidation or from molten core-concrete interaction; the release of fission
products from core-concrete interactions is exceedingly small; the ex-vessel core debris isretained in the containment or, in the event of containment breach, environmental release offission products is exceedingly small; and finally, the debris remains subcritical. The spent fuelinventory is considered in a coolable configuration if all the spent fuel rods, degraded or not,are confined in the pool without the risk of a runaway oxidation reaction; there is no significant
production of hydrogen; release of fission products from degraded spent fuel is small; and spent
fuel remains subcritical.
Accident management actions for long-term management of severe accidents comprise theactions taken to: i) ensure the plant remains in a stabilised controlled state as defined above tillit reaches a long-term safe stable state; ii) prevent and mitigate unintended on-site and off-site
consequences such as uncontrolled releases that could result from actions to maintain the plantin a stabilised controlled state as well as recovery actions; and iii) prepare and complete fueland debris retrieval ensuring that there are no unintended consequences during the retrieval
process and transfer of the same as well as spent fuel to temporary on-site and permanent offsite storage facilities.
Generic long-term on-site accident management actions include: i) maintaining andmonitoring a controlled state; ii) cleaning and decontamination of work environment; iii) cleanup of liquid and solid wastes; iv) on-site processing and temporary storage of wastes; and v) fueland debris retrieval, and transfer to temporary on-site storage facilities, and eventual transfer topermanent off-site facilities; and vi) radiological protection of plant personnel and emergency
management. Likewise, generic off-site actions, not discussed in this report, include:i) monitoring and controlling effluent discharge; ii) monitoring environmental contamination
and airborne activities; iii) decontamination of land mass and water bodies; iv) radiologicalprotection of population; v) preparation for rehabilitation and food-chain recovery; andvi) preparation for re-establishing commerce and other normal civic functions.
The on-site and off-site actions related to radiological protection and public health areconsistent with those defined in various International Commission on Radiological Protection(ICRP) documents as well as various Committee for Radiological Protection and Public Health(CRPPH) documents. The long-term accident management phase defined in the current report
maps to the recovery phase in the ICRP documents. It is characterised by planned or existing
exposure situations and recovery actions in areas contaminated during early and intermediatephases of an accident. Note the topic of radiological protection of general population is notaddressed further within the scope of the current report as the subject is more suitable for the
CRPPH mission. Also, the topics of rehabilitation, re-establishment of commerce, and othernormal civic functions are not considered in the scope of this report.
It is important to emphasise that the precise nature and scope of long-term accident
management actions are heavily dictated by the entry conditions to long-term which, as alreadypointed out above, depend, among other factors, on PDSs. For example, the PDS at TMI-2 reactor(i.e. severely degraded core but no failure of reactor pressure vessel) charted the course of long-102

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APPROACH TO LONG-TERM POST-ACCIDENT MANAGEMENT AND ACTIONS

term accident management actions in a certain way. In contrast, the accidents at the Chernobyl
unit 4 and Fukushima Daiichi units 1, 2 and 3 resulted in PDSs that are far more challenging thanthat at TMI-2 and consequently, charted a different course of specific long-term management
actions.
It is also important to emphasise that the PDS when entering long-term management depends
on the preventive and mitigative actions taken during the short and intermediate terms ofaccident management to stabilise a plant. Thus, it is imperative to have a systematic analysis ofentry PDSs. In Section 3.3, a pragmatic classification for entry plant damaged states is proposed.
The classification is intended to be as generic as possible and highlights how such classification
impacts recovery issues and long-term management actions. A second complementary approach,more detailed and systematic and based on the foundation of level 2 probabilistic safetyassessment, is also provided in Section 3.4. This second approach not only complements the firstone but also provides a sound technical basis to confirm the pragmatic classification.

3.2. Long-term controlled state functions and monitoring
LTM comprises a set of different goals which may be arranged under three categories:•

maintaining the damaged plant in a stabilised controlled state;•decontamination and waste management;•

defuelling.

Each of these LTM top goals requires different technical and organisational approaches.
Goals aimed at maintaining a controlled state for the damaged plant are called functions.
This section is focused on the stabilised controlled state functions which are necessary to
manage safely and successfully the full spectrum of long-term actions.
LTM functions aim at maintaining a controlled state at all units and SFPs of a site when asevere accident occurred in core and/or in SFP. Successful achievement of these functions relies
on knowledge of the PDS plus continuous monitoring of critical parameters as an essential
precondition.
These top-level functions, each of which is necessary yet not sufficient to reach a controlled
state, concern:•

reactivity control;•

decay heat removal;•

mitigation of radioactive releases;•

prevention of combustible gas explosions.

Ensuring these functions during all LTM actions presents specific challenges partly linked
to the difficulty to establish the plant damaged state (e.g. extent of fuel damage, barriers status).
The severity of the accident (e.g. degraded fuel in and ex-vessel, eventually outside confinement,
significant radioactive releases outside confinement) will also affect strongly the extent ofcountermeasures and actions and increase the complexity to comply with these functions.
Finally, depending on the focus of the LTM actions (e.g. maintaining a controlled state in thedamaged plant, decontaminating, defuelling, waste processing), complying with the functions
will require different technical and organisational approaches.
A brief introduction to these functions is provided in Section 3.2, where the fundamentalmeans to accomplish them, together with highlighting of critical aspects, are presented.
Section 3.2 also gives an overview on monitoring of parameters needed to meet these functions.

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APPROACH TO LONG-TERM POST-ACCIDENT MANAGEMENT AND ACTIONSLong-term management functions to maintain a controlled state
Reactivity control
Assessing risks of criticality in various configurations (in-vessel, ex-vessel, in SFP, during fuelretrieval, waste processing) is of special importance to implement, if necessary, propercountermeasures (e.g. use of borated water in some cases, use of absorber materials in others).
Fuel type should be considered in the risk evaluation (enriched fuel, mixed oxide fuel and,eventually, new fuels [ATF]). Some configurations may present more risks than others,
e.g.: in-vessel when control rod or plates degrade and relocate, in-vessel when fuel isfragmented forming porous debris beds, ex-vessel after long time when solidified corium has
fragmented under radiation-induced damages.

Decay heat removal
Damaged fuel/fuel debris cooling on the LT may be provided by active or passive cooling using
liquid or air coolant, and by means of a single or two separated coolant circuits.
Several support systems to accomplish this function are needed, namely a reliable andaccessible – even under the presence of certain radiation levels – heat transfer mechanical andcontrol equipment and ultimate heat sink.

Mitigation of radioactive releases
Limitation of gaseous and liquid radioactive releases entails a wide spectrum of sub-functions
mainly dealing with the preservation of defence in depth barriers:•

Prevention of further RPV deterioration. If in-vessel melt retention (IVMR) succeeded –whether through in-vessel or ex-vessel flooding-, the remaining strength of the RPV after
the accident, if it had been weakened by fuel relocation to the lower plenum, may not beknown implying risk of failure at any time during LT. The mechanical resistance of thesteam generator tubes, hot legs and main steam lines might also constitute an issue sincehigh pressure and temperature differences during the course of the severe accident (byflowing of overheated gases) may have significantly weaken high sensitive componentsof the reactor coolant system (RCS).

•Containment pressure control (avoiding under and overpressure).

•

Identification of release paths (including that after molten corium-concrete interaction[MCCI]).

Prevention of combustible gas explosions
Even if decay heat removal is achieved and radioactive releases are kept below certain threshold,the existence of potential flammable clouds whether inside containment, SFP building (ifapplicable) or any other building, should be identified and duly measures taken to avoid risk of
explosion.

Monitoring quantities needed for LTM controlled state functions
Monitoring, control bands and trends of the safety important parameters should provide
information about the state of key functions described above, together with the characterisationof the existing and evolving PDS.
The facility operator should, as far as achievable, monitor and control the following
parameters (either through installed or portable/movable equipment and instrumentation,
availability depending on physical plant conditions) to ensure that the stabilising functions are
met:

104

•

dose maps and environmental conditions to allow manipulating critical equipment;•

RPV water level and temperature;

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RPV pressure;•

RPV/RCS water injection flow rate;•

for pressurised water reactor (PWRs), steam generator pressure and water level (if RCS
integrity preserved);•

containment water injection flow rate and flooded level;•

containment temperature (in dry well and wet well for boiling water reactors [BWRs]);•

containment pressure;•containment combustible gas concentrations (hydrogen, carbon monoxide, oxygen);•containment 135Xe gas concentration (detection of criticality events);•

containment radiation level;•

for PWRs, auxiliary building temperature and radiation level;•

for PWRs, combustible gas concentrations in buildings connected to the containment;•for BWRs, secondary containment temperatures, radiation levels, water levels, integrity,hydrogen concentrations;•spent fuel pool water level;•

spent fuel pool temperature;•

water chemistry of reactor and SFP;•

radioactivity in waters around the facility.

3.3. Possible accident/plant damaged states classification for long-term management
In this section, the objective is to propose a classification that is as generic as possible fornuclear power plant (NPP) damaged states 2 in terms of their different impact on LTM andrecovery issues and actions. It should serve to identify main LTM and recovery technical issues
and actions as a function of main type of stabilised plant damaged states following a severe
accident.
The goal is not to discuss here systematic and methodological step approaches that could
provide additional information but would probably require some development since they were
not applied to LTM up to this date. Such approaches are discussed in some detail in Section 3.4.
Further, human, organisation and relations to the public aspects are not discussed here even
if their importance has been recognised.
Also, it is acknowledged that the proposed classification uses a very simplified approach forthe building of operational plans for LTM and recovery actions. The establishment ofoperational plans for LTM and recovery actions would have to consider in much more detail:2.

•

plant and site specificities, availability of in-site and off-site resources (installed andportable systems, equipment and instrumentation) and capability to install them;•possible aggravating factors (e.g. strongly damaged infrastructures, multi-units accidents,etc.);•

progress with time in the diagnosis of the situation knowing that it may be highly
challenging depending on the accident severity and available investigation tools to know
about the degraded situation (e.g. damaged fuel distribution in the vessel and thecontainment, containment leak paths);

Plant damaged states discussed here are different from traditional PDS classification that is used inlevel 2 probabilistic safety assessment.

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the evolution of the situation with time (e.g. advancement of recovery actions, status of
mitigating systems, possible cliff-edge effects on structures and systems, change indegraded fuel configuration, etc.);•

identification and prioritisation of risks challenging the damaged installation safety;•

the organisation of various actors of LTM.

The TMI-2, Chernobyl and Fukushima Daiichi accidents experience (Section 2.1) should help
providing some general recommendations related to LTM as discussed further in Chapter 5.
The classification considers as first main criteria the extent of radioactive releases coveringsituations without and with containment failure and as second main criteria the type of
accident (with or without RPV failure, with or without containment failure considering mainmodes of containment failures based on the Rasmussen classification [Rasmussen et al., 1975],
SFP accidents). A summary of main discussed issues is provided in Table 3.1.

Situations with very limited release 3 to the environment during the emergency phase
Plant damaged state 1 (PDS-1): severe accident with intact RPV and containment
The accident results in-core degradation with partially degraded fuel assemblies; some debris
formation and some fuel melting have occurred. Core cooling has been recovered before
extended fuel melting and relocation could induce vessel failure. There is very small
radioactivity release to the environment (less or comparable to limits for design base accidents).
The plant state may be described by:•active or passive cooling of degraded core in-vessel has been reached;•partially degraded fuel assemblies present mechanical fragilities, presence of debris andsolidified corium in the vessel;•

contaminated water circulate in RCS with forced or natural circulation and lixiviate
exposed degraded fuel;•H2 combustion in the containment due to hydrogen production by cladding oxidation
has been prevented (by dedicated severe accident management measures, e.g. passive
autocatalytic recombiners or igniters, inerting);•

the containment building is contaminated. Some contaminated water may be present incircuits and tanks that were connected to the sump and/or RCS during the accident.

For LTM, despite preservation of RCS and containment integrity, such a situation is alreadyrather challenging for ensuring main safety functions:•maintaining degraded core cooling in-vessel and avoid any re-criticality on the LT
considering the issue of damaged RPV resistance on the LT;•

avoid any further containment leakage and radioactive release on the LT;•

manage hydrogen in RCS, containment and circuits containing radioactive liquids on the
LT (produced by radiolysis and corrosion reactions).

Generally, the systems used to provide the required functions such as cooling the coriumand evacuating the residual power out of the containment are not qualified under severe
accident (SA) conditions and are often degraded. It is therefore important to make thesefunctions reliable by restoring the failing trains of these systems or by providing mobile means
capable of replacing the means in service in the event of failure of the latter. This is valid for all
the PDS situations discussed in the present section. We however consider that in the PDS-1 thestabilised state was established by the recovery of plant cooling systems (provided by originalor enhanced design).

3.

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Limited to the level of design basis leakages.
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Also, the challenge will be to determine what will be the best options for recovery actions
in such a situation knowing that one will have to face recovery of partially degraded fuelassemblies with some fragility (which may be a challenge for handling) of fuel debris and ofsolidified corium depending on the extent of core degradation, recovery and treatment of highly
contaminated waters from RCS and containment. These recovery actions will have to be
conducted with a highly contaminated environment in the containment and will have to be
conducted minimising dispersion of active material outside containment.
Learnings from the LTM of the TMI-2 accident are relevant for such situations (Section 2.1).
At TMI-2, in addition to the highly contaminated containment, some contamination was alsotransferred to the fuel handling and auxiliary buildings. The approach used shows that firstrecovery actions dealt with removal or stabilisation of contamination to maintain access andoperability of systems used to ensure main safety functions. Then, removal or stabilisation ofcontamination of large areas of the site’s building started in the fuel handling building and theauxiliary building, then in the containment to prepare access to the RPV and defuelling
operations. During this phase the main issues were:•

The decontamination of highly contaminated non-coated concrete floors and walls
where contamination penetrated at variable depth. In particular, the extremely
contaminated building interior required the removal of the concrete top layer.

•

The treatment of large amounts of contaminated waters with in particular theelimination of Cs and Sr fission products.

•

The monitoring of radiation levels and plant state.

These actions required the development of specific tools, equipment, facilities and technics.
The next recovery actions dealt with degraded fuel and debris retrieval from the reactor
vessel. Planning, development of specific equipment and tools and training of operators forthese actions could only be designed after sufficient knowledge of the degraded configurationin the vessel was obtained and risks and challenges associated to damaged fuel retrieval wereidentified and analysed. These operations were particularly challenging but were conductedsafely and successfully at TMI-2.
The other challenging issue at TMI-2 was the radioactive waste management since theaccident created unique radiological waste characteristics and some waste did not fit into
established regulatory waste classification categories for transportation and disposal but thiswas also managed successfully.

Situations with limited or controlled releases to the site and environment during the
emergency phase
PDS-2: SA with failed RPV and containment maintained without venting
The main difference with respect to the previous situation is that fuel assemblies degradation
went further with large fuel melting. Degraded core cooling is recovered too late to avoid lower
head vessel rupture and debris and corium have spread in the containment building. RCS
depressurisation has been obtained either by RCS discharge or has resulted from the vessel
rupture (without containment failure due, e.g. from energetic direct containment heating). The
state is described by:•active or passive cooling of the degraded core in-vessel and of the debris and corium exvessel have been reached through fixed or mobile equipment;

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some partially degraded fuel assemblies debris and solidified corium can remain in thecore 4 (depending on e.g.: radial core power profile, the extent of core degradation, RPV
breach location and extent and pressure at vessel rupture);•debris and solidified corium are present in the containment, mainly in the reactor pit;•

the RPV failure under high pressure should be eliminated by severe accident
management, still there is residual risk of a high pressure melt ejection scenario (which
would lead to dispersed corium in the containment);•

contaminated water circulate through the breached vessel and lixiviate some exposed
degraded fuel;•contaminated water cool the corium and debris ex-vessel and lixiviate them;•H2 and CO (from oxidation and MCCI) combustion risk has been supressed by dedicatedsevere accident management measures;•pressure build-up in the containment has been managed by heat evacuation meansother than containment venting system;•

the containment building is contaminated, containment leakage might be increased but
still sufficient to allow termination of uncontrolled release by reduction of thecontainment overpressure.

For LTM, despite preservation of containment, this state is more challenging than state 1for:•maintaining degraded fuel cooling in and ex-vessel on the LT with increasing risks ofclogging or failure of cooling water loops and/or avoiding any re-criticality on the LT;•

avoiding any containment leakage, including that from any additional loop that would
be required for ex-vessel corium cooling and containment heat evacuation;•

managing hydrogen production (by radiolysis and eventual corrosion reactions).

The degraded fuel and debris retrieval will be more challenging than in PDS-1 with larger
dispersion in the containment and possible accumulation in areas where accessibility is very
limited. Again, planning, development of specific equipment and tools and training of operators
for these actions should only be designed after sufficient knowledge of the degradedconfiguration in the vessel and the containment is obtained and risks and challenges associatedto damaged fuel retrieval are identified and analysed. These operations are expected to beparticularly challenging as illustrated by the situations at the damaged units at Fukushima
Daiichi.
The other challenging issue will be the radioactive waste management considering higher
volumes of liquid and solid wastes than for PDS-1.

PDS-3: SA with failed RPV and limited or controlled releases (leakage of the containment or
filtered venting)
The main difference with respect to the PDS-2 is that limited atmospheric radioactive releases
have occurred and may affect the site accessibility. Such limited radioactive release may occur
due to intended containment venting or leakage by containment overpressure. The state may
be described as follows:
•

4.

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active or passive cooling of the degraded core in-vessel and of the debris and corium exvessel have been reached through fixed or mobile equipment;

Some scenarios may result in a nearly complete transfer of degraded fuel to the containment notably
when large fuel melting has occurred in the RPV and when the RPV failure favours melt transfer to thecontainment, see e.g. the situation at the Fukushima Daiichi unit 1. High pressure melt ejection
scenarios are also expected to result in large corium transfer to the containment.
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some partially degraded fuel assemblies debris and solidified corium can remain invessel;•

debris and solidified corium are present in the containment;•

highly contaminated water circulate through the breached vessel with forced or natural
circulation and lixiviate some exposed degraded fuel (large volumes may be required dueto the breached vessel);•highly contaminated water cool the corium and debris ex-vessel and lixiviate them;•H2 combustion risk due to hydrogen produced by cladding oxidation and by MCCI has
been suppressed by dedicated severe accident management measures;•pressure in the containment has been managed by heat evacuation means orcontainment venting system (CVS);•

the containment building is highly contaminated.

For LTM and recovery actions, such a state presents similar challenges like PDS-2 with added
constraints resulting from the radioactive release from the containment.

Situations with significant uncontrolled releases to the environment during the emergency
phase
We assume here that all preventive measures and actions provisioned to avoid such severe
situations have failed.

PDS-4: SA with failed RPV and failed containment
The main difference with respect to the previous situations is that containment has either failedby over pressurisation (progressive pressure increase resulting from heat-up and condensablegas build-up or rapid pressure increase resulting from energetic events [direct containment
heating], steam explosion, hydrogen combustion) or due to failure of containment isolation. The
state is described by:•active or passive cooling of the degraded core in-vessel and of the debris and corium exvessel have been reached through fixed or mobile equipment;•

some partially degraded fuel assemblies, debris and solidified corium can remain invessel;•

debris and solidified corium are present in the containment;•

contaminated water circulate through the breached vessel and lixiviate some exposed
degraded fuel;•contaminated water cool the corium and debris ex-vessel with forced or natural
circulation and lixiviate some exposed degraded fuel;•containment failed;•the containment building and the attached buildings–are contaminated;•

the site is contaminated.

For LTM and recovery actions, such a situation presents much higher challenge thanprevious situations with added constraints resulting from the large on-site contamination andthe failed containment. The emphasis of LTM could be placed first in decreasing dose levels onsite to restore site accessibility and recovering containment functions.

PDS-5: SA with failed RPV and failed containment by MCCI (for some designs)
The main difference with respect to the previous states is that containment has failed by MCCIand that significant radioactive releases occurred through the soil below the reactor building(for most designs). Ex-vessel core cooling, if applicable before the basemat melt-through, has
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been inefficient in avoiding concrete containment basemat melt-through by corium. The
containment integrity is lost and cannot be recovered. The situation is described by:•

Cooling of the corium has occurred in the soil below the containment concrete basemat,
its progression in the soil has stopped. For such a situation, it will be highly challenging
to diagnose the situation and establish with confidence that the corium progression has
stopped. Also, defining an entry to LTM is challenging as the degraded fuel in the soil
can theoretically release by leaching radioactive material for an indefinite period of timeand contaminate groundwaters. Containing contaminated waters would be very
challenging in such a situation.

•

Debris and solidified corium are present in the soil below the containment basemat.

•

Large volumes of highly contaminated waters used to attempt to cool the corium havebeen released through the soil after basemat failure and have contaminatedgroundwaters.

•

The containment building and underlying soil are highly contaminated, the containment
integrity is lost and cannot be easily recovered.

•

When the containment fails at high internal pressure then the release to the atmospheremay occur for certain reactor designs.

For this type of situation, it is rather challenging to specify what could be the best strategies
to cool the corium and reduce radioactive releases, notably liquid releases.
Also the challenge will be to determine what will be the best options for recovery actions insuch a situation knowing that one will have to face soil and groundwater contamination, thenwill have to re-establish some sort of containment, then possibly recover corium in the soil andtreat large volumes of highly contaminated soils and waters which are not easily accessible.

PDS-6: SA with containment bypass (e.g. steam generator tube rupture in PWR and liner meltthrough in BWR Mark I and Mark II containments 5)
The main difference with respect to previous states is that the containment is bypassed andsignificant early atmospheric radioactive release may occur. Unmitigated scenario withcontainment bypass leads to large release, site contamination and limited site accessibility.
The state may be described by:•active or passive cooling of degraded core has been reached;•extend of core damage depends on time of cooling recovery;•

the leakage has been isolated eliminating the containment bypass or the internalpressure was decreased to terminate release;•the site is contaminated.

For LTM, that type of situation is rather challenging for:

5.
6.

110

•

maintaining degraded core cooling in-vessel on the LT, avoid any re-criticality 6;•

avoid any further leakage and radioactive release;•manage hydrogen combustion risk on the release path (with severe accident
management provisions for hydrogen mitigation bypassed).

BWRs liner melt-through could as well be classified in PDS-4.
For BWRs liner melt-through scenarios, re-criticality may not be an issue if corium is outside thecontainment.
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APPROACH TO LONG-TERM POST-ACCIDENT MANAGEMENT AND ACTIONSPDS-7: SFP accident
The main difference with respect to the previous situations is that the accident is occurring in
the SFP. Depending on the reactor design, the SFP is either in the containment or in a specificbuilding outside the containment.
We consider a situation resulting in severe fuel assembly damage and significant
radioactive releases in the environment. With containment, the situation could be lesschallenging with notably reduced contamination on-site but the containment building would
be highly contaminated. The situation is described by:•

sufficient water inventory and level in the pool have been recovered;•active or passive cooling of severely degraded assemblies have been reached;•

the pool contains severely damaged fuel assemblies and fuel debris in water;•

the pool contains highly contaminated water;•

H2 combustion risk due to hydrogen production by cladding oxidation has been
suppressed by severe accident management provisions in the containment (when poolis in the containment).

The situation could be highly challenging due to large on-site and SFP building
contamination and severe fuel degradation. Another challenge will be radioactive releases andwaste volumes minimisations. Eventually, degraded fuel ageing and leaching in SFP may
become issues as cooling will have to be maintained for a long time before final defuelling.
For SFP outside containment the risk of severe fuel degradation should be practicallyeliminated by preventive measures. The same requirement applies to plant operational stateswith open containment and unavailable safety systems (e.g. during refuelling). It should be
mentioned that SFP outside the containment may cause high radiation level and site
accessibility problem just due to water level decrease in the pool.

Aggravating factors
In nearly all the situations described earlier, some aggravating factors may render the LTM andrecovery actions more challenging. These aggravating factors are:•unreliability of equipment, instrumentation and systems (permanent or portable) for
LTM diagnosis and actions;•necessity to use other cooling sources than permanent ones (other external or portablesources) with problems which may arise from the use of raw water during the emergency
phase (e.g. necessary chlorine elimination from wastes) and of larger volumes, the useof added cooling loops, etc.;•

damaged infrastructures resulting from e.g. seism, flooding, explosions, structural
collapses. Risk of structural collapse may pose additional LTM challenge (cf. Chernobyl
accident);•

some site specificity such as hydrology (cf. Fukushima Daiichi with groundwaters
flowing in damaged facilities);•

meteorology during the release phases which may concentrate aerosol deposition onsite (e.g. in the case of heavy precipitations), aggravating the situation from the on-site
point of view;•

multi-unit accidents that may result from external hazards (cf. Fukushima Daiichi).

Summary and conclusion
Using a simplified plant damaged states classification, summarised in Table 3.1, functions andsome issues and challenges associated to LTM have been identified and discussed. They are
rather generic in nature for all considered entry plant damages states for LTM; identified issues

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and challenges generally being of increasing complexity with increasing extent and level of onsite contamination, with increasing damaged fuel dispersion in the facility or eventually in theenvironment and with increasing generated wastes volumes. SA prevention and mitigation
measures aim at reducing releases and contamination and damaged fuel dispersion.
Aggravating factors may however strongly affect the extent of LTM.

Table 3.1. Summary table of simplified classification of SA plant damaged states andcorresponding long-term management functions and challenges
Long-term management functions and challenges
Plant damaged state

Maintain stabilised
controlled state

Recovery anddecontamination

Defuelling

Wastesmanagement

Worker and environment
protection

PDS-1• Core partially degraded• RPV damaged but not
failed• No containment
bypass• Containment intact but
heavily contaminated(gas and liquid releases
from RCS)• Very limited fission
product release outsidecontainment (designbasis leakages)• Avoid re-criticality in- • Remove
vessel (e.g. boron
contamination frominjection)buildings (ifcontainment leaked• Maintain coreduring the accident)
coolability in-vessel
and from the highly(using restored RCS
contaminatedcooling) on the LT
containment (issue:(issue: RPV LT
removal of incrusted
resistance)
contamination)• Maintain the• Recover highly
containment
contaminated waters• Manage hydrogenfor treatment
combustion risk on
the LT (production by • Minimise airborne
and liquid releases
radiolysis andduring cleaning
corrosion) in RCS,
operations
containment and in
all wastemanagement
processes• Establish as much as
possible knowledgeof fuel distribution
and on various forms
in-vessel• Design-specific tools
and procedures for
retrieval considering
potential risks(re-criticality,
exposure dose,
dispersion, etc.) andchallenges(accessibility, fragilityof damaged fuel)• Defuel damaged
reactor core

• Monitor doses on-site• Treat highly
and off-site (incl. incontaminatedwaterscomplex and(particularly βcontaminatedemitters
environment)elimination) and• Check efficiency ofstore processedcleaning operations
waters (issue:• Check adequacy andtritium)
efficiency of wastes• Handle and store
treatment andfuel debris
management• Handle and store
other wastes
forms• Consider H2 risk
due to radiolysis
in wastes• Minimise
generated wastes
volumes

PDS-2 and PDS-3• Core severely
degraded, coolingrecovered too late
• RPV failed at low/highpressure• No containment
bypass• Containment intact but
heavily contaminated(releases from RCS plus
debris and corium from
RPV, possibly
dispersed)• Moderate fission
product release inenvironment (SAleakages, filteredventing)• Avoid re-criticality inand ex-vessel• Maintain debris
coolability on the LT
in and ex-vessel after
cooling is restored(issue: resistance of
fixed/mobile coolingloop)• Minimise radioactive
releases (gas, liquid)
to the environment• Manage hydrogen
combustion risk on
the LT incontainment andconnected volumes
and in all waste
management
processes• Idem to PDS-1 withhigher complexity
due to presence ofcorium/debris incontainment(particular challenge
if fuel is highly
dispersed):• Establish as much as
possible knowledgeof fuel distribution
and on various forms
in and ex-vessel• Design-specific tools
and procedures for
retrieval considering
potential risks andchallenges• Retrieve remaining
damaged fuel inreactor core and incontainment• Idem to PDS-1• Idem to PDS-1 with
with additional
additional complexity
complexity due todue to on-sitepresence ofcontamination andcorium/debris inpresence ofcontainment
corium/debris in(particular
containment
challenge if fuel ishighly dispersed)• Minimisation ofwaste volumes
more challenging

112• Clean contaminatedbuildings and thehighly contaminatedcontainment (seedefuelling)• Treat localised highcontamination
outside containment(e.g. leak paths, ventlines)• Recover highly
contaminated waters
for treatment• Minimise airborne
and liquid releases(particularly
groundwatercontamination)
during cleaning
operations

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Table 3.1. Summary table of simplified classification of SA plant damaged states andcorresponding long-term management functions and challenges (cont’d)Long-term management functions and challenges
Plant damaged state

Maintain stabilised
controlled state

Recovery anddecontamination

Defuelling

PDS-4 and PDS-5

Wastesmanagement

Worker and environment
protection• Idem to PDS-2 and• Idem to PDS-2 and• Idem to PDS-2 and
PDS-3 with additional
PDS-3 with additional
PDS-3 with anchallenges linked tochallenges linked toadditional challengehigh on-site doses:
high on-site doses
if debris and corium(delayed accessibility, • Treat highly
are outside• RPV failed at low/highcooling to becontainment
contaminatedzones
pressuremaintained much(containment failed
on-site (particularly• Containment failed
longer) and in
by MCCI)
challenging for(open hatch, early
addition:
extended• Ageing processes
failure, failure by MCCI)
contamination and• Reduce, as much as
may have to be• Significant andcontainment
feasible, containment
considered for
uncontrolled fission
breached by MCCI)
leakagesmaterial behaviour if
product release indefuelling cannot be• Minimise, as much as • Recover highly
environment
performed in lesscontaminated waters
feasible, radioactive
than about 20 years
for treatment
releases (gas, liquid)
after the accident
to the environment• Minimise, as much as
feasible, airborne and• Manage hydrogenliquid releases during
combustion risk on
cleaning operations
the LT in all volumes
where H2 is produced
or transferred• Idem to PDS-2 and PDS-3• Idem to PDS-2
with additionaland PDS-3 withchallenges linked tohigher complexity
failed containment anddue to failed
high on-site doses
containment
Monitor radioactivity in• Minimisation ofliquid and gas releases towaste volumes
the environment
more challenging

PDS-6• Idem PDS-1• Idem to PDS-1
with additional
challenges linked
to bypassed
containment andtreatment of
releasedcontamination at
bypass• Idem to PDS-4 and PDS-5• Establish knowledgeof damaged
assemblies andretrieval plan

• Idem PDS-1 withlarger amount ofdamaged fuel• Dose monitoring in SFP• Core severely
degraded, coolingrecovered too late

• Core severely
degraded, coolingrecovered too late
• RPV possibly damaged
but not failed (if failed,
see PDS-4)• Containment bypass• Significant anduncontrolled fission
product release inenvironment throughbypassPDS-7• Fuel assembly severely
degraded, coolingrecovered too late
• Full drain-down not
avoided• Partial drain-down tolevel not adequate forfuel assembly cooling
Significant fission
product release to theenvironment• Idem to PDS-1 with• Treat highly
additional challenges
contaminated zones
linked to bypassed
on-site (particularly
containment andthe bypass zone)
high on-site doses• Recover highly(delayed accessibility)
contaminated waters
and in addition:
for treatment (liquid• Isolate, as much as
releases by thefeasible, the bypass
bypass)• Minimise, as much as
feasible, radioactive
releases to theenvironment• Fuel leaching may
become an issue
if cooling has tobe maintained fora long time

• Minimise, as much as
feasible, airborne andliquid releases during
cleaning operations• Manage hydrogen
combustion risk on
the LT in all volumes
where H2 is produced
or transferred• Specific challenges• Treatment ofrelated to high on-site
contaminated watercontamination• Decontaminate SFP(accessibility,
structure andhabitability)
atmosphere• Avoid SFP re-criticality • Minimise airborne
Maintain coolability ofand liquid releases
spent fuelduring cleaning• Prevent hydrogen
explosion in SFP andadjacent structures
Mitigate
environmental release• Specific challenges
related to high on-site
contamination(accessibility,
habitability)

operations• Airborne radioactivity
monitoring• Design-specific tools
and procedures for
retrieval considering
potential risks andchallenges (fragilityof damaged fuel)• Clean-up and storageof damaged fuel• Handling and storageof other waste forms

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3.4. Methods and tools for risks and issues ranking for long-term management
Introduction to a structured approach to risks ranking and issues identification
Risk-informed decisions play a significant role in many nuclear safety applications.
Identification of the relative contribution of each piece of the system to plant safety response
has been proven to be relevant to improve system performance by orienting efforts in safety
arrangements.
Risk identification, whether quantitative or qualitative, is carried out through application ofstandard methods fitting well with the available information and nature of the system, that is
to say, type and interrelation of the underlying components.
Contrary to plant normal operation conditions or even accident conditions before coredamage, the long-term phase management of a severe accident addresses many different
systems each of them entailing a set of different components or structures, among which RPV
and containment, but also the areas of the site affected by spread radioactive contamination,solid and liquid waste storage, SFP and SFP building, etc.
Due (i) to the different nature of the systems involved in the LTM, it is not possible to assign
one only risk identification method as it suitability depends, as stressed above, on a fitting
between the method and the nature of the system itself. In addition (ii), the entry conditions to
the long-term phase range among a wide range of possible scenarios.
As a consequence, a comprehensive approach to risk ranking requires as a preliminary step
the identification of the different components involved in the performance of each different
system. Only once such issue identification has been carried out, identification of the available
and most suitable risk ranking tools to each of the components of each different system might
be performed.
This method has been here addressed as structured approach to risk ranking and issues
identification 7.
After briefly detailing the scope and some specific notes on the contents of the proposed
method, the tasks to be followed are afterwards described, among which the need for scenario
categorisation. The main section presents a comprehensive analysis of the main issues to beaddressed under the LTM as the preliminary, fundamental step for risk method identification.
The conducted structuring and arrangement of the different LTM goals and derived tasks is notmandatory in its details and just aims at providing an example of how to perform issue
identification as a precondition for risk method assignation. The section is concluded in bylinking suitable, available methods to each issue identified as gathered in Table 3.2.
The appropriateness of such structured approach for LTM has been recently confirmed by
the recent efforts currently undertaken for LTM of Fukushima Daiichi on following up,
monitoring and reducing risks as briefly described in Section 2.1. The method developed goesvery much in line and shares many aspects in common with the methodology presented
hereafter in this section.
Several preliminary remarks to clarify key aspects in driving identification of risk and issues
ranking methods follow:

7.

114

•

Risks/issues and related ranking/identification methods depend on the scenario
characterisation, i.e. risks/issues are accident sequence and configuration-alignment
dependent.

•

Therefore, a PDS categorisation process should be applied before identifying risks/ issues
and related ranks/methods.

Rather than at a system level, the LTM will be structured upon different LTM goals, where each goalcomprises a set of different, independent activities therefore constituting an independent system.
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APPROACH TO LONG-TERM POST-ACCIDENT MANAGEMENT AND ACTIONS•There is a traced-back relationship linking identification methods with risks and issues,
and risks and issues with PDS categorisation (Fukushima Daiichi challenges largely differ
from those of TMI-2).

•

Even if different analysers can handle different own specific definitions of risk, all ofthem encompasses key elements of frequency and consequences, i.e. potential for things
to change and the magnitude of consequences if they do change (IAEA, 2001).

•

As for the consequence, a reference situation will therefore have to be first taken fromwhich deviations can be measured and, hence, the risk.

Since the current Section only deals with issues identification methods and not issues
identification itself, and such methods depend on the type of issue but not on the LTM PDS, theidentification methods will only be linked with the issue categorisation process.
First subsections will describe the basis for a general structured, traceable, and practical
identification of LTM risks and issues, starting from critical safety functions down to types ofpotentials for reference situations to change, where the latter will constitute the joint withsuitable issues identification methods.

General approach to issues and risks identification
Long-term management of the severe accident covers a wide range of issues and risks along awide range of scenarios. Identification of such scenario-dependent issues and risks relies also
on the identification method itself so that each method’s strengths and limitations can highlightand shadow issues and risks accordingly.
Therefore, in order to provide with practical suggestions for the full spectrum of long-term
management on which method most suitably fits with issues identification and risk ranking, ageneric approach will be applied as shown in Figure 3.1.

Figure 3.1. Generic approach proposed for long-term management
issues identification and risk ranking

Mit. Syst. = mitigation systems.
Source: OECD/NEA (2018).

I. Scenario identification. Severe accidents can differ because of many aspects among which
initiating event, performed operator actions or survivability of safety barriers. To handle suchderived large number of scenarios, categorisation based on aspects leading to different longterm driving actions should be applied. Criteria underlying categorisation might be limited tothose sensitive events and situations influencing long-term critical safety functions and derived
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actions as defined. Each scenario is considered as the single combination of a certain PDS interms of severe accident phenomena having taken place and safety systems configuration tomeet with long-term entry conditions. An example giving suggestions to conduct scenarios
categorisation is discussed later in this section.
II. The second fundamental aspect addressed to suitably identify the most appropriate
method for risks ranking and issues identification consists of performing a structured issue
categorisation wherein scenario identification should be applied. Identification of issues andrisks should be derived from suggested methods application for each category once adapted
according to the scenario identified. An instance giving suggestions to conduct issue
categorisation is discussed later in this section.

Scenario categorisation and analysis for conducting risk assessment and issues ranking
Addressing plant state categorisation contributes to the risk assessment process because of thefollowing reasons:•tasks addressing long-term issues are sequence-specific hence should be derived fromparticular PDS;•

risk ranking and issues identification methods and tools apply to specific accident
conditions;•each severe accident evolution features a different signature in terms of thermodynamic
and radiological characterisation.

A detailed PDS categorisation upon (i) main aspects characterising severe accident signature
and affecting long-term controlled state functions and actions, together with (ii) safety systems
configuration needed to meet with those critical safety functions, helps revealing significant
challenges to the long-term management.
Unlike design basis accidents and – to a lesser extent – the short-term phase of a severe
accident, initial and boundary conditions in the accident long-term phase can widely vary:•plant configuration: multiple mitigating system lineups meeting with long-term
controlled state functions, in addition, with the possibility of using non-conventionaland non-safety systems, on-site but also off-site equipment, fixed but also portable, etc.;•

location of fuel debris: depending on RPV failure mode, significant masses of fuel debris
might be located outside the area below the RPV lower head;•flooding levels both for the primary containment vessel (PCV) and attached buildings to
PCV (including the reactor building for BWRs) that in the very long term might in the end
damage mitigating equipment;•

environmental conditions (temperature, humidity, radiation) in rooms – or open spaces– hosting mitigating systems where human actions are foreseen;•

RPV/PCV physical boundaries, e.g. whether existing breaches communicate with
attached buildings or the environment.

A widely used tool to arrange the entire spectrum of severe accident sequences, currently
available in the majority of NPPs, is level 2 probabilistic risk assessment within which releasecategory classification constitutes one of the mandatory steps of the methodology. Releasecategories are groups of core-damage accident sequences featuring similar source term releasecharacterisation. Underlying criteria in containment event tree design indeed embeds
aforementioned SA-LT categorisation criterion of making distinction upon the three accident
management actions: (action 1) to reach a cooled configuration aimed at achieving a long-term
controlled state, (action 2) to limit radioactive releases (with different source terms: location, typeof failure and magnitude of the release) and (action 3) to switch to defuelling (given that from theaccident categorisation standpoint, defuelling strategies depend on fuel debris distribution).

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APPROACH TO LONG-TERM POST-ACCIDENT MANAGEMENT AND ACTIONSAppendix B provides a simplified example of SA-LT categorisation from level 2 probabilistic
risk assessment release category figures of merit. It should be highlighted that this riskinformed approach to PDS classification differs from the one presented in Section 3.3.

Issue categorisation and analysis for conducting risk assessment and issues ranking
Introduction and main components of a general structured layout approach
In order to generically address issues identification and risks ranking, a structured layout of
long-term scenario category of issues must be undertaken. To help simplify the process andensure the comprehensive nature of the resulting categorisation, such layout should be
developed in a top-down approach stating i) LTM pursued goals, ii) means and actions inmeeting with such goals, iii) reference situation/equipment needed and configuration to
perform such means or actions, and iv) potential types of deviations in failing to maintain suchreference situation. Suitable issues identification and risks ranking tools and methods will thenbe assigned according to the nature of the issue/risk and the type of deviation 8.
Figure 3.2 depicts the general structured layout for issues identification according to theaforementioned components in a top-down approach.

Figure 3.2. General structured layout for issues identification

Source: OECD/NEA (2018).

From a non-analytical view, such approach starts by distinguishing and reducing LTM
objectives into categories of objectives which are independent of one another, i.e. each objective
can be met without partly or fully correlatively meeting with the rest of the objectives. From ananalytical view, each objective stands as a category as long as it constitutes an independent set
of mandatory taken actions/measures, i.e. the carried out actions within each objective should
be different of one another.
Stepping one level down, each action or measure is categorised as an independent meansof accomplishing the overarching objective. At the action/measure-level, independency isestablished by featuring differences regarding some of the following criteria:

8.

•

addressed object which undergoes the action;•

addressed subject who performs the action;

Deviations stand for situations departing from reference situations where conditions are met for
maintaining a controlled state. For instance, a situation where the hydrogen concentration in thecontainment increases above safety threshold limits is a deviation. The reference situations and related
deviation scenarios within each LTM goal discussed in the following sections attempt to cover the most
relevant aspects for LTM but the lists of items are not necessarily fully comprehensive.

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area or component where the action takes place;•

constraints the action is subject to.

Each measure or action is in turn materialised into a specific set of supporting systems,
configuration, scenarios, etc., each one constituting a so-called reference situation. Finally, thelowest level corresponds to the different causes making each action to depart from eachreference situation.

Long-term management goals
According to the LTM definition (Section 3.1), accident management actions for long-term
management are taken to reach the top goals of i) maintaining a stabilised controlled state;ii) prevent and mitigate uncontrolled releases; and iii) prepare and perform defuelling. 9 Each ofthese three top goals comprises in turn a set of related measures or actions. Since meeting withseveral of these goals depend on similar actions, LTM goals may be reclassified as depicted in
Figure 3.3.

Figure 3.3. LTM goals for the risks and issues identification
as derived from the LTM definition

Source: OECD/NEA (2018).

Following the LTM definition, the first goal of ensuring a stabilised controlled state can be
broken down into ensuring coolability, mitigating radioactive releases, avoiding explosions andpreventing any further defence in depth barrier degradation regarding their state at the time ofentering into LTM phase. In turn, coolability condition is broken down into avoiding generation
and accumulation of combustible gases, fission product releases and re-criticality either invessel, ex-vessel and in the spent fuel pool.

9.

118

The followed structure is arranged upon LTM goals as specified within the definition in Section 3.1. An
alternative, more exhaustive and detailed structure might rely on the generic long-term on-site and offsite actions as indicated also in the definition.
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APPROACH TO LONG-TERM POST-ACCIDENT MANAGEMENT AND ACTIONS

The resultant simplifying transformations shown in Figure 3.3 lead to the following
combined top goals:•

Heat removal: fission product releases and flammable gas generation are kept undercontrol in case of effective cooling of the fuel debris no matter their location. Even if thefuel debris is cooled down effectively, further failure of RPV and containment could occur
on the LT, potentially challenging fuel debris heat removal. Thus, we opted for including
prevention of further challenges to barriers under this current function/goal 10.

•

Decontamination and mitigation of radioactive releases: actions involved in meetingwith this objective can substantially differ from the rest of long-term actions sincecooling can be achieved and barriers can maintain their status and at the same time
significant leakages can result in radioactive releases to the outside environment or
attached buildings (provided containment leak tightness has been lost).

•

Prevention of explosion: risk of combustion of gases generated during the emergency
phase has been as far as achievable suppressed. Moreover, since combustible gases
might be further generated on the LT even though at very low rates once effective fueldebris heat removal is achieved, devoted actions should be considered in the LT in thisrespect.

•

Defuelling preparation and completion: actions looking at removing fuel debris fromcontainment, RPV and SFP belong to a separate set of actions.

With respect to the sub-criticality top-level issue, it initially addresses a goal different thanremoving heat from the affected buildings so it should be classified accordingly. Sub-criticality
is ensured through monitoring and backup system aimed at keeping the fission reactions as low
as possible, for instance by rapidly injecting boric acid.
Nonetheless, since both actions share their reference situations hence related deviations
with some of the actions subsumed into the other categorised goals, sub-criticality as a top
objective will not be further analysed. As an instance dealing with the boric acid injection, thereference situation will be a system injecting boron solution into the compartment where fueldebris are located, e.g. RPV or reactor pedestal, which fully coincides with the maintaining LTM
entry conditions reference situation within the plant performance action within the corium heatremoval top goal. And the same is valid for the monitoring action which can be found under
different goals.

Fuel debris heat removal
Fuel debris heat removal top-level function/goal comprises the following fundamental actions
which can be figured out as different defence in depth levels as depicted in Figure 3.4:•

Plant performance: plant configuration is likely drastically modified so that systems incharge of accomplishing with the long-term controlled state functions should perform
well along the LTM phase.

•

Working conditions: actions performed on working mitigating systems, recovery actions,managing actions conducting the process, etc.

•

Ongoing severe accident evolution: synchronically, entry conditions do not constitute thefull list of relevant magnitudes to ensure keeping with a stabilised and controlled statethrough the entire long-term phase of the accident. Diachronically, entry-condition
related magnitudes can evolve along time, whether in a slow degraded process or by
abruptly coming across so-called cliff-edge effects so that critical quantities, so far kept
below safety levels, would depart from a stabilised and controlled state.

10. Sub-criticality issues will be subsumed under the fuel debris heat removal category because related
actions share reference situations and derived deviations.

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New accident conditions: measures against the onset of initiating events and evolution
potentially leading to the loss of corium heat removal or any other long-term controlled
state function capacity should be carried out moreover when plant configuration
performance will likely lack of preventive safety measures other than working mitigating
systems themselves 11.

Figure 3.4. Fundamental actions affecting the long-term management
goal of fuel debris heat removal

Fuel debris heat removal: Plant performance
Source: OECD/NEA (2018).

Brief description of the issue
Mitigating systems meeting long-term controlled state functions are subject to fail intrinsicallyeither because of internal component failures or so-called cliff-edge effects related to backup
systems depletion. According to IAEA (2016), a cliff-edge effect “in a nuclear power plant is an
instance of severely abnormal plant behaviour caused by an abrupt transition from one plant status toanother following a small deviation in a plant parameter, and thus a sudden large variation in-plantconditions in response to a small variation in an input”.
Provisions for maintenance actions dealing with depletion are advisable in order to facespecific long-term supplying type of problems since mission times usually managed in normal
and accident conditions do not go far beyond 72-96 hours. Recovery actions dealing with internal
failures should also be preaddressed and included as preventive measures to avoid dramatic
equipment replacement gaps and minimise the corresponding corrective human actions.
Within the action of plant performance, ensuring that the working systems perform well,together with the recovery actions and the status of available guidelines to drive the related
processes are analysed.

Reference situations and deviation types
Overall plant performance in meeting with corium heat removal can be analysed through thefollowing reference situations each of which subject to specific potential deviations:•

Maintaining LTM entry conditions: Mitigating systems ensuring plant stabilisation
should be kept under control.

11. The issue on new accident conditions actually affects also the “mitigation of releases” LTM goal insofar
existing structural material, e.g. deposits of stored contaminated water, might be challenged becauseof the onset of an accident. For the sake of simplification, it would only be addressed within the “fueldebris heat removal” LTM goal.

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APPROACH TO LONG-TERM POST-ACCIDENT MANAGEMENT AND ACTIONS–

Mitigating systems alignment: Intrinsic deviations associated to mitigating systemalignment meeting with fuel debris heat removal stem from internal failures thatcan either be mechanical, electrical or hydraulic of the safety system itself or
supporting systems.

–

Mitigating systems can also deviate from normal performance because of occurring
conditions arising from slowly-increasing performance degradation going beyond
equipment design technical specifications as a result of environmental and nonenvironmental burdens typical of unfolded severe accidents in the long term. Even
though system suppliers might confirm that equipment performs well under certain
test profile conditions through equipment qualification testing or survivability
analysis, high uncertainty might arise when severe accident spans over long periodsof time. For instance, sump blockage (highly reducing the available net positive
suction head), heat transfer degradation by aerosol deposition (for instance, in thecontainment heat removal system cooling batteries), or mechanical fatigue (forinstance, in the filtered venting system), might cause losing some of the safety
critical functions such as containment overpressure or MCCI mitigation for somespecific NPP designs. Ageing as a cause for conditions of the facilities getting
moderately worse in the future also falls under this deviation type.

–

Actions dealing with depletion: In dealing with depletion actions, as listed in Safetyof Nuclear Power Plants: Design, Specific Safety Requirements (IAEA, 2016), the following
examples illustrate the possible sources of internal cliff-edges:

–

•

–

depletion of batteries in blackout conditions;–depletion of water sources for primary circuit for loss-of-coolant accidents whenrecirculation is not possible;–

depletion of water sources for secondary make-up in case of unavailability ofclosed circuit for heat removal through steam generators (this is applicablemainly for PWRs for transient events and events with secondary side line
ruptures or secondary side valves failing to reclose).

Certain short-term actions aimed at driving the plant to a stabilised and controlled
state might negatively impact on the long-term phase (NEA, 2000). Corrosion
problems derived from acid solutions present in the containment and coming fromchemical additives, high dose rates in the auxiliary buildings as a consequence ofcontainment hard venting, or thermal shocks because of high injection rates of cold
water into the RPV, are instances of actions that should be analysed in advance tofind a compromise solution between the short-term and long-term management.

Recovery actions on equipment that can play a safety role in the long term.
–

Instrumentation capability to monitor main parameters evolution. Instrumentation
availability deserves special attention since almost all actions and overarching
functions/goals rely on instrumentation reading and feedback. In order to transit to
LTM phase, a minimum set of instrumentation to check compliance with LTM entry
conditions must be available, e.g. flammable gas concentration monitoring system,
containment and RPV pressure and temperature instrumentation, etc. However, as
noted above, the plant can depart from a stabilised and controlled state whether
synchronically or diachronically. In order to track such safety challenge,
instrumentation should be ensured beyond minimal requirements in meeting with
LTM entry conditions.

–Standard safety equipment: recovery actions focused on standard safety systems
might contribute to give support to fuel debris heat removal. Allocated resources,
planning of activities, people involved, etc., will dedicatedly address this topic. With
respect to recovered standard safety equipment not being used in achieving safe
stable state condition, special attention should duly take to ensure that no newchallenging issues arise from such equipment as for instance damaged piping whose
integrity has been lost or radioactive products spreading to areas hosting that

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equipment. Therefore, deviations in this topic stem from worsening the current
situation by putting in place standard safety equipment to reinforce fuel debris heatremoval capabilities.
•

Back-up guidelines supporting operator actions.
–

Existing procedures and guidelines, e.g. severe accident management guidelines(SAMGs). Once LTM entry conditions are met, current existing SAMGs, in the mostfavourable case, will only give instructions to follow-up generic actions. Therefore,challenges in this topic deal with gaps in managing and conducting corium heatremoval related actions as specified above.

Fuel debris heat removal: Working conditions

Brief description of the issue
Human actions have traditionally played a very significant role as risk contributors to safety
figures of merit, staying usually at the top of minimum cut sets contribution to core damage
frequency in level 1 probabilistic risk assessment applications. Since SA-LT human actions largely
differ from those characterising standard normal and abnormal operational conditions(i.e. normal operation and accidents with baseline plant equipment and configuration), it is highly
advisable to adequately characterise human actions along the SA-LT phase from the perspectiveof so-called performance shaping factors (PSFs) and equivalent or related formulations.
Reaching a stabilised and controlled state as defined does not ensure that environmental
conditions will long be kept under safe, sustained values – as already mentioned when
distinguishing between synchronic and diachronic reasons to departing from a stabilised andcontrolled state. Variables dealing with meeting LTM entry conditions might not fully satisfy
the entire set of variables affecting long-term human actions, e.g. radiation values, flooding
levels, habitability (humidity and temperature) conditions, or existence of structural obstacles,or they might be satisfied only on a temporary basis.
It is worth noticing that the reason why human performance is analysed under the fueldebris heat removal issue is because the remaining actions, i.e. plant performance, ongoing accident
evolution and new accident conditions, constitute the supporting foundations of the entire LTM
building so that without keeping the reactor under a stabilised and controlled state no further
action such as mitigations of releases, decontamination or defuelling might take place.
Moreover, those actions are not only relevant from a safety standpoint but also because they
are likely the most demanding ones in terms of human error precursor factors such as available
time or stressful environments.

Reference situations and deviation types
Human performance is carried out by means of the following reference situations each of which
subject to specific potential deviations:•

122

Environment conditions: This topic refers to environmental conditions affecting humanaction performance since harsh conditions in terms of high temperature, humidity orradiation will severely impact the success of the human action. Contribution of local
actions in the SA-LT phase will likely take a far-reaching predominant role than standard
normal and accident conditions so that new environments will have to be faced by field
operators. Actions dealing with manipulating equipment, replacement, replenishment,or simply isolating/connecting areas by closing/opening safety gates will require a set ofminimum working conditions to be able to perform that action:–

Challenges to working conditions come from unbearable environments resulting
from the ongoing accident itself or as a consequence of taken actions to reach andmaintain LTM stabilised and controlled state.

–

Physical impediment to get access to specific areas and rooms and freely movethroughout the plant as necessary.

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APPROACH TO LONG-TERM POST-ACCIDENT MANAGEMENT AND ACTIONS•Human action performance: Deviations in human actions taken as a whole fromexpected performance in carrying out fundamental tasks under fuel debris heat removal
deserve dedicated treatment:–

Aside from environmental constraints, human actions can depart from their
expected performance because of so-called categories of omission or commission,under which precursors such as lack of conducting guidance, ambiguous
information, environmental stress, and many other factors are gathered upon.

Fuel debris heat removal, ongoing accident evolution (non-system-related cliff-edge
effects)

Brief description of the issue
Loss of stabilised and controlled state can also be caused by cliff-edge effects leading to sudden
modification of thermal-hydraulic conditions. Prolongation of previously under-control
variables coming from short-term severe accident phase whose natural evolution encounters acliff-edge effect can dramatically overturn its safe time trend.

Reference situations and deviation types
Potential issues coming from ongoing SA evolution affect the following reference situations
each of which subject to specific potential deviations:•

Status of the environment in the primary/secondary system and attached buildings tocontainment.
–•

Prevention of further challenges to barriers.
–Equipment performance can be challenged if the environment conditions go beyond
design specifications. Such modifications might derive from the accident evolution
in the long term and, alongside the containment, it can affect the auxiliary buildings
hosting safety equipment. For instance, for PWR designs provided with passive
autocatalytic recombiners (PARs) as the only available means to cope with flammablegases, if the containment runs out of oxygen, the PARs would not be able to keep thehydrogen and carbon monoxide concentrations under a certain limit. This C-E effectphenomenon might be relevant in case the non-flammability state of thecontainment still relied on PARs (even in the absence of MCCI, some hydrogen andcarbon monoxide might continue being generated which, together with a continuous
steam condensation, might in the end increase the flammable gases concentration).
Progressively lower decay heats resulting in a negative pressure evolution might also
challenge the containment pressure design limits and/or the flammability limits.
Such pressure evolution might cause a decrease in the saturation temperature, andas a consequence, large steam condensation rates that in the end might result inflammable conditions.

Even if the fuel debris has been effectively cooled down, there might be a risk forundergoing leakage in the containment or the RPV. Contributors to this risk mightcome from unexpected pressure increases – for instance, related to continuously
higher flooding levels pressurising a particular compartment –, or from a slow yet
continuous increasing in the flooding level until reaching a specific mitigating
system elevation.

Fuel debris heat removal, new accident conditions

Brief description of the issue
Contrary to standard normal and accident initial and boundary conditions, SA-LT phase ischaracterised by plant configurations where internal initiating events can only be derived frommitigating systems themselves, namely plant complexity degree has substantially been reduced
since the number of plant systems is limited to only those ones in charge of achieving and

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maintaining the long-term controlled state functions (together with their support systems aswell). Therefore, both in case of external and internal events, initiating events will always leadto direct loss of long-term controlled state functions (as there is no other plant working system
performing functions different than safety functions).
Moreover, since the plant might have shifted to an almost entirely new configuration wheremost of previous existing systems to avoid initiating event propagation are no more available,
preventive actions 12 should be accordingly undertaken to replace such systems.
External and internal events should be analysed and weaknesses that could lead to loss oflong-term controlled state functions.

Reference situations and deviation types
Potential issues coming from new accidents affect the following reference situations each ofwhich subject to specific potential deviations:••

External safety barriers:–

Protection against new external hazards such as earthquake, flooding or extreme
temperature should be analysed from the view of portable mitigating systems
directly exposed to outside environments thereby to all type of extreme climate
events without any kind of protection.

–

Consequence on the status of defence in depth barriers by external hazards should
also be analysed.

Internal safety barriers stemming from systems different than mitigating systems:–

As a result of a drastic modification in the available systems, the plant might face anupcoming initiating event in a degraded situation. The lack in the long term of
several standard safety measures operable in normal operation and limited accident
conditions might result in the available structure (safety systems, but also including
existing emergency procedures) falling short to correctly respond to the accident.
Among them, the lack of the engineered safeguards actuation and sequence system,
or of any redundancy should a system failure occur, might significantly contributeto the overall risk of reaching new unstable conditions. Therefore, the identificationof the most vulnerable failures directly propagating to a sudden loss of a top-level
function/goal turns to be a fundamental issue.

Mitigate radioactive releases
Mitigate radioactive releases top-level function/goal comprises the following fundamental
actions:•decrease radioactive gas emissions;•decrease radioactive liquid effluent emissions;•decontamination of highly contaminated floors and walls;•

radiological protection 13.

Mitigate radioactive releases: Gas emissions

Brief description of the issue
LTM entry conditions ensure that generated radioactive gas source term has decreased to
exceedingly small values. Nonetheless, radioactive gas could still be present in containment
and attached buildings so that duly measures should be taken in an attempt of reducing theirreleases to outside environment. Moreover, if containment bypass leakage exists (beyond

12. Preventive regarding the new stabilised situation in the long term taken as a reference.
13. Disregarded here since not fully treated in the current report.

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allowable leakage), or there is a breach in containment integrity or containment isolation
system has failed, and the containment or RPV is under saturated conditions, radioactive gas
will be continuously generated and released with steam.

Reference situations and deviation types
Treatment of radioactive gas releases is carried out through the following reference situations
each of which presents particular deviation types:••

Identification of release locations: Radiation area monitors placed at each potential
source term release point can help pinpointing radiological release locations.
Deterministic and probabilistic analysis conducted in the plant might highlight those
several equipment and containment interface locations more prone to fail and act assource of radioactive release. Actions should therefore duly be taken to check if thosepoints have been challenged, e.g. whether they have undergone mechanical integrity
failure, stuck-open positions, uncovered emissions (non-water-sealed or water-scrubbed
releases), etc.:–

Radiation monitor failure can hinder the process of identification.

–

In case of not having access to the monitor reading, it will also be difficult to establish
a radiological map of the plant detailed enough to identify potential source term
releases.

Filtering of containment gases and radionuclides: Auxiliary buildings filtered ventilation
system together with in-containment filters, e.g. those belonging to filtered containment
venting (FCV), can have a strong impact in decreasing radiation values both inside theplant and released to the outside environment, furthermore increasing habitability:–

slowly decreasing filters performance (reaching threshold performance limits);–

lack of feedback through appropriate available instrumentation to check effectivenessof carried out actions, non-reliable readings.

Mitigate radioactive releases. Liquid effluent emissions

Brief description of the issue
Release of radioactive liquid effluent may occur under the special circumstances of containment
flooding level reaching failed-to-close penetrations, because of containment bypass orcontainment failure.

Reference situations and deviation types
Control and minimisation of liquid effluent emissions are carried out through the following
reference situations each of which presents particular deviation types:••

Separate treatment and storage of high and medium-low contaminated waters:–

stable storage threatened by external hazards;–

radioactive spills derived from logistics issues such as short-term temporary storage
during the first instants after the accident, sufficient storage space or free area toinstall due connections with containment and treatment equipment;–

failures in water treatment and disposal (including fission product control, monitor
and sampling).

Prevention of liquid effluents entering/exiting containment:–

drawbacks from building additional barriers for liquid effluent confinement, both onsite (e.g. by isolating different contaminated areas of auxiliary buildings attached tocontainment) or off-site;
drawbacks from temporary one-through liquid treatment and non-treatment water
released to outside environment.

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Mitigate radioactive releases: Decontamination of highly contaminated floors andwalls

Brief description of the issue
Decontamination efforts pursue maintaining access and operability of plant systems, in the firstplace, and removal of any remaining contamination for long-term storage in the second place.

Reference situations and deviation types
The following reference situations share common deviation types all of them related to issues
stemming from monitoring and control of radiation spreading.
•identification and division of plant contaminated buildings into different levels ofcontamination;•site buildings clean-up: sludge removal and surface decontamination;•

debris and sludge characterisation;•soil and groundwater monitor and decontamination;

Prevention of risk of explosions
Prevention of risk of explosions as top-level function/goal comprises the following fundamental
actions:•monitoring;•clearing;•

removal.

Prevent risk of explosions: Monitoring

Brief description of the issue
One of the LTM entry conditions consists of keeping combustible gas generation underthreshold limits either from metal oxidation or from molten core-concrete interaction.
Therefore, in order to transit from short-term to long-term phase of the severe accident
management, monitoring of combustible gases should be performed.

Reference situations and deviation types
Monitoring of combustible gases is carried out through the following reference situations eachof which presents particular deviation types:••

Analysis, monitoring and filtering of containment combustible gases, especially in thoselocations more prone to hydrogen and carbon monoxide build-up:–

loss of monitoring devices (including non-reliable readings and samplings);–

inaccessibility issues to monitor readings.

Environmental analysis of attached buildings to containment including reactor building,especially in those locations subject to hydrogen and carbon monoxide accumulation:–126

inaccessibility to rooms hosting safety equipment or other areas susceptible toexplosion.

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Brief description of the issue
Even if combustible gas generation has been drastically reduced, flammable clouds may still bepresent in containment and attached buildings, e.g. small confined compartments or eveninside pipes, so that duly measures should be taken in an attempt of preventing risk of
explosions in the long term.

Reference situations and deviation types
Clearing of combustible gases is carried out through the following reference situations each ofwhich presents particular deviation types:••

Venting, whether filtered venting through a chimney or manual venting throughmanually opening of hatches located at top floors of attached buildings to containment,
as well as accessible doors increasing natural circulation thereby flammable gas mixing
and dilution with air:–inaccessibility due to radiation;–

loss of automatic opening system;–

slowly-increasing filters performance degradation (reaching threshold performance
limits).

Switch-on of heating, ventilation, and air conditioning system (or equivalent) to favour
flammable gases dilution with air:–lack of supporting systems (power, actuation);–

inaccessibility to locations from where the system must be activated.

Prevent risk of explosions: Removal

Brief description of the issue
Even if corium has been quenched, flammable gases may be continuously generated for
instance through water radiolysis.

Reference situations and deviation types
Removal of combustible gases is carried out through the following reference situations each ofwhich presents particular deviation types:••

recombiner device performance:–

oxygen depletion would interrupt hydrogen and carbon monoxide oxidation;–

recombiners may only face low flammable gases generation rates.

injection of inert gases:–•

loss of injection of inert gases such as nitrogen may conduct to flammable gas
escalation.

igniters:–

Loss of igniters supporting systems would prevent igniters performance. Moreover,after significant elapsed times along which igniters have not been working, thereby
flammable gases have been building-up, turning on igniters might drive tochallenging explosions.

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venting 14.

Defuelling preparation and completion
Contrary to the other three main areas of activities, no urgent issues should be immediately
addressed related to the topic of preparation and completion of defuelling.
As top-level function/goal, it comprises the following fundamental actions:•collect information from in-vessel/ex-vessel and SFP inspections and observations of thestatus, location and characterisation of fuel debris, including analysis performed withcode simulations;•training of defuelling operators and development of new tools and equipment;•

defuelling and disposal operations;•safety requirements for defuelling.

Preparation for defuelling: Collection of information

Brief description of the issue
First issue to tackle with when planning defuelling activities consists of achieving thoroughunderstanding on the extent of the events in terms of status, location and characterisation of
degraded core, e.g. whether RPV failure occurred, corium relocation to other parts of thecontainment aside from the area placed right below the vessel, etc.

Reference situations and deviation types
Information gathering is carried out through the following reference situations each of which
presents particular deviation types:
•

•

•



available information about the accident follow-up in terms of status, location andcharacterisation (e.g. mass) of the degraded fuel:–ambiguity in operators/technical support centre in diagnosing the accident/
interpreting the available information;–

absence of information on RPV/SFP state and fuel debris distribution andcharacteristics.

inspection by means of dedicated devices such as visual cameras, muon techniques, etc.:–

impossibility of introducing devices into containment/RPV/SFP;–

deficient provided information.

analysis with dedicated severe accident simulation codes:–

discrepancy in the results when using different codes;–

uncertainty in the results whether because of a deficient the state of the art or
because of a lack of information.

Preparation for defuelling: Training

Brief description of the issue
In order to address training on defuelling and disposal, personnel should be trained in using
dedicated new tools and equipment.

14. Already included within the previous clearing action.

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Reference situations and deviation types
Training activities may include simulations in real mock-ups and development and handling ofad hoc tools, e.g. to visually inspect containment, RPV or SFP, or remotely load debris canisters.
No generic deviation type in this action deserving issues identification and risks ranking
methods and tools application is foreseen.

Preparation for defuelling: Defuelling operations

Brief description of the issue
The area of activities focused on defuelling operations puts an end to the long-term
management of the severe accident. It includes a wide series of ad hoc activities performed bytrained personnel and spanning along a large time frame that can last several months.

Reference situations and deviation types
Defuelling operations are carried out through the following reference situations each of which
presents particular deviation types:•

Provide the plant with suitable temporary defuelling platforms:–••

Collection and characterisation of debris samples:–worsening scenario by related handling operations;–

extrapolation feasibility of sample characterisation.

Collect corium as deposited during accident evolution distinguishing between debris
containing degraded fuel and additional debris material such as cladding, structural andcontrol materials:–•

Deviations match those ones already indicated for collection of corium reference
situation.

Transport all collected kind of high activity inside-containment debris, together withaforementioned by-products, to an on-site safe, temporary storage location:–•

Deviations match those ones already indicated for collection of corium reference
situation.

Disposal of solid waste by-products used in the clean-up and treatment of effluentsphase including filters, demineralisation devices, clothing, tools and equipment:–•

Hazards involved in related operations such radioactive dust dispersion, high dose
rates, modification of plant scenario resulting from containment flooding, radiation
contamination, re-criticality, etc.

Place all collected kind of high activity inside-containment debris into debris containers
or canisters:–•

No generic deviation type in this action deserving issues identification and risks
ranking methods and tools application is foreseen.

Deviations match those ones already indicated for collection of corium reference
situation.

Final transportation to off-site dry, safe storage:–

Deviations match those ones already indicated for collection of corium reference
situation.

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Different provisions for site modification dealing with following topics: containment
isolation, full area decontamination, tracking of nuclear materials, establishment of aregular monitoring programme to ensure continued safety of the plant:–

Deviations match those ones already indicated for collection of corium reference
situation.

Preparation for defuelling. Safety requirements

Brief description of the issue
When developing new tools and equipment and planning defuelling operations, safety
requirements should be clarified and shared among the relevant stakeholders. This is important
not only to protect people, including public and workers, and the environment, but also tofacilitate the defuelling through avoiding unnecessary rework for preparation based on
unclarified requirements.

Reference situations and deviation types
These safety requirements should build on the reality of the facilities, e.g. availability of existing
facilities and effectiveness of human actions as safety functions, and consequently may bedifferent from those for the operational power reactors.
Deviation could come from too much conservatism due to lack of information. This may
result in extended period of time and increased work for preparation, which in turn would
extend time at risk and increase occupational exposure.

Issue identification and risk ranking methods
Once each LTM goal has been analysed by having referred to their actions and related reference
situations, within which a set of main potential deviations have been anticipated, issues
identification and risks ranking methods will be applied based on the nature of the identified
deviations. In order to identify such methods, the main source of information will come fromindustry, and more specifically, from nuclear safety and risk applications.
Identification and ranking tools might be easily categorised in qualitative and quantitative
methods. Since LTM has not long been one main focus of attention within nuclear safety, someof the suggested methods have not ever been applied to this particular field. Nonetheless,extrapolation might be carried out provided the method applies to the same category of treated
events. As an instance, this is the case of fault tree analysis (Appendix C) addressing portableequipment rather than traditional equipment, or SA system codes in extending the calculations
at long-term time frames typical from LTM, e.g. in the order of weeks.
The suggested methods are linked to each identified deviation and they are followed byeither an “I” or “R” letter standing for methods which address issues identification or risk
ranking.
For each goal/action/reference situation, possible issue identification and risk ranking
methods are listed in Table 3.2.

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Table 3.2. Issue Identification and risk ranking methods applicableto long-term management and actions (LTMA)
Goal/action/reference situation

Deviation

Issue identification/risk ranking methods

Corium heat removal
Plant performance

Deviations from intrinsic equipment failures• Single-system fault tree analysis (See Appendix C) (I & R)
• Failure mode and effect analysis (I & R)
• SA system codes (I)• Extended fault tree analysis (See Appendix C) (I & R)
• Engineering analysis to test the reliability of the equipment
operating in extreme conditions during long-term periods(I)

Slowly-increasing performance degradation• Hazard and Operability Study (I & R)
• SA system codes (I)• Phenomena identification ranking table (I & R)
• Engineering analysis to test the reliability of the equipment
operating in extreme conditions during long-term periods(I)

Depletion-related issues• Failure mode and effect analysis (I & R)
• Extended fault tree analysis (I & R)

Maintaining LTM
entry conditions

Recovery actions

Back-up guidelines supporting
operator actions

Impact of short-term actions in the long term

• Research survey (I)Instrumentation capability• Redundancy (I)• Analytical calculations (I)Standard safety equipment (including worsening
consequences)• Pipe and instrumentation and instrumentation and control
drawings (I), systems performance procedures (I)Once LTM entry conditions are met, current existing
SAMGs, in the most favourable case, will only give
instructions to follow-up generic actions• Probabilistic risk assessment• SA system codes• (See Appendix IV) (I)

Unbearable environments resulting from theongoing accident itself or as a consequence of taken
actions to reach and maintain LTM safe stable state• Human Reliability Analysis techniques (identification stage)(I & R)• Walkdown (I)• SA system codes (I)Working conditions

Environment conditions

Physical impediment to access to specific areas and• Walkdown (I)rooms or freely move throughout the site
Human action performance(holistically estimated)

Human actions can depart from their expected
performance because of so-called categories ofomission or commission, under which precursors
such as lack of conducting guidance, ambiguous
information, etc.

• HRA techniques (I & R)
• Operating experience (I & R)
• SA-LT drills (I)Ongoing SA evolution

Environment in primary/secondary
system and attached buildings

Environmental characterisation on safety systems
constraints and equipment survivability as aconsequence of sharp modifications in accident
evolution in rooms and buildings hosting safety
equipment

Prevention of further challenges todefence in depth barriersContainment and RPV leak tightness could be
jeopardised even if corium is effectively cooled
down• SA system code simulations (I)• Risk Oriented Accident Analysis Methodology (I & R)

New accident conditions• Probabilistic risk assessment event trees dealing withexternal events (See Appendix II) (I & R)
• Risk Oriented Accident Analysis Methodology (I & R)

External safety barriersProtection against new external hazards andoutside environmental threatens against directly
exposed portable equipment

Internal safety barriers• Failure mode and effect analysis (I & R)
Internal failures not stemming from mitigating
systems directly propagating to sudden loss of top- • Fault tree analysis (I & R)
level function/goal• Risk Oriented Accident Analysis Methodology (I & R)

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Table 3.2. Issue Identification and risk ranking methods applicable to LTMA (cont’d)
Goal/action/reference situation

Deviation

Issue identification/risk ranking methods

Mitigation of radioactive releases
Gas emissions
Radiation monitor failure
Identification of release locations

Filtering of containment gases andradionuclides

Inaccessibility to monitor readings• Dedicated movable radiation monitor devices (I)• SA system code sequence results fed by available online
accident evolution information (I & R)
• Radiation modelling and tracking codes for auxiliarybuildings to containment for indirect confirmation (I)• Risk Oriented Accident Analysis Methodology (I & R)

Slowly-increasing performance degradation(reaching threshold performance limits), both incontainment and in attached buildings tocontainment where radioactive cloud has leaked
into• Hazard and Operability Study (I & R)
• Phenomena identification ranking table (I & R)
• Engineering Tests fed by Long-Term SA system code
simulations (I)• Risk Oriented Accident Analysis Methodology (I & R)

Lack of feedback from available instrumentation tocheck impact of carried out actions• Indirect measures (e.g. pressure evolution in potential
released points, radiation meter evolution on-site or off-site
close to the release point, etc.) (I)

Stable storage threatened by external hazards• Failure mode and effect analysis applied to water treatment
new equipment coupled with containment/auxiliarybuildings (I & R)

Logistic issues• Preliminary design of water treatment and disposal
facility (I)• Operating experience (I)Failures in water treatment (including fission
product control, monitor and sampling)• Failure mode and effect analysis (I & R)
• Operating experience (I)Drawbacks in building additional barriers for liquid
effluents confinement• Engineering and manufacturing analysis (I)• Operating experience (I)Drawbacks in temporary one-through liquid
treatment or non-treatment water released tooutside environment• Preventive measures (I)• Operating experience (I)Liquid effluent emissions

Treatment and storage of highly
containment waters

Minimisation/prevention of liquid
effluents entering/existing
containment

Decontamination of highly contaminated floors and walls
Identification and division of plantcontaminated buildings into
different levels of contamination

Wrong schedule of activities
Wrong diagnosis

Site buildings clean-up
Debris and sludge characterisation
Off-site soil and groundwater
monitor and decontamination

Inaccessibility due to environmental constraints incase of radiation monitor failure• Dedicated movable radiation monitor devices (I)• SA system code sequence results fed by available online
accident evolution information (I & R)
• Radiation modelling and tracking codes for auxiliarybuildings to containment and off-site area for indirectconfirmation (I)Prevention of risk of explosions
Monitoring
Loss of monitoring devices
Analysis, monitoring and filtering of
flammable gases

Environmental analysis of attachedbuildings to containment

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Inaccessibility issues to monitor readings

Inaccessibility to rooms hosting safety equipment
or other areas where explosion might bepropagated• Redundancy (I)• Analytical calculations (I)• Indirect measures (I)• SA sequence simulations (I)• Risk Oriented Accident Analysis Methodology (I & R)
• Flammable gas clouds tracking and build-up in auxiliarybuildings simulations resulting from bounding SA
sequences to detect whether safety equipment hostingrooms might be affected by leaked flammable gas fromcontainment. (I & R)
• Risk Oriented Accident Analysis Methodology (I & R)

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Table 3.2. Issue Identification and risk ranking methods applicable to LTMA (cont’d)
Goal/action/reference situation

Deviation

Issue identification/risk ranking methods

Clearing

VentingHeating, ventilation, and air
conditioning (HVAC) system switch-on

Slowly-increasing filters performancedegradation• Hazard and Operability Study (I & R)
• Phenomena identification ranking table (I & R)
• Engineering Tests fed by Long-Term SA system code
simulations (I)• Risk Oriented Accident Analysis Methodology (I & R)

Inaccessibility due to radiation
Loss of automatic opening systems
Lack of supporting systems (power; actuation)
Inaccessibility to locations where systems can be
activated• Application and plant-specific analysis, e.g. to instructpersonnel to promptly ventilate auxiliary buildings throughoperation of HVAC or opening of doors/hatches (I)RemovalRecombiners
Injection of inert gases

Oxygen depletion
Flammable gases generation peaks
Loss of injection of inert gases conducting to
flammable gas escalation• SA sequence simulations (I)• Engineering analysis providing countermeasures(e.g. redundancy)• Dedicated procedures instructing operators to follow
suitable alternatives (I)• Fault tree analysis (I & R)
• SA sequence simulations (I)Igniters

Loss of supporting systems

Venting(already identified)(already identified)

Ambiguity/disagreement in operators/technicalsupport centre diagnosis
Absence of information on RPV/SFP state andfuel debris distribution and characteristics
Impossibility of introducing devices into
containment/RPV• Reaching a clear, unified picture by contrasting and solving
contradictory pieces of information between all sources of
information (available monitor readings – including
radioactivity area measures), analysis of code results andvisual inspections (I)• Use of alternative best-estimate stand-alone codes specificfor ex-vessel corium relocation and deposition (I)• Operating experience (I & R)
• International code benchmark exercises (I)• Research and development efforts in uncertainty
assessment, modelling and experimental analysis (I)Preparation for defuellingCollect information
Provided information during theaccident follow-up

Inspection by means of technicaldevices
Analysis with dedicated ex-vessel
phase simulation codes

Deficient provided information
Discrepancy in the results
Uncertainty in the results

Training
Training activities

N/A• Operating experience (I & R)
• Dedicated facilities (I & R)
• Virtual reality devices (I)Defuelling operations
Suitable temporary defuelling platforms
Collection and characterisation ofdebris samples

N/A

N/A

Worsening scenario by related handling
operations• Human performance (I & R)
• Operating experience (I & R)
• Best-estimate stand-alone codes specific for debris
characterisation (I)• Operating experience (I)• Usual human performance tools (e.g. documentation
evidence field walk-down track, pre-job brief, problem
statement analysis causes testing solution, etc.) (I)• Operating experience (I)Extrapolation feasibility of sample
characterisation

Collect deposited corium
Place all collected kind of high activity
inside-containment debris into debris
containers or canisters
Disposal of solid waste by-products
used in the clean-up and treatment ofeffluents phase
Transport of high activity debris

Hazards derived from related operations(radioactive dust dispersion, high dose rates,
worsening scenarios, etc.)• Human performance (I & R)
• Operating experience (I)• Current existing dedicated guidelines (from affected plantand beyond) (I)

Final transportation to off-site dry safe
storage
Different provisions for site modification

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3.5. Action identification and ranking table exercise
Introduction, scope and objectives
The phenomena identification and ranking table (PIRT) is a structured elicitation process used
to achieve consensus on subject matters of interest. In the context of Long-Term Managementof Actions for a Severe Accident in a Nuclear Power Plant (LTMNPP), the process entails reachingconsensus on commendable LTM practices based on lessons learnt from past nuclear power
plant accidents. The management actions are informed by the state of phenomenologicalknowledge of severe accident progression and consequences in the long term as well as any
gaps therein. The actions are also informed by guidance, procedures, and availability andreliability of prevention/mitigation systems. For brevity, the long-term management PIRT will
henceforth be referred to as the AIRT.
The scope of the AIRT is focused on all necessary accident management actions to maintaina plant in a stable post-accident state and prepare for plant decommissioning. The
decommissioning action itself is not in the scope. Specifically, the scope focuses on continued
cooling of debris, managing confinement of radioactivity and managing risk from combustible
gaseous products, water and solid wastes and effluent management, site clean-up anddecontamination mitigating in-plant dust dispersion, defuelling, safe disposal of fuel debris andradioactive waste, and on-site emergency management. The last of these items covers hazard
assessment, occupational health protection, countermeasures taken, and any emergency
management and administrative actions that may arise during the recovery process. The topic
of public health is not included in the scope even though being an important element of LTM.
Other important elements of LTM, not covered in the AIRT and in consistency with the scope of
the LTMNPP project, are regulatory oversight, public interactions, environmental remediation,
rehabilitation, and economic considerations.
Acknowledging the wide types of actions addressed under the long-term management
phase, a LTM-top-goal-driven approach is taken so that rather than placing all actions together,each of them will be binned according to the objective pursued by the action.
A total of six tables have been considered, the first five corresponding to different LTM top
goals, whereas the sixth table addresses management cross-sectional issues thus affecting
actions gathered under the other five tables:1.

maintaining coolable configuration and managing confinement;

2.

water waste, solid waste, and effluent management;

3.

site clean-up and decontamination;

4.

defuelling of damaged reactors;

5.

fuel debris and radioactive waste disposal, SFP fuel removal;

6.

long-term management.

Given the aforementioned scope, the AIRT process consists of taking each major LTM goaland dissecting it into several actions (some inter-dependent) and determining what each action
is with regard to scope; practical means of achieving the scope; and gaps in technology,
procedure, and other factors that may pose a challenge in accomplishing the objectives. In
parallel, each action is evaluated with regard to their role and the degree of success inaccomplishing the overall particular LTM objective. In the final step, gaps in technology,
procedure, and other factors are ranked and prioritised, leading to a set of recommendations
for improving the LTM practices to achieve enhanced safety.
Procedural aspects of conducting the AIRT are as follows. For each major long-term accident
management action identified within the scope, the AIRT members were provided a templatecontaining information about the scope of actions associated with each major goal, and tools
and technology as well as procedures to perform each action. The members were asked to rank,
based on their knowledge and expertise, the likelihood of success of each action inaccomplishing the LTM objectives. Individual ranking was performed in a qualitative scale ofhigh (H), medium (M), and low (L) with regard to how, in the individual’s opinion, a particular134

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management action is likely to succeed in accomplishing the LTM objectives. The members
were also asked to provide individual input on gaps in technology, procedures, and other factors,
as well as rank them in a similar qualitative manner. The individual scores in both the likelihoodof success and the technology gaps were then discussed among the AIRT members, at theconclusion of which a consensus score was assigned in each category.
The objective of the AIRT is to identify open issues and gaps whether stemming fromchallenges dealing with the likelihood of success, whether from any gap such as technological,
procedural, knowledge-based, etc., linked to the main actions carried out during the long-term
phase of the severe accident as currently defined in this project. In this regard, the most relevant
AIRT outcomes to pay attention to will be those actions ranked as low (L) consensus score with regard totheir likelihood of success and high (H) consensus score with regard to gaps.
Notwithstanding the above, the entire spectrum covered by the long-term phase of thesevere accident presents to a more or less extent certain degree of uncertainty. That is to say,issues or gaps can be identified for nearly the whole list of related actions.
This is why additional relevant issues should also be lifted up from those actions ranked as
medium in terms of likelihood of success and/or gaps, since such “medium” means that limited
yet existing challenges and/or gaps for the action to be accomplished have been identified.
Therefore, and even if not as highly relevant as the former ones categorised with L/H
combinations, medium ranked actions should be analysed to highlight the associated open
issues and/or gaps. In this way, a threshold to focus the attention to will be fixed between thoseactions presenting a medium or high value against those ones only ranked with low significant
related marks. And within the selected actions, conclusions derived from those ranked as L/Hshould be prioritised.
It is worth noting that unlike traditional PIRT where actions are ranked upon their importance, allaccident management actions are deemed important and necessary by definition; thus their importancewill be neither discussed nor ranked.

Elicitation process
Individual process of assessment
Besides the AIRT panel member’s expertise, any kind of ranking activity largely relies on
subjective cognitive thinking processes taking place along the time interval an issue is being
taken into consideration. Factors such as scenario characterisation and forecast, or assumed
hypotheses taken as initial and boundary conditions, might contribute to discrepancies in
judging the success likelihood of an action.
To address such lack of transparency and provide with an as-objective-as-possible process,the success likelihood ranking – together with the technology gaps ranking – will be further
described based on the particular aspects characterising the long-term management actions as
included in the tables.
The idea is to provide with a traceable path whose application, on one hand, allows readers
to have a clear understanding on the reasons behind the given rankings, and on the other hand,allows panel members ranking the actions based on the same rationale.

Likelihood of success
When looking at the actions developed along the LTM phase, their likelihood of successcomprises the different challenges found throughout the arrangement settings, disposition andperformance of an action in accomplishing its goals. This means that in order for an action tocarry out and fulfil its functions, several barriers of different nature must be overcome. Suchbarriers cover the entire process from the very beginning – at the design or knowledge phase –to the very end – at the time of actual implementation of the action on-site. Figure 3.5 below
graphically represents such process where the main items are afterwards explained.

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Figure 3.5. Aspects considered for deciding on the likelihoodof success of a given long-term management action

Technologically available/knowledge issue: Tools
An action usually relies on devices 15 , and these devices may be available worldwide andtechnologically mature enough to perform its intended functions.
Harsh environment conditions typical of a severe accident might challenge LTM via two
different ways: by jeopardising the equipment performance because of going beyond design
specifications (addressed under this issue), or by preventing the implementation or handling of
the equipment (addressed under the accessibility issue).
The success of an action will also largely depend on the knowledge of i) the environmental
conditions for the use of the device ii) the device functioning in the environmental conditionsiii) challenges associated to the action to perform (e.g. characteristics of matter to treat for fuelretrieval, for waste management, etc.). One has thus to consider uncertainties associated to thedetailed knowledge of plant damaged state.

Available on-site/logistics: Provisions
While the device can be found in the industry, i.e. it has been designed and manufactured, suchdevice must be found on-site in order to be implemented. Depending on the issue,
accommodating the physical needs underlying every action can be more or less demanding. For
instance, water contamination handling may need large free areas to accommodate the storage
tanks and filter equipment that might not be easily available at the plant site.

Operable/constraints: Accessibility
Even if the device exists in the industry and is found on-site, the device must feature operability.
Such operability may require performing local actions to connect the device or just place it where
it is supposed to perform its function. Uncertainties on the scenario where such device must
fulfil its goal, or on how well the device copes with the harsh environmental conditions, can leadto a device which is technologically available yet where there is reasonable doubt on whether thedevice can be placed and implemented exactly when needed. For instance, devices to remove
the core from the containment may exist and be technologically demonstrated and evenavailable on-site yet they cannot perform their function because of high radiation level
preventing operators to handle the device inside containment.

15. In this context, the word “device'” plays the equivalent role than the acronym SSC (structures, systems
and components), that is to say, it refers to any kind of equipment, device, system or even entire
installation whose correct performance is necessary to fulfil the pursued function.

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Going through the three fundamental challenges jeopardising the implementation of anaction, each of them may be qualitatively ranked into three different categories according toTable 3.3a.

Table 3.3a. Classification of challenges associated with implementationof long-term management actions
Challenge

State

Tools

Provisions

Accessibility

True (T)The device is properly designed
and qualified to tackle withenvironmental conditions andchallenges linked to action
fulfilment.

The necessary tools, related
human actions, logistics,
allocated surface, etc., do exist
on-site and a dedicated plan toimplement the action is available
or the action is simple enough sothat no dedicated plan should be
followed up.

No constraint of any typepreventing the action to be
implemented exists dealing with
context concerns such as physical
impediment, high-level dose, etc.

Partlytrue (PT)

Uncertainties stemming from
coping well with any type of LTM
scenario at the technology level
exists.

The necessary equipment (aslisted above) does not exist onsite but it is likely (including time
constraints) to bring it into theplant and accommodate it whenneeded.

Reasonable doubts due toaleatory or epistemic uncertainty
in the accident evolution – before
and after entering into the LT
phase – exist to access to thelocations where the equipment
must be placed to perform its
function.

False (F)The technology to build up thedevice used to perform the action
has not been designed or
manufactured.

The equipment (as listed above) is
not available on-site and it cannot
be brought to the plant easily(including time constraints).

Environmental constraints arelikely too demanding to avoid
getting closer to the location
where the equipment is intended
to perform its function.

Technology gaps
Regarding the second ranking process of each action, namely the “technology gaps”, thiselement covers a broader scope than the above mentioned challenge of “tools” affecting the
likelihood of success. In particular, any gap in technology but also in procedures or necessary
knowledge to fulfil the action should be estimated here. Therefore, a false condition in tools willlead to a low ranking in technology gaps but not necessarily the other way around.
Table 3.3b collects the material conditional logical connectives between the challenges tothe likelihood of success and the two ranked elements for each action.

Table 3.3b. Logical connectives used to fill in Table 3.4
Tools

Provisions

Accessibility

Likelihood of successTechnology gaps

True

Low

Partly true

Medium

False

Low
False

Medium/low

True

Partly true

MediumPartly true

True

Medium
False

High

Low

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It is worth noting that in contrast with a false value in tools or accessibility, a false value inthe provisions side does not necessarily lead to a low value in the likelihood of success sincedepending on the infrastructure status and the LTM goal evolution, e.g. available time for theoperators to maintain the plant under a safe stable state by only using the existing resources
on-site, the function might be successfully met or not. On the contrary, a false value either intools or accessibility directly leads to a low likelihood of success for obvious reasons,i.e. whether the equipment has not even been designed and manufactured or – even if available
on-site – the plant conditions do not allow the action being implemented, the device will not be
operative so that the likelihood of success must be necessarily low. 16.
The likelihood of success and the technology gap will then be ranked as a set of particular
combinations between states and challenges according to Table 3.4, where the most relevant
outcomes, i.e. L/H combinations, have been highlighted within a black-frame box. Therefore,
the likelihood of success is derived from the three aforementioned issues of tools, provisions
and accessibility, whereas the state of the tools issue directly relates to the technology andknowledge gaps.

Table 3.4. Combination of classification of long-term management challenges andderived connection with “likelihood of success” and “technology gaps”
Challenge

Likelihood of
success

Technology
gaps

T

High

Low

PT

Medium/high

Low

F

Low

Low

PT

T

Medium/high

Low

T

PT

PT

Medium

Low

T

PT

F

Low

Low

T

F

T

Medium

Low

T

F

PT

Medium/low

Low

T

F

F

Low

Low

PT

T

T

Medium/highMediumPT

T

PT

MediumMedium

PT

T

F

Low

MediumPT

PT

T

MediumMediumPT

PT

PT

Medium/low

Medium

PT

PT

F

Low

Medium

PT

F

T

Low

Medium

PT

F

PT

Low

Medium

PT

F

F

Low

Medium

F
F

T
T

T
PT

Low
Low

High
High

F

T

F

Low

High

F

PT

T

Low

High

F

PT

PT

Low

High

F

PT

F

Low

High

F

F

T

Low

High

F

F

PT

Low

High

F

F

F

Low

High

Tools

Provisions

T

T

T

T

T

T

T

Accessibility

16. Unless accessibility can be recovered through decontamination and cleaning.

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APPROACH TO LONG-TERM POST-ACCIDENT MANAGEMENT AND ACTIONS

Collective process of assessment
AIRT activities usually reach consensus in individual ranking elicitation whether by integratingor agreeing the different votes.
Integration processes can follow quantitative or qualitative methods. Looking at the former,
they are usually based on applying a weighted average to the number of votes ranking the item
differently, and then linking different intervals, within which the numerical result is assigned,back with a specific ranking level. Such approach can be of importance for large PIRT panels
whereas for the current case, due to the relatively low number of participants in the elicitation
process, reaching common consensus through open discussion was taken as the optimal
solution.
After collecting and summing up all the individual votes, the panel went through each item
during a two-day meeting so that any concern and discrepancy in the interpretation andassessment was raised up and openly argued by the panel members. This consensus-based
method leads ultimately to a single ranking level in most cases, and any divergent opinion anddiscussion therein are incorporated as side notes to the table.

Presentation of the results
The AIRT is organised upon LTM goals. For each goal, a table of actions will be presented. Withineach table, each action or item will feature a twofold common-consensus ranking in terms oflikelihood of success and technology and knowledge gaps, together with an additional last box
highlighting the main challenges and open issues related to that item.
Alongside with each table, the two given rankings will be explained per item according to
the arguments presented during the panel discussions. If non-reducible discrepancies were
found, the reasons behind such lacking of consensus will be incorporated into the text and themain insights raised up during the meeting will be added in the third column.
It is worth noting that the supporting comments describing the underlying reasons anddiscrepancies for each action are as important as their final votes since they give the clue forhaving a good understanding of the results.

Comments and results of the action identification and ranking tables
General preliminary aspects
In order to assess and rank the actions, all panel members should share a common view of theinitial and boundary conditions characterising the LTM scenarios, while recognising the wide
spectrum of accidents falling under the long-term phase of a severe accident. The minimal
conditions that all LTM scenarios should meet are the following:
A severe accident involving limited or extended core damage has occurred but the current
status meet the LTM entry conditions as specified in the definition (Section 3.1):•

the decay heat is successfully being removed;•

no flammable situation exists on-site and there is no further hydrogen and carbon
monoxide generation from steam-metal reaction or MCCI;•

the fuel is maintained subcritical at all locations;•the containment pressure and RCS pressure are kept low;•

the release of fission products is exceedingly small;•same conditions apply to the SFP.

Three options have been considered in the ranking exercise:•a severe accident has occurred in the reactor;•

a severe accident has occurred in the SFP;•

a severe accident has occurred in both.

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However, it appeared during the ranking exercise that some panel members mostly focusedon in-reactor severe accidents where others did consider the three options.
Due to the wide variation LTM scenarios embrace, and since each action is assigned to asingle ranking level, this level should ideally bound all type of LTM scenarios, and in case ofhighly depending on the magnitude of the accident, it should look at the worst situation to beon the conservative side. Appropriate side notes may then clarify the issue. Therefore, the AIRT
tables should not be linked to any particular accident scenario.

Comments and results
4)

Maintain coolable configuration and manage confinement

Inject water in-vessel
Likelihood of successTechnology gaps

Challenging/open issues

High

Low

Long-term performance of equipment uncertain. Back-up
systems should be pre-planned.

Likelihood of success: Provided equipment is appropriately qualified and water sources
adequately provisioned. The action has already been successfully implemented as afundamental piece of the LTM entry criteria meeting conditions hence the ongoing part of theaction is the only one here being assessed.
Technology and knowledge gaps: Robust designs and criteria exist for equipment and systems;
procedures are validated in simulated environment. The necessary allocated resources in termsof structure, system and components, i.e. valves, pumps and heat exchangers, do not cope withhighly ambiguous scenarios after having reached LTM entry conditions.
Raised concerns: The main arguments claimed by the panel members not agreeing with thehigh likelihood of success have to do with the technological issue on one side and theaccessibility issue on the other. Regarding the former, the in-vessel injection, as well as thecontainment injection, can be carried out in permanent direct mode or by means ofrecirculation mode using either fixed and/or mobile equipment. The recirculation mode, even
if a commendable practice because of avoiding further issues stemming from handling large
masses of contaminated water, can contribute highly to risk as its good performance rely on aconsiderable number of components sensitive to fail. For instance, loss of long-termcore/containment cooling can result from debris-induced direct clogging of the recirculation
cooling loop. Efficiency of heat exchangers can also degrade because of fouling by deposits onto
the surfaces coming out of contaminated water. When the recirculation system takes thecontaminated water out of the containment, then it will be source of radiation and in the caseof failure also source of leakage. Looking at the logic combinations as stated above in Table 3.3,
doubts on technological performance will lead in average to medium likelihood of success.
Regarding the accessibility issues, depending on the level of radioactive releases and spread,reasonable doubts on necessary operability-related actions, e.g. maintenance and replacement,
might be an issue difficult to cope with. A similar concern was raised if the action is performed
through emergency core cooling system in recirculation mode since contaminated water will
circulate outside containment (for certain designs), preventing field operators to get closeenough to implement eventual necessary actions that must be taken locally. Looking at the logic
combinations as stated above in Table 3.3, lack of accessibility will directly lead to a medium oreven low likelihood of success.

Inject water in containment

140

Likelihood of successTechnology gaps

Challenging/open issues

High

Low

Long-term performance of equipment uncertain. Back-up
systems should be pre-planned.

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Likelihood of success and technology and knowledge gaps: the action of injecting water
directly to the containment to remove the decay heat thereby depressurising and cooling down
the containment shares almost every view and issue with the former action of injecting water
into the vessel: the likelihood of success will then be ranked high and technology gaps low
because of the same arguments presented above.
Dealing with raised concerns, the same issues as underlined above are here highlighted,
i.e. debris-induced problems in recirculation-related devices; taking contaminated wateroutside containment/reactor building; accessibility issues in case local actions are needed; andportable equipment – thus related personnel in charge of operating the system – exposed tohigh dose environments, internal equipment failure along the long-term phase; similar clogging
issues as mentioned in the previous action; etc.

RPV pressure control
Likelihood of successTechnology gaps

Challenging/open issues

High

Low

Plants should be equipped with reliable pressure control
systems.

Likelihood of success: pressure control procedure and technology are robust and should not
be an issue for LTM; provided equipment is appropriately qualified and strategy anticipated forstation black-out (SBO) conditions.
Primary pressure will only contribute to meeting LTM goals under very limited scenarios:
once achieved LTM entry conditions, this action will only be relevant provided the RPV has not
failed. The primary system can be depressurised by correctly following SAMGs usual first actionof depressurising the primary system through the pressuriser pilot operated relief valves or
through the secondary side by means of steam generators cooling down for PWR designs, or
through the primary system safety relief valves for BWR designs, or unintentionally in case of aloss of primary system integrity.
The need of reliable primary depressurisation for PWR was clearly recognised and existing
plants are being upgraded by depressurisation systems reliable in severe accident conditions.
Technology and knowledge gaps: redundancies built into current designs and utilities haverobust procedures; technology gap, if any, is largely a material performance issue.
Raised concerns: safety relief valves can play an additional role in PWR designs in case of
secondary side recovery acting as a heat sink since the RCS will have to be first repressurised byclosing the safety relief valves. Such heat transfer mechanism can substantially improve the LTM
management by avoiding further issues related to handling large masses of contaminated water,and by significantly reducing the sources of risk threatening the goal of maintaining a coolable
configuration by removing heat through a clean system fully separated from the fluid in contact
with the radioactive source. Safety relief valve performance under highly contaminated, dirty
water and harsh environment conditions should be therefore further investigated.

Maintain RPV integrity
Likelihood of successTechnology gaps

Challenging/open issues

MediumMediumKnowledge of late-phase melt progression. Remaining strengthof damaged RPV cannot be assessed. Materials performanceissues; embrittlement, corrosion.

Likelihood of success: RPV failure may be avoided in design extended condition B-likescenarios – extended core-damage scenarios with partial or full core relocated to the lower
plenum by injecting water into the vessel and/or into the reactor cavity. Significant
uncertainties exist on the damaged status of the RPV (with possible fragilities induced locallyby contact with hot materials) and whether the damaged RPV will bear the thermal andmechanical loads in the long term, even if it had resisted in the short term.
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According to Table 3.3, a partly true statement in tools corresponding to uncertainties
stemming from coping well with any type of LTM scenario at the technology level, directly yields
medium ranks for the likelihood of success.
Technology and knowledge gaps: re-establishment of the in-vessel injection after
successfully prevented RPV failure by ex-vessel cooling, long-term creep and embrittlement
effects on the lower head materials, long-term resistance of damaged RPV and risk of failure,
long-term effect of minor leakages, solidified corium retrieval operations from the RPV, etc.
Raised concerns: no additional long-term safety concerns coming from discrepant opinions
other than those already pointed out above.

Containment pressure control
Likelihood of successTechnology gaps

Challenging/open issues

High

Low

Related to hydrogen control, see next item.

Likelihood of success: after entering into the LTM phase, the containment pressure shouldhave been kept low enough to minimise further environmental release of radioactivity. Thiscondition – together with the fact of meeting with the rest of the LTM entry criteria – makes that
containment pressure control will hardly become an issue in the long term.
Several reliable means of containment pressure controls exist including containment sprays,
hard venting and filtered containment venting system, decay heat removal by portableequipment or standard emergency core cooling system in injection (temporary) or recirculationmode, etc. Procedures for implementing these systems in any combination are well established.
Technology and knowledge gaps: robust designs and criteria exist for equipment andsystems; procedures are validated in simulated environment.
Raised concerns: containment leak tightness partly relies on containment penetrationmaterials exposed to typical harsh environment conditions. Experiments on penetrationmaterials have been made by using nitrogen instead of hydrogen so that aggravated corrosion
mechanisms have not been taken into account on their survivability. In addition, containment
liner may undergo galvanic punctual corrosion thereby leading to containment leakage. With
respect to containment cooling systems, such as fan coolers or containment sprays, aerosol
deposition onto the heat exchanger surfaces or sprays nozzles can soundly degrade the heattransfer mechanism. Recirculation sumps blockage is also an issue if containment sprays is theremoval heat mechanism. Portable equipment suffers from the same problems than the invessel injection action.

Containment hydrogen control
Likelihood of successTechnology gaps

Challenging/open issues

High

Low

Radiolysis needs to be considered.

Likelihood of success: once reached LTM entry conditions, hydrogen (and carbon monoxide)
production must have lowered below safety thresholds. Provided control and monitoring ofhydrogen is established and mitigation means appropriately designed, the likelihood of successshould be high.
Technology and knowledge gaps: the only significant sources of hydrogen generation arewater radiolysis and corrosion reactions, which will hardly stop even after entering into the LTM.
Even though the generation rates are likely to be low enough to be easily detected before
departing from a safe state, the main concern arises from flammable gases accumulation inplaces where no detection nor mitigation measures are provided. Such build-up situation is not
limited to the reactor containment.
Raised concerns: N/A142

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Confinement of fission products
Likelihood of successTechnology gaps

Challenging/open issues

MediumMediumUncertainties regarding status of confinement, chemistry andremobilisation processes of fission products in the long-term
affecting releases.

Likelihood of success: fission products confinement on the LT very much depends on thecontainment barrier status when a coolable configuration is reached and on the extent of
releases either airborne (gas and aerosols) or liquid that occurred through containment breaches
during the emergency phase of the accident.
If the containment leak tightness is not affected during the emergency phase, fission
products confinement is achievable. Then with unaffected containment, the likelihood of
success of fission products confinement is high.
Maintaining confinement of fission products is of course much more challenging if thecontainment leak tightness had been lost during the emergency phase. In such situations, thelikelihood of success of fission products confinement is lower. Containment decontaminationand debris removal will be more challenging.
With these considerations, the likelihood of success was classified medium.
Technology and knowledge gaps: potential remobilisation of gaseous iodine species in the
weeks following the accident from pools, sumps, surface deposits (dose effects, evolution ofwater phase chemistry in scrubbers, in suppression pools, in sumps, radiolytic decompositionof deposited aerosols on surfaces and filters, effect of impurities and corrosion reactions).
Leaching and transfer of Cs isotopes, Sr, Pu, Am in waters on the LT depending on water phasechemistry.
Raised concerns: N/A

Minimise environmental release (airborne and aqueous)
Likelihood of successTechnology gaps

Challenging/open issues

MediumMediumKnowledge/technology gaps mainly concerning aqueous
release.

Likelihood of success: airborne and aqueous release control requires previous analysis of
the status of the containment and identification and control of the potential leakages, extending
those activities to potential containment bypass situations, e.g. through the steam generators(for PWR designs) or primary interfacing systems, or recirculation interfacing system loss-ofcoolant accident (interfacing loss-of-coolant accident during the recirculation switch
connecting containment/primary systems with the outside environment). Depending on thesevere accident evolution, pinpointing and control of releases might be a very complex issue totackle (cf. confinement of fission products).
Technology and knowledge gaps: current ways for analysing and pinpointing the exact
release locations are not satisfactory enough and they can become crucial to more suitable andfaster face the problem and isolate the leakage; leak tightness of penetrations; of loops
conveying contaminated liquid; survivability and trapping efficiency on LT of loaded filters infiltered containment venting systems; remobilisation from aerosol deposits and from pools andsumps (iodine); transfer to liquid phases (Cs, Sr, Pu, Am).
Raised concerns: N/A

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APPROACH TO LONG-TERM POST-ACCIDENT MANAGEMENT AND ACTIONSMinimise likelihood of criticality
Likelihood of successTechnology gaps

Challenging/open issues

High

MediumRisk of re-criticality sometimes difficult to assess.

Likelihood of success: likelihood of re-criticality small and effects of re-criticality events
probably limited; re-criticality assessment is possible but challenging due to uncertainties ondamaged fuel distribution and composition; monitoring and control is possible; non-borated
water injection increases the risk. Criticality can be monitored by taking containment gaseous
samples or by placing online monitors at containment ventilation outlet path, for instance, by
measuring 135Xe concentration and can be controlled by borated water injection.
Technology and knowledge gaps: risk of re-criticality is difficult to assess, Pu clusters
formation upon cooling of corium in particular for mixed oxide fuel may increase the risk; clear
water injection or boron dilution may increase the risk. Also, when starting to retrieve debris,risk of criticality is difficult to assess (e.g. when reflooding some years after the accident dried
aged debris).
Raised concerns: potential delays between event detection and implementing appropriate
countermeasure actions, e.g. boron acid injection.

Maintain SFP cooling and minimise risk
Likelihood of successHigh (loss of cooling)
Low (loss-of-coolant)

Technology gaps

Challenging/open issues

MediumValidated tools to calculate progression and consequences of
SFP accident.

Likelihood of success: the action is maintaining SFP cooling on the LT following a loss ofcooling or loss-of-coolant accident in the SFP.
The likelihood of success is considered high for loss of cooling accident where limited fuelassemblies degradation and radioactive releases can be expected and due to the possible
various means at effectively injecting water into the SFP.
The likelihood of success is considered low for unmitigated loss-of-coolant accident wherelarge fuel assemblies degradation and radioactive releases can be expected. The success of theaction may be challenged by SFP structure damages and reduced accessibility (in case of large
releases on-site, particularly for SFP outside containment).
Loss-of-coolant accidents with fuel melting in SFP should be “practically eliminated”
according to the Western European Nuclear Regulatory Association (WENRA), hence provisions
for prevention of SFP accidents should be adequately verified.
Technology and knowledge gaps: tools to calculate progression of loss of cooling and lossof-coolant accidents in SFP and assess their consequences, with consideration of plant-specific
SFP configurations, need further development and validation.
Raised concerns: N/A
The resulting common ranking for actions aiming at maintaining the coolable configurationis shown in Table 3.5.

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APPROACH TO LONG-TERM POST-ACCIDENT MANAGEMENT AND ACTIONS

2) Water waste, solid waste, and effluent management
Minimise and collect contaminated water
Likelihood of success

MediumTechnology gaps

Challenging/open issues

Low

If containment is leaking, it is a considerable challenge tohandle contaminated water. Also, if cooling is not done on
secondary side, large amounts of contaminated water are
generated from cooling.

Likelihood of success: the aspects making collecting contaminated water a challenge arelogistics in nature while technologically speaking this action does not constitute an open issue.
The complexity of successfully connecting the necessary devices to collect the water highly
depends on whether the decay heat is removed through an open or a closed loop, and evenmuch more relevant, on whether the containment has been kept isolated or it has breached.
Even if the operators manage to make the contaminated water circulate through a closed loop
provided with a pump and heat exchanger, in case there is a leakage in or out containment, partof the water will leave the containment so that it should have to be identified, stored andsuitably treated. As a matter of fact, if one of the most critical issues at Fukushima Daiichi iscurrently collecting and treating large masses of contaminated water, it is just because theunderground water coming into the reactor building is mixing with the water leaking from thecontainment. The underground water would not be a challenge if it could be pumped out or
simply be deviated outside the building. Large efforts are currently implemented at Fukushima
Daiichi NPP in that purpose.
Therefore, the action was ranked as medium because of partly true in terms of provisions.
Technology and knowledge gaps: robust designs and criteria exist for equipment andsystems. The necessary allocated resources in terms of devices, i.e. valves, pumps and heatexchangers, do not cope with highly demanding scenarios.
Raised concerns: several panel members expressed different views on how likely this action
is to be successfully implemented. Along with the specific challenge in implementing the action
wherein the emphasis was placed by each member – basically whether on tools or provisions –,divergences were mainly rooted in the different scenarios foreseen since the success in
implementing this action is highly scenario-dependent. According to the degree of core damage– arrested in-vessel or ex-vessel–, the status of the containment, and even more theconfinement of radioactive products, to handle the resulting contaminated water might range
from an easy action to deal with to a complex one.

Decontaminate collected waste water
Likelihood of success

High

Technology gaps

Challenging/open issues

Low

No technology available to separate tritium in very large
volumes and very low concentrations; capacity may be a
problem; water treatment facilities may not be readily
available as well as characterisation techniques (not easily
detectable nuclides).

Likelihood of success: once arrangements for contaminated water collection are set up,
equipment needed to decontaminate water is not technologically challenging (cf. TMI-2 andFukushima Daiichi NPP feedback). Use of raw water or sea water may however complicate
decontamination operations.
Logistics issues dealing with handling and storing large amounts of water, together withsuitable further decontamination of derived by-products do not fall under this action. Thisaction also shares with collecting contaminated water the lack of NPP preparedness in allocating

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the necessary devices and planning the necessary projects should it be needed. Nonetheless,such allocated efforts are ranked as not as demanding as those required for collecting
contaminated water. This is why rather than a medium likelihood of success, this action isranked as high.
Technology and knowledge gaps: equipment needed to processing contaminated water isappropriately qualified and extensively tested. The only remaining open issue deals with thecurrent unavailability of separating tritium in very large volumes of water.
Raised concerns: tritium issue; logistics issues but this last one actually assessed within theformer action so not applying here.

Recirculate decontaminated water for core cooling
Likelihood of successTechnology gaps

Challenging/open issues

High

Low

Damaged systems/piping may pose a problem.

Likelihood of success: when decontaminated water is available, its use for the core coolingis likely to be successful.
Decontaminated water for the core cooling is not necessarily completely clean water. It iswater cleaned from any dirt, oil, dissolved salts etc. and decontaminated from the mainradiation sources. The objective is that this water should not cause damage to pumps, valvesand piping of the cooling system and the dose rate around the system should be kept in
acceptable range.
Technology and knowledge gaps: no lack in the necessary equipment to perform this action
is found.
Raised concerns: arrangement for a reliable recirculation alignment of necessary equipment
including pumps, heat exchangers, valves and piping might be an issue if making use of devices
or piping having undergone damage yet not easily identified because of harsh environment
conditions, i.e. accessibility problems. From this point of view, the action should not be
categorised as high but as medium in terms of the likelihood of success.

Collect and store waste water
Likelihood of successTechnology gaps

Challenging/open issues

High

Low

The sheer volume of waste water may be problematic.
Release criteria should be defined.

Likelihood of success: when focusing on the storing side of the more overarching action
aimed at collecting and storing contaminated water, this action becomes as highly likely of
being successfully implemented.
Technology and knowledge gaps: no lack in the necessary equipment to perform this action
is found.
Raised concerns: the storage of large amount of contaminated (or in certain cases almost
clean) water may prohibit other activity on-site due to lack of space. If this issue is taken into
account, this action should be reassessed as medium in terms of the likelihood of success.

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APPROACH TO LONG-TERM POST-ACCIDENT MANAGEMENT AND ACTIONSMonitor and control fission products including tritium in water
Likelihood of successTechnology gaps

Challenging/open issues

High

Low

Monitoring techniques may not be readily available for all
radionuclides.
Monitoring of some not easily detectable nuclides.

Likelihood of success: monitoring is well established for major isotopes (e.g. Cs, Sr) thus thelikelihood of success was considered high. Also this action does not present strong challenging
issues in terms of necessary equipment or knowledge, provisions (allocated resources could
already be on-site or easily and rapidly delivered) and accessibility (no need to perform local
actions where harsh environment conditions might prevent their correct implementation).
Technology and knowledge gaps: technology available and proven measurement methods
well established for major isotopes. However, the monitoring of other isotopes (e.g. actinides,
some activation products) is more challenging in highly active waters. “Routine” measurements
techniques and methods for monitoring of all important nuclides in highly active waters for
LTM could be further established and shared.
Raised concerns: additional development of devices for routine activity measurement andcontamination evaluation required; international standards for measured values such as “grossbeta” should be more openly advertised; procedures for results analyses should be promoted;
different techniques and methods may be required for monitoring radionuclides for very
different water contents.

Control and minimise effluent discharge
Likelihood of successTechnology gaps

Challenging/open issues

Low (if containment leaktightness has been lost)

MediumKnowledge of containment status (leak paths).
Technics/methods to recover the containment.

Likelihood of success: the arrest of existing leaking paths might be extremely difficult tocope with not because of the necessary tools falling short in meeting their function – since theaction itself merely consists of plugging, welding, etc., the undesired radioactive leakages –, but
because of highly demanding accessibility issues. These issues are mainly related to identifying
the release location and to implementing the necessary local actions.
Technology and knowledge gaps: actions aimed at controlling and minimising radioactive
liquid effluents do not feature important gaps in terms of knowledge/devices. Possibly, there
might be some interest in developing technics and methods for recovering the containment leaktightness.
Raised concerns: no additional issues raised aside from those already included above.

Effluent characterisation
Likelihood of successTechnology gaps

Challenging/open issues

High

Low

Cf. challenges mentioned for fission product monitoring inwaters.

Likelihood of success: this action requires applying well-known techniques from feedbackof accidents, knowledge of fission products of interest and adequate measuring techniques.
Significant challenges neither found in tools, provisions nor accessibility.
Technology and knowledge gaps: no fundamental gaps are found; long-term sampling isfeasible; analyses of highly active waters may be challenging (multi-nuclide analyses).
Raised concerns: no additional issues raised aside from those already included above.
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Collect, store solid wastes and monitor fission product contents
Likelihood of successTechnology gaps

Challenging/open issues

High

Low

Monitoring of fission product content in highly active and complex
solid structures.
Accessibility and handling of highly active solids (cf. defuelling).

Likelihood of success: here we consider solid wastes other than wastes related to damaged
fuel and SFP fuel assemblies’ retrieval (treated later in point 5). Collection and storage of solid
wastes is considered with a high likelihood of success. Tools and technologies to collect andstore them temporarily on-site are considered well known. However, monitoring of fission
products in highly active and complex solid structures and accessibility to and handling of mostactive wastes may be challenging.
Technology and knowledge gaps: no fundamental gaps are found, possibly development oftechnics and methods for more “routine” fission product content measurements for highly
active wastes could help solid waste management (storage and disposal).
Raised concerns: no additional issues raised.

Storage and disposal of filtered containment venting system water and filter
Likelihood of successTechnology gaps

Challenging/open issues

High

Low

Monitoring of fission product content in highly active solid andliquid wastes (see above).
Accessibility and handling of highly active solids (see above).

Likelihood of success: see addressed arguments above dealing with collecting and storing
solid and liquid wastes.
Technology and knowledge gaps: idem.
Raised concerns: no additional issues raised with the exception that filtered containment
venting may still play a role in the long term for containment pressure and hydrogen control
and to limit radioactive releases.
In case of liquid scrubber-type filter, provisions are set so that the contaminated liquid canbe periodically removed if required. Issues associated to collection and treatment of highly
active contaminated waters apply.
For solid filters (solid filter stages exist both in solid- and liquid-type filtered containment
venting systems), removing of the filters might not be feasible for a long period of time if venting
has to remain operable. This is favourable for power and dose decay in the filter with theelimination with time of short-lived isotopes (e.g. iodine isotopes decay).
In case filter clogging occurs during the accident, this may prevent possibilities for further
venting. The action of removing the clogged filter will be difficult to implement in particular due
to high doses on the filter.
Solid filters collecting aerosols (in particular pre-filters implemented upstream some
venting lines) are expected to concentrate large activities even on the LT (Cs aerosols filtration).
Then the action of retrieving the filter may be difficult to implement due to high doses on thefilter.
The resulting common ranking for actions aiming at water waste, solid waste, and effluent
management is shown in Table 3.6.

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3) Site clean-up and decontamination
Plant site clean-up and rubbles/sludge removal (outside of the reactor)
Likelihood of successTechnology gaps

Challenging/open issues

High

Low

If major releases have occurred, the likelihood of reaching green
field status may be low.
Rubbles/sludge characterisation for waste management.

Likelihood of success: technological gaps are not foreseen, as well as significant
shortcomings in accessibility issues.
Technology and knowledge gaps: none foreseen. Site infrastructures may be developed tofacilitate site clean-up if an accident occurs.
Raised concerns: while the suitable technology exists, removal of sludge and cleaning up of
the site might be difficult to implement in those areas with low accessibility due to high dose
rates. Also, large infrastructures damages (e.g. resulting from natural hazards or explosions or
fires) will make the action implementation more challenging. If such conditions are considered,
the likelihood of success should be ranged as medium.
Active rubble/sludge characterisation (fission product content) for waste management could
be challenging (cf. solid waste management).

Surface decontamination
Likelihood of successTechnology gaps

Challenging/open issues

Low

Accessibility may be limited due to contamination, physicalconstraints, water coverage. Contamination may havediffused into structures (e.g. in concrete).

High (outside damaged plant)
Medium (in highly contaminatedzones of damaged plant)

Implementation of strippable or washable coatings in zones
were high contamination is expected (e.g. in containment).

Likelihood of success: likelihood of success of surface decontamination (including soil) onsite outside damaged buildings is expected to be high referring to recovery actions that were
implemented for instance to clean the Fukushima Daiichi site.
However, the success of surface decontamination in highly contaminated zones of damaged
plants has been ranked as medium because accessibility may be limited and contamination
may not be easily removable as was evidenced at TMI-2 and Fukushima Daiichi. Despite manydifferent decontamination techniques used for surface decontamination in TMI-2 containment,
decontamination was very demanding and very abrasive technics had to be used. At Fukushima
Daiichi, for example decontamination at the level of the refuelling floors to give access to the
RPV, still appears very challenging. This is in great part due to contamination which diffused
into concrete structures or accumulated in non-accessible zones or pipes.
Technology and knowledge gaps: many technics for surface decontamination exists but
encrusted contamination is difficult to remove. Implementation of strippable or washable
coatings in zones where high contamination is expected may be of help (e.g. in containment).
Raised concerns: similar comments dealing with accessibility as set forth in above action
dealing with site clean-up apply.

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Rubbles and sludge characterisation
Likelihood of successTechnology gaps

Challenging/open issues
Robotics and remote measurement technology could be helpful.

High

Low

Composition, content of burnable substances, etc. are also ofinterest for transport and disposal.
The sheer amount of debris may be a challenge.

Likelihood of success: this action has been analysed by most of panel members as if it only
comprised the task of characterising debris itself. From this aspect, the action has been ranked
as high since debris/sludge characterisation is not overly challenging except for fission productcontent characterisation for waste management as discussed above.
Technology and knowledge gaps: depending on the situation, debris characterisation can be
performed by already existing devices. See also the related open issues.
Raised concerns: as many other actions involving local actions, the emphasis might not be
put on characterising the debris itself but rather on the prior, necessary action to access thedebris, which can be locally prevented by high dose rates.

Monitor cooling pond (and sea) contamination
Likelihood of successTechnology gaps

Challenging/open issues

High

Low

None

Likelihood of success: as any other action dealing with monitoring activities, the current
one is ranked as high – see for instance above the monitor and control fission products including
tritium in water. This should be “routine” measurement.
Technology and knowledge gaps: none.
Raised concerns: none.

Monitor and control groundwater contamination
Likelihood of successTechnology gaps

Challenging/open issues

Low

Low

Specific technological solutions to limit groundwater contamination
are highly situation/plant specific.

Likelihood of success: the fundamental reason why this action features a low ranking relieson the difficulty to controlling the groundwater contamination, mostly in the light of the
Fukushima Daiichi NPP highly challenging scenario dealing with the groundwater issue. Further,an accident where the containment concrete floor would be breached by MCCI could also leadto significant groundwater contamination.
Such low likelihood of success derives from accessibility problems involved in theimplementation of countermeasures to prevent groundwater to be contaminated, in particular
because such countermeasures should aim at recovering containment leak tightness.
Technology and knowledge gaps: none.
Raised concerns: consensus in this action was achieved through long clarification discussions
about different plant characteristics stemming from extremely different severe accident
signatures, i.e. negligible importance in many scenarios (both looking at the severe accident
evolution and the plant site, e.g. TMI-2) but highly relevant in some other ones, e.g. the worst casescenarios likely represented by the Fukushima Daiichi NPP accidents and the Fukushima Daiichi
NPP site. In order to take the bounding case, the likelihood was finally ranked as low.

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APPROACH TO LONG-TERM POST-ACCIDENT MANAGEMENT AND ACTIONSRadiological monitoring
Likelihood of successTechnology gaps

Challenging/open issues

High

Low

Technology is available; robotics and remote measurement technology
would be helpful. Challenging in very high dose environment.

Likelihood of success: the same arguments for a high likelihood of success as stated abovein monitoring-related actions apply.
Technology and knowledge gaps: none. Technology is available; robotics and remotemeasurement technology would be helpful.
Raised concerns: none.

Characterisation of fission products distribution in contaminated spaces
Likelihood of success

MediumTechnology gaps

Challenging/open issues

MediumMethods for source distribution analysis; improved techniques for hot
spots detection and characterisation; difficulties to access highly
contaminated areas; technology is available except for molten fuellocation detection; codes cannot give this much detail.

Likelihood of success: reasonable doubts on tools and accessibility directly lead to a
medium/medium ranking according to Table 3.3.
Depending on the extent of the contaminated areas, dose rate data analysis can be very
complex and the retrieved data can be scarce from areas with low accessibility.
Technology and knowledge gaps: the main shortcomings as set forth in the likelihood of
success paragraph would significantly decrease with advancement of clean-up and additional
measurement campaigns. Efforts should be placed in developing innovative devices andinstrumentation to have a better characterisation of fuel and radionuclides distribution in adamaged plant.
Raised concerns: aside from a highly scenario-dependent action in terms of its likelihood of
success, no additional concerns were raised and a common agreement was achieved.

4) Defuelling of damaged reactors
Detect fuel debris locations and estimate debris masses (at start of defuelling)
Likelihood of successTechnology gaps

Challenging/open issues
Feedback from accidents shows this action remains highly
challenging. Particularly difficult for accident with failed RPV and/or
larger fuel dispersion.
With existing monitoring (imaging), no reliable information until
access to damaged fuel is possible (accessibility, dose).

Low

High

Distant monitoring with sufficient resolution not yet developed.
SA codes not sufficiently predictive.
Required/wished progress in technics and analytical tools scientificallychallenging.
Damaged fuel and debris distribution poorly known up to start ofdefuelling, selection of appropriate defuelling options and techniquesdelicate at this stage.

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Likelihood of success: the feedback from TMI-2, Chernobyl and Fukushima Daiichi NPP
accidents shows determination of damaged fuel distribution after a severe accident remains ahighly challenging issue. In the first five to ten years following such accidents, damaged fuelcannot be easily accessed due to high doses, even remotely, which prevents having information
on its distribution in the damaged plant. Further, accidents with a failed RPV and with larger
fuel relocation or dispersion outside RPV will make the action even more difficult. The
traditionally used monitoring means such as optical imaging by camera generally conveyed byrobots fail to provide reliable information in high radiation environment (limited lifetime,
localised measurements, low resolution and often accessibility problems).
Alternative technics offering distant monitoring could be further developed. Muon
tomography technique, recently applied at Fukushima Daiichi NPP, provided useful by limited
information on damaged fuel location – essentially providing evidence of presence of fuel or notin the RPV after the accident in unit 1 and 2 (no quantification possible due to low resolution).
At TMI-2, knowledge on fuel distribution in RPV was gained progressively by visualinspections, first at RPV opening about six years after the accident and then upon progress ofdefuelling operations.
On the analytical side, as evidenced by recent international benchmark exercises (NDA, 2011;
NEA, 2015a), severe accident calculation codes (e.g. Accident Source Term Evaluation Code(ASTEC), Modular Accident Analysis Programme [MAAP], MELCOR, SAMPSON) do not yet predictreliably enough RPV failure and damaged fuel mass distribution in-vessel and, eventually, exvessel.
Based on these arguments, panel members considered that, at start of defuelling,
establishing damaged fuel distribution has a low likelihood of success. Thus, since damaged
fuel and debris distribution is expected to be poorly known up to start of defuelling, anticipative
selection of appropriate defuelling options and techniques is delicate.
Technology and knowledge gaps: non-intrusive instruments dedicated to measure during
the accident any significant fuel movement in the RPV could be searched for (e.g. detection
during the accident of massive fuel relocation in RPV, of fuel relocation outside RPV). Such
instruments could be highly valuable both for severe accident management and LTM.
Alternative technics allowing distant monitoring of damaged fuel, such as muon tomography,
could also be further developed for LTM. Such technical developments appear highly
challenging especially if debris masses estimates are searched for.
On the analytical side, efforts are underway to improve modelling of SA codes, particularly
on late phases of core degradation in-vessel.
Raised concerns: progress both in technology and SA codes pose scientific challenges.

Identify state of degraded fuel/fuel debris including control rod and structural material
Likelihood of successTechnology gaps

Challenging/open issues
Determining the composition, fission products content and mechanicalbehaviour of degraded fuel and debris is highly challenging withoutsampling and analysis of real material.
Predictive analyses can be made using thermodynamic calculation tools
but these tools are only partially valid for complex compositions.

Low

High

Simulant materials testing may help but experimentation representativeof actual compositions is challenging.
Knowledge from performed fuel degradation experiments to beconsidered but not necessarily easily transposable to specific accident
transients.
Knowledge of ageing phenomena limited.

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Likelihood of success: as for getting information on debris distribution, identifying the stateof damaged fuel/fuel debris by optical imaging has a low likelihood of success for the reasonsearlier exposed. When accessible, low resolution imaging of damaged fuel/fuel debris may beobtained providing information on their morphology and appearance. Determining thecomposition, fission products content and mechanical behaviour of degraded fuel and debris ishighly challenging without retrieval and analysis of real material.
Predictive information on composition in-vessel and ex-vessel and physical properties canbe searched for using i) knowledge gained through performed fuel degradation experimentsii) experimentation performed using materials simulating accidental complex compositionsiii) analytical calculations such as SA codes and/or thermodynamic calculation tools. This is theapproach currently deployed at Fukushima Daiichi NPP. However, the transposition of suchknowledge to specific accident transients is not straightforward.
Ageing processes (incl. long-term leaching) may also have to be considered for cases wherefuel retrieval can only be engaged (due e.g. to dose and accessibility limitations) long times after
the accident (cf. Chernobyl).
Technology and knowledge gaps: progress in analytical tools (SA codes, thermodynamic
databases) searched for, but this is scientifically challenging. It could be valuable to generate aknowledge database related to damaged fuel/fuel debris recovered and analysed in accidents
and in relevant testing (composition and physical properties and ageing). Limited knowledge on
ageing processes (dose effects, leaching) on the LT.
Raised concerns: progressing in analytical tools
experimentation are considered scientifically challenging.

andperforming

representative

Collection of damaged fuel/debris samples and characterisation
Likelihood of successTechnology gaps

Challenging/open issues
It may be actually challenging to retrieve samples. Specific remotecontrol technologies may have to be developed depending on
damaged fuel/debris characteristics.

MediumLow

As seen above, high dose, knowledge of debris characteristics.
Characterisation techniques for damaged fuel/debris are well known
from previous accidents and degradation experiments. Large experiencein hot labs.

Likelihood of success: specific remote control technologies (cutting, lifting, collecting,handling) may have to be developed depending on damaged fuel/debris characteristics(e.g. mechanical behaviour), location (accessibility from top, side, etc.) and conditions (underwater or in air). Since getting information on damaged fuel/fuel debris distribution andcharacteristics is challenging, organising and designing tools for samples collection isconsidered to have a medium likelihood of success. With progress in damaged fuel retrieval andknowledge of its characteristics, the likelihood of success of the action should increase. Morechallenges will arise in accidents where damaged fuel has relocated in many different places(failed RPV, fuel dispersion).
Since characterisation techniques for damaged fuel/debris are well known and large
experience was built in hot labs, there was no raised concern related to damaged fuel/debris
characterisation.
Technology and knowledge gaps: as discussed above, damaged fuel/debris characteristics.
Raised concerns: none.

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Fuel debris examination and characterisation for storage
Likelihood of success

High

Technology gaps

Low

Challenging/open issues
Knowledge of damaged fuel characteristics should be obtained during
fuel retrieval operation
Specific tools may have to be developed for characterisation related tocriteria (mass, dose, fission products content) set for waste storage

Likelihood of success: damaged fuel/fuel debris characterisation should be conductedduring fuel retrieval operations. Specific device may have to be developed to get data relevant
for waste storage management (cf. point 2 related to waste management). Panel members
considered that once challenges associated to damaged fuel retrieval have been solved, thelikelihood of success of the action is high.
Technology and knowledge gaps: see above.
Raised concerns: none.

Minimise environmental release during defuelling
Likelihood of success

MediumTechnology gaps

Challenging/open issues

MediumDepending on damaged fuel distribution, locations and characteristics
may be challenging to minimise releases during fuel retrieval operations
Development of cutting techniques limiting aerosol generation anddispersion considering damaged fuel characteristics

Likelihood of success: the likelihood of success of the action has been considered medium
as depending on damaged fuel distribution (e.g. fuel relocated outside RPV in different areas),
location (in air or under water, low accessible and high dose areas) and characteristics (hardness,brittleness, eventually considering ageing), it may be challenging to minimise releases byimplementing countermeasures (confined area, filtering, etc.). Further, new cutting techniques
limiting dust generation and dispersion may have to be developed considering damaged fuelcharacteristics (e.g. hardness, brittleness). Some developments are being pursued in the
preparation of damaged fuel retrieval at Fukushima Daiichi NPP.
Technology and knowledge gaps: medium for the reasons exposed above.
Raised concerns: none.

5) Damaged fuel/fuel debris and radioactive waste disposal, SFP fuel removal
Temporary on-site storage of damaged fuel/fuel debris
Likelihood of success

High

154

Technology gaps

Challenging/open issues

Low

Some challenge may arise due to volume, dose and space restrictions
on-site. More challenging for multi-units, SFP accidents with significant
fuel damage, accidents with significant fuel dispersion.
May necessitate large-scale constructions (temporary confinement
before retrieval, installations for storage after retrieval).
Criteria for site location and/or storage containers not always defined.
Criteria for appropriate handling and storage approach to be definedconsidering materials characteristics (dose, mass, material composition,
fission products content).

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Likelihood of success: the likelihood of success of the action was considered high even if
challenges may arise for accidents where fuel retrieval is particularly complex (e.g. dispersedfuel) and when large volumes of highly active material are generated (SFP or multi-units
accident). Then the temporary on-site storage of damaged fuel or fuel debris may necessitate
large-scale constructions (e.g. temporary confinement before retrieval such as at Chernobyl,
installations for storage of retrieved material).
As for any nuclear waste management (cf. point 2), criteria (dose, mass, material
composition, fission products content) for handling and storage approaches have to be defined.
Technology and knowledge gaps: as discussed earlier, damaged fuel/debris characteristics.
Raised concerns: none.

Temporary on-site storage of structural debris
Likelihood of success

High

Technology gaps

Challenging/open issues

Low

Space restrictions may be a serious problem depending on debris
volumes.
Criteria for appropriate handling and storage approach to be definedconsidering materials characteristics (dose, mass, material composition).

Likelihood of success: similar concerns as the one expressed above for damaged fuel/fueldebris were expressed in case of large volumes of structural debris.
Technology and knowledge gaps: structural debris characterisation.
Raised concerns: none.

Temporary on-site storage of filters used in water decontamination process
Likelihood of successTechnology gaps

Challenging/open issues
Dose and space restrictions if large volumes of contaminated water haveto be treated.

High

Low

Criteria for appropriate handling and storage approach to be definedconsidering filters loading (dose, mass, fission products content).
Waste may need to be stabilised (limit remobilisation).

Likelihood of success: the likelihood of success of the action is considered high even if morechallenging, due to space restrictions considerations, when large volumes of contaminatedwater have to be treated. Filters loading should be controlled during the water decontamination
process. Contamination captured in the filter may need to be fixed to minimise any
remobilisation.
As for any nuclear waste management (cf. point 2), criteria for handling and storage
approaches have to be defined.
Technology and knowledge gaps: none.
Raised concerns: none.

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Material control and accountability
Likelihood of successTechnology gaps

Challenging/open issues
Necessary level of detail in accounting to be defined depending on
considered material, not the same as under normal operation.

High

Low

Adequate devices and methods for “routine” material control andaccountability to be designed especially when large volumes of wastes
have to be treated.

Likelihood of success: the likelihood of success of the action is considered high even if alarge variety of wastes has to be treated. Adequate devices and methods for “routine” material
control and accountability have to be developed based on defined criteria.
Technology and knowledge gaps: none.
Raised concerns: none.

SFP defuelling
Likelihood of successTechnology gaps

Challenging/open issues
Complexity depending on extent of fuel damage in SFP, SFP building
damage (if external to confinement) and on-site radiological situation(accessibility to SFP).

High (loss of cooling)

Low

Identification, characterisation, collection and handling of mechanicallyweakened fuel assemblies, damaged fuel, fuel debris (similar to fuel inreactor, cf. point 4).
Damaged racks.

Likelihood of success: the action of SFP defuelling is considered to have a high likelihood of
success when considering a loss of cooling accident with moderate fuel degradation (no loss offuel assembly integrity). The situation would be much more challenging for a loss-of-coolant
accident which would result in large fuel damage. Then technical challenges would be similarin nature to those associated to fuel retrieval in a damaged reactor, possibly with higher
difficulties arising from the volumes of damaged fuel to recover, handle and store.
If the SFP is external to confinement, a temporary confinement may have to be establishedover the damaged SFP to limit radioactive releases prior to and during damaged fuel retrieval.
Technology and knowledge gaps: see reactor defuelling and open issues associated to LT
SFP cooling.
Raised concerns: see above.

6) Long-term management
Radiological and security monitoring on-siteLikelihood of success

High

Technology gaps

Challenging/open issues

Low

Define reliable and secure sensors network for radiological and security
monitoring on-site (dose, cameras, etc.) considering the plant damaged
state.
Define adequate instruments to follow effective clean-up,
decontamination, fuel retrieval operations.

Likelihood of success: high, no raised concerns.

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APPROACH TO LONG-TERM POST-ACCIDENT MANAGEMENT AND ACTIONSTechnology and knowledge gaps: none.
Raised concerns: none.

Occupational health protection
Likelihood of successTechnology gaps

Challenging/open issues
Reliable monitoring of dose map in intervention zones, hot spots
detection and treatment in complex environment.

High

Low

Assess radiological risk and exposure associated to complex operations(clean-up, decontamination, defuelling, waste treatment and storage).
Plan interventions to minimise radiological exposure.
Monitor radiological exposure of workers.

Likelihood of success: the action is considered to have a high likelihood of success if proper
dose monitoring, radiological risk evaluation (see below) and planning of interventions are
made. Also, monitoring of radiological exposure of workers is an important aspect.
Technology and knowledge gaps: possibly, radiological risk assessment for complex
operations.
Raised concerns: none.

Emergency equipment maintenance
Likelihood of successTechnology gaps

High

Low

Challenging/open issues
Assess reliability of safety important equipment on the LT.
Equipment behaviour on LT in harsh conditions (dose, chemistry).

Likelihood of success: the action is considered to have a high likelihood of success if propermonitoring, repairing or replacing of equipment is ensured.
Technology and knowledge gaps: equipment behaviour on LT in harsh conditions (dose,
chemistry).
Raised concerns: none.

Risk assessment related to LT actions
Likelihood of success

MediumTechnology gaps

MediumChallenging/open issues
Little experience in the assessment of risks in complex situations on theLT, hence success probability is not known.
Risk assessment and mapping will change substantially during LT.
Non-trivial development of risk assessment tools may be required for LT.

Likelihood of success: risk mapping and implementing in response specific procedures,countermeasures and organisation is a challenge for complex situations. Risk mapping should
also follow plant damaged state changes (e.g. upon recovery actions) and consider LT actions
which may have different scope (maintain plant safety with damaged fuel in reactor or in SFP,
clean-up, decontamination, defuelling, waste treatment and storage).
Technology and knowledge gaps: many risk ranking methods exist (Section 3.4) but little
experience in application to LT management, with the exception of recent application to
Fukushima Daiichi NPP. Methods applicable to LTM should be developed.
Raised concerns: none.
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External hazards assessment and countermeasures
Likelihood of success

MediumTechnology gaps

MediumChallenging/open issues
Failure risk assessment for damaged or weakened structures andequipment contributing to LT plant safety (reduced margins).
Uncertainties concerning external hazards.
Assess margins recovery by countermeasures.

Likelihood of success: risk assessment for damaged or weakened plant and implementing
specific countermeasures (e.g. structure reinforcement, redundancy of emergency equipment)
is not straightforward. Characterisation of the damaged state and remaining margins to failure
is a difficult task. Further, recovery of margins set at the design stage may not be possible
despite countermeasures.
Technology and knowledge gaps: uncertainties remain on external hazards assessment.
Difficulty to assess margins reduction without a proper knowledge of structure or equipment
damaged state (difficult to attain). No tool to appreciate resistance of weakened structures andsystems.
Raised concerns: none.

Table 3.5. Maintain coolable configuration AIRT
ActionLikelihood of
success

Technology andknowledge gaps

Challenges and open issues

Inject water in-vessel

High

Low

Inject water incontainment
RPV pressure control

High

Low

High

Low

Maintain RPV integrity(for in-vessel melt
retention strategy)
Containment pressure
control
Containment hydrogen
control
Confinement of fission
products

MediumMediumHigh

Low

Long-term performance of equipment uncertain.
Back-up systems should be pre-planned
Long-term performance of equipment uncertain.
Back-up systems should be pre-planned
Plants should be equipped with reliable pressure
control systems
Knowledge of late-phase melt progression. Remaining
strength of RPV cannot be assessed. Materials
performance issues; embrittlement, corrosion
Related to hydrogen control, discussed below

High

Low

Radiolysis needs to be considered

MediumMediumMinimise environmental
releaseMinimise likelihoodof criticality
Maintain SFP coolingand minimise risk

MediumMediumHigh

MediumUncertainties regarding status of confinement,chemistry and remobilisation processes of fission
products in the long term affecting releases
Knowledge/technology gaps mainly concerningaqueous release.
Risk of re-criticality sometimes difficult to assess

158

High (loos of cooling) Medium
Low (loss-of-coolant)

Validated tools to calculate progression andconsequences of SFP accident

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Table 3.6. Waste water, solid waste and effluent management AIRT
Likelihoodof successTechnology andknowledge gaps

Minimise and collectcontaminated water

MediumLow

Decontaminate collected
waste water

High

Low

Recirculate decontaminatedwater for core coolingCollect and store
waste water
Monitor and control fission
products including tritium inwater
Control and minimise effluent
discharge
Effluent characterisation

High

Low

High

Low

High

Low

Low

MediumHigh

Low

Collect, store solid wastes and High
monitor fission products
content
Storage and disposalHighof filtered containment
venting system water and filter

Low

ActionChallenges and open issues
If containment is leaking, it is a considerable challenge tohandle contaminated water. Also, if cooling is not done on
secondary side, large amounts of contaminated water isgenerated from cooling
No technology available to separate tritium in very large
volumes and very low concentrations; capacity may be a
problem; water treatment facilities may not be readily available
as well as characterisation techniques (not easily detectable
nuclides)Damaged systems/piping may pose a problem

Low

The sheer volume of waste water may be problematic. Releasecriteria should be defined
Monitoring techniques may not be readily available for all
radionuclides
Monitoring of some not easily detectable nuclides
Knowledge of containment status (leak paths)
Technics/methods to recover the containment
Cf. challenges mentioned for fission products monitoring inwatersMonitoring of fission products content in highly active andcomplex solid structures
Accessibility and handling of highly active solids (cf. defuelling)
Monitoring of fission products content in highly active solid andliquid wastes (see above)
Accessibility and handling of highly active solids (see above)

Table 3.7. Site clean-up and decontamination AIRT
Likelihood of
success

Technology andknowledge gaps

Plant site clean-up and sludge
removal (outside of the reactor)
Surface decontamination

High

Low

Debris and sludge
characterisation

High

low

Monitor cooling pond
(and sea) contamination
Monitor and control
groundwater contamination
Radiological monitoring

High

Low

Low

low

High

Low

Characterisation of fission
products distribution incontaminated spaces

MediumMediumActionHigh (on-site)
Low
Medium (indamaged plant)

Challenges and open issues
If major releases have occurred, the likelihood of reachinggreen field status may be low
Accessibility may be limited due to contamination, physicalconstraints, water coverage. Contamination may havediffused into structures (e.g. in concrete).
Implementation of strippable or washable coatings inzones were high contamination is expected (e.g. incontainment)
Robotics and remote measurement technology would be
helpful.
Composition, content of burnable substances, etc. are alsoof interest for transport and disposal.
The sheer amount of debris may be a challenge
None
Specific technological solutions to limit groundwatercontamination are highly situation/plant specificTechnology is available; robotics and remote measurement
technology would be helpful
Methods for source distribution analysis; improved
techniques for hot spots detection and characterisation;difficulties to access highly contaminated areas; technology
is available except for molten fuel location detection; codes
cannot give this much detail

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Table 3.8. Defuelling of damaged reactors AIRT
Likelihoodof successTechnology andknowledge gaps

Detect damaged
fuel/fuel debris
locations andestimate masses (atstart of defuelling)

Low

High

Identify state ofdegraded fuel

Low

High

Collection ofdamaged fuel/fueldebris andcharacterisation
Damaged fuelexamination andcharacterisation forstorage
Minimiseenvironmental
releasesMediumLow

MediumMediumHigh

Low

ActionChallenges and open issues
High challenge particularly with failed RPV and/or large fuel dispersion.
No reliable information by imaging until access to damaged fuel ispossible.
Distant monitoring with sufficient resolution not yet developed.
SA codes not sufficiently predictive.
Required progress in technics and analytical tools scientifically challenging.
Selection of appropriate defuelling options and techniques delicate at startof defuelling.
Determining the composition, fission products content and mechanicalbehaviour of degraded fuel and debris without sampling and analysis ofreal material.
Predictive analyses using thermodynamic calculation tools.
Representative simulant materials testing.
Transposition of knowledge from performed fuel degradation
experiments.
Ageing of damaged fuel.
Specific remote control technologies may have to be developed
depending on damaged fuel/debris characteristics.
High dose, knowledge of debris characteristics.
Knowledge of damaged fuel characteristics should be obtained during fuelretrieval operation.
Specific tools may have to be developed for characterisation related tocriteria set for waste storage.
Depending on damaged fuel distribution, locations and characteristics
may be challenging to minimise releases during fuel retrieval operations.
Development of cutting techniques limiting aerosol generation anddispersion considering damaged fuel characteristics.

Table 3.9. Damaged fuel/fuel debris and radioactive waste
disposal, SFP fuel removal AIRT
ActionLikelihoodof successTechnology andknowledge gaps

Temporary on-sitestorage of damaged
fuel/fuel debris

High

Low

Temporary on-sitestorage of structuraldebrisTemporary on-sitestorage of filters
used in water
decontamination
Material control andaccountability

High

Low

High

Low

High

Low

SFP defuellingHigh (loss of Low
cooling)

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Challenges and open issues
Volume, dose and space restrictions on-site. More challenging for multiunits, SFP accidents with significant fuel damage, accidents withsignificant fuel dispersion.
May necessitate large-scale constructions (temporary confinement before
retrieval, installations for storage after retrieval).
Define criteria for site location and/or storage containers.
Define criteria for appropriate handling and storage.
Space restrictions may be a serious problem depending on debris
volumes.
Define criteria for appropriate handling and storage.
Dose and space restrictions if large volumes of contaminated water haveto be treated.
Define criteria for appropriate handling and storage.
Waste may need to be stabilised (limit remobilisation).
Define necessary level of detail for control and account.
Design adequate devices and methods for “routine” material control andaccountability particularly when large volumes have to be treated.
Complexity depending on extent of fuel damage in SFP, SFP building
damage (if external to confinement) and on-site radiological situation.
Identification, characterisation, collection and handling of mechanicallyweakened fuel assemblies, damaged fuel, fuel debris.
Damaged racks.

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Table 3.10. Long-term management AIRT
Likelihood of
success

ActionTechnology andknowledge gaps

Challenges and open issues

Radiological and security
monitoring on-site

High

Low

Define reliable and secure sensors network for radiological andsecurity monitoring on-site (dose, cameras, etc.) considering theplant damaged state.
Define adequate instruments to follow effective clean-up,
decontamination, fuel retrieval operations.

Occupational healthprotection

High

Low

Reliable monitoring of dose map in intervention zones.
Assess radiological risk and exposure associated to complex
operations (clean-up, decontamination, defuelling, waste
treatment and storage).
Plan interventions to minimise radiological exposure.
Monitor radiological exposure of workers.

Emergency equipment
maintenance

High

Low

Assess reliability of safety important equipment on the LT.
Equipment behaviour on LT in harsh conditions.

Risk assessment associated Medium
to LT actions

MediumAssessment of risks in complex situations on the LT.
Risk assessment and mapping will change substantially during LT.
Non-trivial development of risk assessment tools may be requiredfor LT.

External hazards
assessment andcountermeasures

MediumFailure risk assessment for damaged or weakened structures andequipment contributing to LT plant safety.
Uncertainties concerning external hazards.
Assess margins recovery by countermeasures.

MediumConclusions derived from the AIRT exercise results
The AIRT should serve as a filtering process looking at identifying the most critical challenges
and open issues related to the main actions carried out during the long term of the severe
accident. The priority taken by such challenges and issues should be in line with the importanceof the actions as hierarchically ranked in the exercise.
In order to limit the recommendations (Chapter 5) naturally aimed at closing the identified
issues and gaps by orienting future efforts, a qualitative threshold should be established. As
stated above in Section 3.5, such barrier has been placed at the “medium” rank level,
i.e. challenges and open issues will be identified only coming from those actions assessed as“medium” or “low” in terms of the likelihood of success, and/or as “high” or “medium” in termsof gaps. And within these actions, challenges and open issues linked to those featuring L/H
combinations should deserve a distinct, more careful treatment.
From the 6 LTM-goal-oriented tables, there are only 3 out of 42 actions (7%) featuring an L/H
combination, all of them under the defuelling goal:•

detect fuel debris location and estimate debris masses;•

identify the state of the degraded fuel;•design and implement defuelling alternatives.

However, most of the actions present at least one M value whether in the likelihood of
success or gap side, in particular, 21 actions from the 6 tables, meaning 45% of the entire list:•

maintain RPV integrity (medium in technology/knowledge gaps);•

confinement of fission products (medium in technology/knowledge gaps);•

minimise environmental release (medium in both ranks);•

minimise likelihood of re-criticality (medium in technology/knowledge gaps);•

maintain SFP cooling and minimise risk (medium in technology/knowledge gaps);

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minimise and collect contaminated water (medium in likelihood of success; high intechnology gaps);•

decontaminate collected waste water (medium in both ranks);•

control and minimise effluent discharge (low in likelihood of success; medium intechnology/knowledge gaps);•

collect, store solid wastes and monitor fission products contents (medium in likelihoodof success);•plant site clean-up and sludge removal (medium in technology/knowledge gaps);•

surface decontamination (medium in technology/knowledge gaps);•

debris and sludge characterisation (medium in both ranks);•

monitor and control groundwater contamination (low in likelihood of success; medium
in technology/knowledge gaps);•

characterisation of fission products distribution in contaminated spaces (medium inboth ranks);•collection of damaged fuel/fuel debris characterisation (medium in both ranks);•

damaged fuel examination and characterisation for storage (medium in technology/
knowledge gaps);•

minimise environmental releases during the defuelling process (medium in both ranks);•

risk assessment associated to LT actions (medium in both ranks);•external hazards assessment and countermeasures (medium in both ranks).

It is worth noting that the conducting efforts should be oriented according to the type ofidentified issue/gap. For instance, medium or low likelihood ranks can just derive from the lackof dedicated arrangements on-site. If this were the case, efforts on how addressing the issue
should not be oriented towards further research and development but on providing
recommendations to the plants to accommodate the necessary measures allowing the action to
be suitably performed in place. Therefore, it will be fundamental to link the challenges/ open
issues with appropriate recommendations in conducting future activities aimed at closing thatissue/gap. Among others, Chapter 5 will address these recommendations taking as one of thesources of information the AIRT results coming from the 24 actions ranking medium/low inlikelihood of success and/or high/medium in technology gaps as listed above and upon tracing
back the underlying challenge/open issue.

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Chapter 4. Summary of challenges and main issues
for long-term management

Significant challenges, open issues and knowledge and technological gaps for the long-term
management (LTM) of severe accidents synthetised in this chapter were identified through the
feedback of the Three Mile Island unit 2 (TMI-2), Chernobyl and Fukushima Daiichi accidents(Section 2.1), through the review of the status of post-accident long-term management andactions (LTMA) in NEA member countries (Section 2.2) and through the various approaches to
LTMA (Chapter 3).
Though the TMI-2, Chernobyl and Fukushima Daiichi accidents posed tremendous
challenges for LTM, each accident was different in nature and had different consequences. Thus
challenges and issues may be different for each of them. TMI-2 LTM can be considered ascompleted as the fuel has been transferred from the site to long-term storage facilities whereas
Chernobyl and Fukushima Daiichi LTM are ongoing with fuel retrieval to come. Thus for these
two accidents the learnings for LTM will surely evolve in the years to come.
In comparison to the TMI-2 accident where the reactor vessel, the primary circuit, and theconfinement integrity were maintained after the accident, other challenges arose for the LTMof the Chernobyl and Fukushima Daiichi accidents. This was due to highly damaged
infrastructures caused by explosion and fires at Chernobyl, and by earthquake, tsunami,
explosion, and fires at Fukushima Daiichi, loss of confinement due to explosion at Chernobyl
and mostly due to containment over-pressurisation at Fukushima Daiichi and consequent large
release of contamination. Among these challenges, extensive site cleaning and construction
works had to be performed to reduce on-site radiation levels, facilitate damaged plant
accessibility and protect the environment (hydrosphere, atmosphere) from further radioactive
releases (e.g. sarcophagus at Chernobyl, sea and land walls, groundwater bypass systems at
Fukushima Daiichi). These extensive works were largely specific to the plant damaged state andsite, particularly works to protect the hydrosphere.
Though the confinement had been maintained at TMI-2, the accident posed significant LTM
challenges with unprecedented actions, organisations, and technical means (infrastructures,
systems, equipment and instrumentation) to recover and treat highly contaminated waters,
decontaminate highly contaminated buildings (in particular the reactor building where most of
the radioactivity was confined), defuel a severely damaged reactor core and manage transportand storage of unique wastes.
In the sections below, the main challenges and issues that were highlighted in the TMI-2,
Chernobyl and Fukushima Daiichi accident feedbacks, in the information provided by utilities
and safety organisations through a questionnaire as described in Section 2.2, and through theplant damaged state (PDS) classification and action identification ranking table (AIRT) exercises
are summarised.

Regulatory and organisational aspects
The feedback from past accidents showed that unique regulatory and licensing requirements
had to be developed for LTM at TMI-2 and Fukushima Daiichi 1 with necessary “flexibility”
compared to existing regulation. New regulatory practices had to be applied as changes in the1.

Little feedback was provided on these aspects for Chernobyl LTM.

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facility’s post-accident mode of operations required unique regulatory and licensing actions. In
order to properly reflect evolving plant status, the regulators issued orders, modified thoseorders, approved licence amendment requests, and granted relief from certain regulatory
requirements. Exemptions were necessary because of the plant’s damaged configuration andchanging status during clean-up. This had to be done for very different LTMA with sometimes
limited knowledge of plant status and risks for their implementation, with sometimes use of
unique systems and equipment and involving objects difficult to fit in regulatory classification(e.g. degraded fuel, radioactive wastes).
Not only the regulatory requirements, but also the organisational structure and functions of
the utilities and their support organisations changed during the long-term management of theaccidents. Various organisations formed working groups to provide guidance on addressing
specific issues, problems, and research activities. Independent oversight groups reviewed,
monitored, and advised on the overall direction of recovery and clean-up plans and activities.
Also, both at TMI-2 and Fukushima Daiichi, international collaborative activities were
established in support of the long-term management and decommissioning of the plants.
Concerning plant staff, long-term management programmes are intense and staff working
conditions should be considered in long-term recovery activities and staff turnover should beplanned where appropriate.

Relation to the public during long-term management
Increased public scrutiny after an accident was reported as an important issue both for the
TMI-2 and the Fukushima Daiichi accidents where organisations dedicated to relations to the
public were established and public relations plans were made as described in Section 2.1. Even
with extensive efforts, gaining public trust was proven to be a significant challenge. In manycases it was more difficult to obtain public trust and acceptance than to find a technical solution.

Risk evaluation for long-term management
For all accidents, it has been reported that risk assessment for LTM is a challenge due to lack ofknowledge on plant damaged state when implementing some LTMA, e.g. damaged fueldistribution and characteristics when starting fuel retrieval at TMI-2 and Fukushima Daiichi,
and its evolution during LTM. The need of continued risk assessment during LTM implementing
new knowledge on plant damaged state for risk-informed LTM is highlighted, particularly for
Fukushima Daiichi LTM. Risk assessment is further complicated by lack of experience on risk
assessment in the long-term management of severe accidents, and the lack of validated
methods to apply to this task. Some approaches to the risk assessment have been discussed in
Chapter 3 and, in connection with the Fukushima Daiichi experience, in Section 2.1. It wasrecognised that further work is needed to develop methods capable to effectively guide longterm management.
More specific aspects have been highlighted from LTM feedback as well as analyses
presented in Chapter 3:

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LT reliability of infrastructures and of some important safety systems, e.g. failure of
weakened infrastructures related to ageing or external events may result in a new
unstable state. Examples of this are the non-stabilised first sarcophagus built in
Chernobyl to protect the damaged reactor, and the reactor buildings of units 1, 3 and 4 at
Fukushima Daiichi before consolidation works.

•

Situations posing increasing challenges with time should be particularly assessed,
e.g. ageing of the first sarcophagus at Chernobyl, waste accumulation at Fukushima
Daiichi, leaching and ageing of damaged fuel material and structures at Chernobyl and
Fukushima Daiichi.

•

Assessment of risk of criticality.

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The possibility of generation of combustible gases.

•

The risk of release of radioactive compounds from damaged fuel due to leaching andageing, and release of fission products from surfaces and water volumes due to ageing
as well as de-commissioning activities.

Both at Chernobyl and at Fukushima Daiichi, the present state of the fuel-containing
materials is not sufficiently known which creates nuclear and radiation risks for defuelling
works.

Radioprotection in long-term management
One of the primary objectives for LTM is to improve working conditions by reducing dose rates
on-site by simultaneously avoiding further release of activity or increase of the dose rates. The
challenges come from the unknown status of the plant and the unknown distribution of theradionuclides. The difficulty to acquire sufficiently accurate dose rate mapping that could guideradioprotection implementation for LTMA and the search for improved data acquisition andanalyses methods were highlighted for the past accidents and identified as a knowledge andtechnological gap in the AIRT exercise.
The work is challenging due to high dose rates in the contaminated areas, and thepotentially severely damaged structures. The dose rate reduction is done, as far as feasible, inthe damaged plant(s) by cleaning, fixing or removing contamination and by implementing
protections wherever necessary. Different approaches have been implemented effectively
easing progressively interventions in more and more areas in the damaged plants. In manycases, dose reduction by shielding has proven to be more effective than decontamination. The
use of robotics to perform monitoring and decontamination is invaluable in reducing radiation
exposure to workers and in allowing operations in areas not accessible to humans, as evidenced
by the feedback from past accidents as well as highlighted in the AIRT exercise.
To manage radioprotection, specific radioprotection organisations were established at
TMI-2 and Fukushima Daiichi. As low as reasonably achievable (ALARA) approach was set as an
objective. Many programmes and activities were established to help improve radiation
protection practices. New approaches were needed in a number of basic worker protection anddose reduction areas, including protective clothing, respiratory protection, dosimetry, radiation
field and contamination characterisation, exposure-tracking systems, dose reduction planning,procedures, and training.
It should be emphasised that at Chernobyl and Fukushima Daiichi, damaged plants are stillfacing considerable radioprotection challenges related to future defuelling and decommissioning
actions. A significant challenge will be to assess and mitigate the risks of contamination
dispersion related to these actions. Also, difficult configurations in terms of accessibility anddoses may pose radioprotection challenges.

Coolable configuration and confinement of radioactivity
The three past accidents in TMI-2, Chernobyl and Fukushima Daiichi had very different accident
progression and end state of the reactors. Therefore, we give a short summary of each.
In TMI-2, the reactor pressure vessel was intact and this enabled coolant water injection
into the reactor even though the pressure control did not function and non-condensable gases
needed to be released not to block the heat removal. The state of the core was not known until
first visual observations were made and it was seen that the core was severely damaged. As thestate of the core was not known, a high boron concentration was maintained in the coolant toensure that the core would not become critical.
The main challenges to the confinement of radioactivity in the long term were the large
quantities of contaminated water in the reactor building, and the release of fission gases fromthe damaged fuel. The fission gases caused airborne contamination inside the auxiliary building

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and measures were taken to remove them. A significant risk of release of radioactivity to theenvironment in the long term was related to the large quantities of contaminated water in thereactor building. Water management was one of the most challenging activities in the longterm management of TMI-2.
In Chernobyl, the explosion and fire destroyed the structures designed to cool the reactor
core and to confine the radioactivity. A sarcophagus was erected within six months of theaccident to limit the release of radioactivity mainly to the atmosphere. However, thesarcophagus was not entirely leak tight. Cooling of the corium inside the sarcophagus was
carried out by installing sprays which also reduced the amount of airborne radioactive aerosols
formed due to ageing of the corium materials. Contamination release through water pathways
was controlled by building a plate under the reactor, and by several measures to limit the flowof water to the nearby rivers.
In Fukushima Daiichi, the reactor pressure vessels and containments in all three damaged
units have been observed to be compromised thereby providing a path for release of radioactive
compounds. A closed-loop cooling has been established to control the releases with water. All
three containments are filled with water even if to a different level, and a large fraction of thecontamination in the containments is expected to be in the water thereby controlling the air
releases. Continuous coolant injection keeps the temperatures in the reactor pressure vessels
and containments at low levels.
Even though very different in progression and end state, all accidents described here had
similar challenges related to cooling the reactor core and confining radioactivity. The AIRT
exercise especially highlighted the challenges related to the unknown state of the reactor
pressure vessel (RPV) in the long run and material performance issues, as well as to confinementof aqueous releases and uncertainties regarding remobilisation of radioactivity as elaborated
below.
First and foremost, the state of the core was not known when long-term management
actions had to be designed and decided. This indicates that both analytical and measurement
methods are incapable of determining the accident progression and its end state accurately
enough as also emphasised by the AIRT process. Only after visual observations and samplingsof the active materials, it was possible to determine the state of the core in TMI-2. In Chernobyl
and Fukushima Daiichi, samplings have been carried out and they have given indications of theextent of the damage, but with the lack of instrumentation capable of working under theconditions prevailing in the reactors, there are still uncertainties concerning the state of thecore and distribution of the fission products.
Secondly, there is not much knowledge about the processes that take place in the long term
under humid, high irradiation, possibly high temperature atmosphere. Structures which have
kept the coolant flowing and confined the radioactivity may deteriorate with time. This meansthat long-term accident management measures need to be implemented taking into accountthat the status of the plant is changing, but it is difficult to predict these changes.
And third, it was not possible to determine which systems in the plants were operable andcould be used in the long-term management of the accidents. Therefore, long-term management
actions had to be planned using systems without always knowing if those systems would work
and without instrumentation to support these actions with measurements.
It should be emphasised that these issues and challenges are not accident progression
specific and therefore would probably exist for most severe accidents as highlighted by the AIRT
process in Chapter 3.

Development and use of unique systems, equipment and instrumentation
Unique systems, equipment and instrumentation had to be developed for LTM of the three
accidents. Important cited aspects are not always common to the three accidents:•

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robotic or remote control means for decontamination, for investigations in harshenvironment (imaging, monitoring, sampling) and for defuelling;

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sampling systems for contaminated waters and active debris and related analyticalmeans;•

instrumentation for detailed and accurate radiation monitoring;•

methods for determining the distribution of degraded fuel, e.g. muon tomography at
Fukushima Daiichi;•

systems and instrumentation for monitoring sub-criticality, hydrogen and radioactive
dust;

•

liquid waste treatment facilities and related analytical means;•transport canisters for solid wastes;•

temporary radioactive waste storage facilities for low, medium and highly active liquidand solid wastes.

LTM of the three accidents highlights the importance of containment media and analyses, ofmonitoring critical data (to avoid reaching a new unstable state) and instrumentation providing
reliable information for a risk-informed LTM. The monitoring of sub-criticality remains achallenge, even though the detection of re-criticality events is possible. The importance ofdeveloping further remote or robotic techniques of investigations to guide LTMA should be also
highlighted. These technological gaps were also pointed at in the AIRT exercise which indicatedthat available monitoring and robotic technologies are not always compatible with conditions in
the vessel and/or containment due to high radiation and high temperature, and that there are
improvement needs for imaging and distant monitoring in high radiation zone.

Contaminated water treatment
Common to the three past accidents was the management of large amounts of highly
contaminated liquids. This required the design and construction of treatment facilities toeliminate, through filtering systems, Cs, Sr, and other radionuclides. At the same time, the water
treatment facilities generated highly active solid wastes in the form of filters and sludge.
Tritium cannot be removed from the liquid using standard liquid waste treatment. In TMI-2,
water contaminated by tritium was evaporated. In this way any contamination of nearby riverby tritium was avoided and release of residual activity of other radionuclides remaining in thewater was avoided. Due to the use of sea water during the emergency phase of the accident at
Fukushima Daiichi for cooling of damaged reactors, specific separation processes had to be
implemented to eliminate chlorine from contaminated waters. Information about water
treatment systems for TMI-2 and Fukushima Daiichi are provided earlier in this report.
Management of large amounts of highly contaminated liquids would probably be a challenge
for most severe accidents as highlighted by the PDS classification (Section 3.3) and AIRT
(Section 3.5) exercises.
Leaching of the fuel and fission products would also pose for most severe accidents a
specific challenge to the water treatment systems. At Fukushima Daiichi, the reactors arecontinuously cooled with water, and some fission products have been found to be transferredto the coolant water. Due to groundwater intake in the damaged reactors, contaminated waters
are posing specific problems at Fukushima Daiichi with accumulation of very large volumes thathave to be stored on-site. A lot of efforts have been implemented to reduce the produced
volumes: implementation of close loop cooling, of sea and land walls, and of groundwater
bypass.
At Chernobyl, water in the sarcophagus presents increasing Pu concentrations which may
be a sign of leaching and ageing effects. The water originates from condensation inside thesarcophagus, sprays used to limit dust dispersion, and leakage of rain water into thesarcophagus. Extensive efforts were made over the years to successfully limit the release ofradioactivity to the groundwater and the surrounding rivers.

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Site clean-up, de-contamination and waste management
In past accidents, different strategies have been implemented to reduce as much as feasible theradioactive inventories and risks of further radioactive release.
For instance, at TMI-2 the approach was progressive with first recovery and treatment of
the highly contaminated waters to give access to the plant buildings, followed bydecontamination of the auxiliary and the reactor building to give access to RPV and damaged
fuel for defuelling.
At Fukushima Daiichi, many aspects had to be treated in parallel such as recovery andtreatment of large volumes of contaminated waters, cleaning of rubbles from buildings
damaged by the explosions, decontamination of the reactor building above the refuelling floor
and of the refuelling floor, fuel retrieval from spent fuel pools (SFPs), etc. Access to primarycontainments remains a challenge in the end of 2017.
At Chernobyl, the prime objectives were to re-establish a confinement by building the firstsarcophagus within six months, and to ensure its safety. There too, access into rooms
containing damaged fuel and corium remains a challenge in the end of 2017. Even thoughsamples have been taken of the fuel-containing materials, de-fuelling has not started yet.
As shown by the Fukushima Daiichi feedback and through approaches to LTM described in
Chapter 3, risks identification and ranking could efficiently guide LTM strategies, but related
tools and methods require development and validation.
At TMI-2 and Fukushima Daiichi, decontamination of highly contaminated surfaces wasdone using many different techniques (chemical or mechanical treatment) and sometimes by
robotic means. Cleaning and decontamination operations were reported to be challenging
especially where contamination had penetrated deeply into the contaminated material, inparticular in concrete and in some areas with low accessibility. Then fixing contamination,e.g. by spraying anti-scattering agents, or implementing protections and shielding where
necessary are alternatives that have been used to decrease radiation level and provide safe
working conditions. This should be a challenge for most severe accidents as highlighted in
Chapter 3.
In common to the three accidents, radioactive wastes had unique characteristics not fitting
regular waste classifications and technologies for reprocessing, transport and disposal.
Therefore, specific equipment, containers and installations for waste treatment and temporary
waste storage had to be constructed within limited amount of time to manage the large
quantities of radioactive waste. Even though AIRT process estimated a high likelihood of successto these actions, it identified challenges due to space restrictions, time management, andselection of criteria for a suitable treatment method due to the unknown characteristics of thewaste.
For TMI-2, special design was developed for fuel canisters, which were then suitable for thetransfer and storage using filling material to stabilise the fuel-containing solid waste.
Solutions for temporary and final waste storage have yet to be established for Chernobyl
and Fukushima Daiichi.

Fuel retrieval
Regarding the fuel retrieval from the damaged reactors, all three past accidents have very
distinctive characteristics. In TMI-2, the reactor pressure vessel remained undamaged, and fuelretrieval could be carried out from the top of the reactor vessel. In Fukushima Daiichi, thereactor pressure vessels as well as primary containments in all three damaged units arecompromised, and some fraction of the corium is expected to be on the containment floor withpossible molten corium-concrete interaction (MCCI). Therefore, fuel retrieval will be more
complicated. In Chernobyl, the explosion destroyed the confinement, and the fuel is spread toseveral rooms both vertically and laterally.

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SUMMARY OF CHALLENGES AND MAIN ISSUES FOR LONG-TERM MANAGEMENTAt TMI-2, the fuel retrieval strategy and equipment had to be revised when information was
gained on damaged fuel status after RPV opening and investigations in the RPV. The fuelretrieval was completed with specific equipment (defuelling platform) for which exercises were
performed beforehand to limit workers exposure.
At Chernobyl, the fuel retrieval strategy is yet to be designed and planned. The degradedfuel material distribution and characteristics were studied in detail but it is recognised thatageing, mostly through leaching in water and interactions with atmospheric gases, may affectthe integrity of the damaged fuel with possible fine dust formation on the long term. The longterm behaviour was investigated but the knowledge is not sufficient to assess the risk of finedust formation. The new sarcophagus was designed to avoid release of radioactivity to theenvironment during fuel retrieval actions.
At Fukushima Daiichi, the fuel retrieval is planned to be started around 2021. It isemphasised that Fukushima Daiichi presents an added complexity with respect to TMI-2 since
it is a three-unit accident with different accident progression in each unit. Consequently, theoptimal fuel retrieval strategy may be different for each unit. It is noted that presently, someuncertainty still exists regarding the amounts of material in the reactor pressure vessel and in
the containment, as well as regarding the extent of MCCI in the three units despite
investigations conducted so far (analytical investigations, muon tomography, robotic
investigations in the different units). Important efforts are planned from 2018 to increaseknowledge on degraded fuel for preparation of plans for fuel retrieval operations.
As highlighted in Chapter 3, fuel retrieval from damaged reactors or SFPs after a severe
accident should remain a tremendous challenge as long as there are remaining knowledge andtechnology gaps which prevent obtaining information that could effectively guide defuelling,
before direct observations and characterisation of damaged fuel are possible. In addition, further
development of remote control technologies was emphasised also for these actions.

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Chapter 5. Recommendations for enhancing long-term post-accident
management and for future researchThis chapter presents recommendations for improving the long-term management (LTM),
arranging the recommendations upon whether they address general, cross-cutting issues
covering all aspects of LTM, or specific LTM actions.
Recommendations are related either to necessary knowledge development and consolidation
to enhance LTM and are then addressed to research organisations, safety organisations 1 andindustry (operators and designers), or to development of accident management provisions andthey are then addressed more specifically to plant designers and operators.
Organisations such as the Nuclear Energy Agency (NEA), the European Commission (EC) andthe International Atomic Energy Agency (IAEA) could foster international initiatives aiming at
knowledge development and capitalisation.
It should be emphasised that proposed recommendations are based on LTM feedback forpast major accidents (Section 2.1 and Chapter 4) and on risks, challenges, issues and knowledge
and technological gaps identified through discussed approaches to LTM (Chapters 3 and 4).

5.1. Recommendations related to knowledge consolidation for cross-cutting issues
Calculation tools and methods for severe accident analysis in reactor and spent fuel pool
Consolidation of reactor and spent fuel pool (SFP) severe accidents knowledge base and calculation
tools (e.g. severe accident (SA) codes, thermochemical databases) should be pursued in the future toenhance capabilities to predict i) the effect of mitigation measures used to reach a stabilised state andii) the stabilised plant damaged state following an accident, e.g. distribution and properties ofradioactive material in- and ex-vessel in a damaged reactor.
Methods to validate reactor and SFP accident mitigation strategies to reach a plant stabilised state,including strategies using non-conventional approaches and equipment should be consolidated.
These research efforts should be targeted at providing information, data and methods to enhancevalidation of reactor and SFP accident mitigation strategies, addressing in particular scaling issues
and uncertainties for main reactor designs.

LTM of past SAs, including the ongoing LTM for the Fukushima Daiichi accident, hasevidenced that the existing knowledge base and calculation tools for reactor and SFP SA are notsufficiently developed and validated to provide some critical information on the stabilised plantdamaged state after an accident that would support LTM implementation (e.g. containment
status, distribution and properties of radioactive material in the damaged plant). Thoughadditional information and data should be obtained and verified after reaching a stabilised state,
LTM would greatly benefit from more predictive tools.

1.

Safety authorities and their technical support organisations.

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Further, positive and negative effects of some accident mitigation measures, e.g. oncomponents, equipment, systems and structures, remain difficult to assess with existingknowledge. Failure or success of these mitigation measures will in great part determine theplant damaged state after stabilisation and risks and challenges for LTM.
This is particularly true for example for strategies and actions involving water injections invessel, ex-vessel or in SFP in view of recovering the cooling of the degraded fuel and strategies
involving pressure control of the containment (e.g. venting) as these actions, if improperly
implemented, generate risks to lose the confinement of radioactive material and of increased
radioactive releases in the LT. For instance, water injections can either result in positive(e.g. effective degraded fuel cooling in or ex-vessel, in SFP) or negative effects when applied
inadequately.
The assessment of confinement failure risks and associated releases is part of SA analysis.
However, an accident with failure of the confinement will obviously lead, due to contamination
by radioactive releases on-site, to a challenging LTM with complex recovery actions. Even after
the plant has been stabilised, it may be a continuing challenge to control and limit further
releases particularly through contaminated water transfers to the environment and avoid
groundwater contamination. As learnt from the Fukushima Daiichi accident, this may lead to theneed to implement highly technically complex and costly measures to limit radioactive liquids
releases and manage large amount of liquid wastes in the long term.
After the Fukushima Daiichi accident, significant efforts have already been internationally
undertaken to enhance some SA mitigation strategies. However, validation of these strategies
should still be examined with:•full consideration of past accidents knowledge, including knowledge that will beobtained through post-Fukushima Daiichi accident projects (e.g. NEA Benchmark Studyof the Accident at the Fukushima Daiichi Nuclear Power Station, PreADES, TCOFF and
ARC-F and future post-Fukushima Daiichi LT projects);•

full consideration of knowledge obtained through major international research
programmes aiming at severe accident management guideline (SAMG) validation(e.g. the EC H2020 In-vessel Melt Retention [IVMR] project);•

increased knowledge and feedback on implementation of non-conventional approaches
and equipment;•

further consideration of site and plant specificities and expected resources inchallenging situations.

Additional research efforts that are being conducted within NEA and EC initiatives for the
consolidation of SA mitigation strategies (e.g. in-vessel or ex-vessel melt retention, andradioactive release mitigation) and of SFP accident mitigation should be supported. Such
consolidation, implementing in particular knowledge gained through LTM of the Fukushima
Daiichi accident, will necessarily enhance severe accident management including for LT phases.

Status of equipment, systems and structures in the long term of a severe accident
Knowledge of status of equipment, systems, including passive ones, components and structures in
the long term of an SA should be consolidated with emphasis on those that are expected to contribute
maintaining a stabilised state on the LT.
Development of harmonised methods should be fostered at the international level to assess response
and reliability of systems, including passive ones, equipment components and structures during an
SA considering LTM.

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With recognition that progressing in the field may be a complex task due to the variabilityof reactor designs and of their systems and equipment, shared common knowledge base andapproaches would be highly valuable for safety assessment processes of nuclear power plants(NPPs). Such an action could also result in proposing further industrial developments for
systems and equipment better qualified to prototypic SA conditions.
In particular, the capacity to assess reliability of cooling systems, including nonconventional ones, and status of confinement (leakages paths and rates and their evolution) on
the LT is of critical importance for LTM.
Uncertainties around the containment status are a major issue, particularly whencontainment pressure and temperature loads have resulted in containment failure andsignificant radioactive releases to the environment. With respect to knowledge of theconfinement leak tightness, further testing and characterisation for prototypic SA conditions offlanges and seals used to isolate containment openings, of concrete walls, of liners and coatings
would certainly be valuable to determine most probable releases paths to the environment andwhere one can expect main radioactivity deposition in the damaged plant. The ageing ofstructures and equipment could as well be considered in such investigations. Materials,
conditions and configurations to be tested could however be very containment design-specificand defining generic approaches to gain knowledge and develop methods to assesscontainment leakages and their evolution with time should be undertaken as recommended.

Severe accident long-term phenomenaInternational sharing and capitalisation of knowledge on phenomena that could affect LTM should be
fostered to define appropriate future research on such phenomena.
Appropriate research should be conducted to acquire sufficient knowledge for LTM and related risk
evaluation for re-criticality, hydrogen production, phenomena that could induce mechanical failureof safety important components (with consideration of cliff-edge effects) such as corrosion-erosion
reactions, sediment formation and clogging of recirculation loops, solidified corium and debris
leaching, ageing and fuel dusting and dispersal (in particular during debris retrieval operations).
Such actions should benefit from knowledge sharing between various communities (SA, ageing andwaste management experts).
Further development of knowledge on some of these phenomena through investigations on real
samples from passed accident should be supported, e.g. debris leaching and ageing from Chernobyl
accident (some testing may be performed in the NEA TCOFF project). These investigations should also
provide important knowledge for Fukushima Daiichi LTM.

LTM of an SA after damaged plant stabilisation could certainly be improved with betterknowledge of long-term processes affecting systems, equipment and structures (e.g. damaged
or weakened reactor pressure vessel [RPV], containment, SFP) but also radioactive products
deposits and solidified corium and fuel debris. This is of special importance for risk evaluation
related to LTM actions (cf. next section), including recovery and decommissioning actions. It has
already been identified that knowledge of effects of corrosion and radiolysis reactions andrelated hydrogen production, of long-term radiation-induced damages, of fuel debris leachingand exposition to humidity and of radiobiological processes is scarce. From Chernobyl
observations on lava ageing, it can be expected for instance that LT radiation-induced damages,leaching and/or exposition to humidity and radiobiological processes may affect solidified
corium and fuel debris mechanical properties and their retrieval.
Knowledge sharing and capitalisation on long-term phenomena could be partly tackled in
short-term projects which will be launched to prepare the decommissioning of the Fukushima
Daiichi damaged reactors (e.g. NEA PreADES, TCOFF and ARC-F projects) with the objective to
identify additional research that should be undertaken in the field and in which frame. From
now on however, some support to further investigations on Chernobyl samples could beorganised.

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Methods or systems for risk assessment for long-term management
Opportunities to launch at the international level a joint effort to progress in methods and/or expert
systems for risk evaluations for LTM and LTM optimisation should be fostered.

Such an action would certainly benefit from a common work between risk and safety
analysts from safety organisations and operators, taking full advantage of past accidents LTM
feedback.
Expected results are methods and criteria for LTM optimisation in terms of safety to reachsite remediation.
Conducting a risk-informed LTM of a damaged plant is necessary but challenging at thispoint since, as earlier pointed, i) knowledge of the stabilised plant damaged state and remaining
safety margins for critical components is limited as well as their evolution with time andii) knowledge and existing analytical tools are not developed to assess effects of long-term
phenomena. In addition, methods and tools for risk mapping and ranking for LTM, tracking
evolution of the stabilised plant damaged state and considering substantially different
objectives of recovery actions (e.g. maintaining a stabilised state, managing wastes, cleaningand decontaminating accessible areas, retrieving fuel from SFP or from damaged reactors, etc.)
and nature of risks are not well established. Possible approaches have been discussed in thisreport but they should be reviewed and developed further for concretely supporting long-term
management and actions (LTMA) implementation.
In addition to the development of methods for risk assessment, there could be some interest
to develop methods and criteria for LTMA optimisation in reaching site remediation.

5.2. Recommendations related to provisions development for cross-cutting issues
Enhance monitoring of plant damaged state and of long-term management actions
Further joint investigations to better define critical data and instruments (existing or contemplated)that would effectively guide management of all phases of an SA including LTM and to promote thedevelopment of innovative approaches and technics should be fostered.
This should include, more specifically for LTM, robotic means and remote technologies. Enhanced
technologies for direct or distant imaging and more accurate radioactivity and dose mapping in harshenvironment would strongly support LTMA implementation.
Provisions to enhance monitoring of both short-term SA progression and related releases and LTM
should be developed.
These should in particular include instrumentation dedicated to provide information on degradedfuel/debris distribution in-vessel and ex-vessel (at least detection of vessel rupture during theaccident), to provide information on containment leakage paths, to monitor re-criticality risks forvarious degraded fuel/debris configurations, to monitor combustible gas contents in premises wherecombustion could affect plant safety, to monitor atmospheric and liquid radioactive releases andradioactive products deposits in the damaged plant.

Large efforts have been undertaken in the past on instrumentation for normal operation
and accidents in the emergency operating procedures domain but less on SA instrumentation.
In the SA field, particularly after the Fukushima Daiichi accident, the focus of international
attention was on consolidation of existing techniques and systems that support
implementation of mitigation strategies on the short term, looking particularly at improved
methods for better assessing their capability for SA conditions.

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While this is of importance, less effort seems to have been undertaken internationally i) to
identify if major possible monitoring improvements that would efficiently support
comprehensive management strategies for all phases of an SA in a reactor or in an SFP –including emergency planning and LTM of damaged plants – are possible, ii) to identify available(including simple, robust and remote technologies), under development or promising newtechnologies that could provide such major improvements, and iii) to conduct research thatwould help better assessing critical instruments (existing or contemplated) performance in SA
and promote development of innovative approaches and technics.
It should be emphasised that the research in this field should benefit from technologies bothfrom the nuclear and the non-nuclear fields and should consider the development of fixed andmobile systems and their protection, including remote technologies and of robotically operated
systems.
The following actions would help defining a well-targeted research programme to better
assess critical instruments performance in SA and promote development of innovative
approaches and technics:
1. Review of critical information for severe accident management, emergency planning,
LTM (up to decommissioning after an accident) with required criteria (range, accuracy,
response and mission time) – based on approaches developed or under development for
NPPs worldwide – that would efficiently support comprehensive management strategies(both for reactor and SFP accidents). With consideration of 2), identification of major
potential improvements (additional supportive critical information and relevantmeasuring technics) for different rector designs.
2. Review of already implemented instruments (including design basis accident
instruments), their expected performance in SA (level of qualification, reading validity,
margin to failure) and their possible use in severe accident management, emergency
planning and LTM approaches for NPPs. And also identification of additional needed
research, if any, to better assess their performance.
3. Identification of available technologies and/or technologies under development and/or
innovative technologies that would provide these major improvements, considering
possible implementations, protections and expected environmental conditions andcovering all accident phases.
LTM could largely be improved with a reliable instrumentation providing pertinent
information to characterise and monitor the stabilised plant damaged state and its evolution
with recovery actions, to monitor the performance of systems and equipment – including backfitted and non-conventional ones – involved in maintaining a stabilised state on the LT and tosupport the optimisation of LTM actions and their risk monitoring, including for clean-up,
decontamination, fuel retrieval and decommissioning operations. Identified critical
instrumentation is earlier listed with the recommendations.

Upgrades of components, equipment, systems and structures
Provisions to enhance reliability and availability of systems, equipment, components and structures
should be implemented, including for LTM. Replacement, repairing and maintenance feasibility
wherever contemplated should be assessed.
Provisions for post-accident samplings (contaminated liquids and atmosphere) in the damaged plant,
retrieval, transfer and analyses should be prepared.

Numerous equipment and systems upgrades have already been proposed or implemented
to various extents by operators in NPPs worldwide after the Fukushima Daiichi accident as aresult of international stress tests to reinforce severe accident management. Some of these
upgrades should be beneficial for LTM, in particular those implemented for strengthening
mitigating systems as e.g. mobile equipment in common protected storage in different
countries.

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Some general guidance for upgrades – including for LTM – has already been provided ininternational documents (see for instance ENSREG [2012] and NEA [2014a]):•Components, systems and equipment should remain operable to maintain a stabilised
state for an extended period (several months or more) or they should be made easily
replaceable. They should also be resistant or protected against on-site and off-site
hazards. This concerns both fixed and mobile systems and equipment including passive
systems.

•

Provisions should be made to maintain mitigation measures for long-term phases of an
accident. Due to the contamination, it may be difficult to repair mitigation means.
Therefore, easily implementable substitution means should be provided.

•

Sampling systems and analytical means for plant surveillance and following recovery
actions should be provided.

•

Treatment, storage and management of radioactive wastes (solid, liquid, gas) should beconsidered.

This complements guidance for strengthening mitigating systems for accidents such as the
Fukushima Daiichi accident:•

Management and operation of mobile or non-conventional equipment should be welldefined (storage, maintenance, testing, transfer and training) and should consider
possible large contamination. Strategies for deployment in the event of anticipatedhazard should be developed (e.g. for situations with degraded infrastructures and accessdue to extreme external events and/or explosions, accidents with prolonged loss ofelectrical power and heat sinks, SFP accident, multi-units accidents, etc.).

•

Implementation of several connections for mobile equipment (e.g. water make-up
pumps and diesel generators) should be considered for relevant safety functions.

•

Systems and equipment for cooling the reactor core and spent fuel pools should be
independent of normal/emergency power supplies and heat sink (e.g. diesel or turbinedriven pumps with alternate water sources). Sufficient battery capacity or portable
power supplies need to support systems, equipment and instrumentation in an extended
period of station blackout.

•

Reinforced or alternative emergency management facilities and additional or alternativeequipment dedicated to providing emergency electrical power supplies and water
sources should be considered.

With respect to systems and equipment, including post-Fukushima Daiichi upgrades, thetask group estimated that efforts remain to be done to assess their response and reliability
under SA conditions including for LTM phases.

Develop harmonised practices and technical means to limit workers occupational
exposure during long-term management
Capitalisation and sharing of good practices and development of guidance to minimise workers
occupational exposure in LTMA implementation should be organised.
Development of additional technical means to enhance management of workers occupational
exposure should be fostered.
The second recommendation relates for instance to enhancement of existing radiation monitoring
systems and related analyses methods for more accurate dose mapping in harsh environment.

Feedback from past accidents has shown that it is of prime importance to provision in an
NPP sufficient and adequate workers protection and monitoring means to manage interventions
during an SA emergency phase. Improvements both in terms of protection and dose mappingand monitoring could further limit worker occupational exposure.

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Dose monitoring and mapping in harsh environments that could guide radioprotection
implementation and provide information on radioactive deposits in a damaged plant remainchallenging with currently available instrumentation. Further technological and methodologicaldevelopments in the field should be supported.
Management and operations of mobile and non-conventional equipment for LTMA should
be well defined and drills carried out to limit as much as possible workers exposure.
Chernobyl and Fukushima Daiichi damaged plants are still facing considerable
radioprotection challenges related to future defuelling and decommissioning actions. Thus,
important feedback for radioprotection issues for major LTMA is still expected in the futurefrom these two accidents.

5.3. Recommendations related to knowledge consolidation for specific long-term
management
Maintaining long-term cooling
Knowledge should be developed on:- long-term RPV resilience for situations where RPV failure has been prevented;- long-term reliability of systems and equipment involved in maintaining a coolable
configuration to assess failure risks due to induced mechanical weakness (particularly for
fixed systems that have withstood challenging SA conditions), due to clogging with debris
and due to corrosion-erosion reactions;- degradation under SA conditions (high dose and temperature) of materials that may form
significant amount of debris, posing challenges for the LT cooling.
These recommendations come in addition to recommendations related to implement provisions for
replacement of critical systems and equipment for LT cooling (as far as feasible depending on design).

One of the challenges of LTM is to ensure long-term cooling of damaged fuel/fuel debris,
possibly for years depending on the accident scenario, up to their retrieval from the damaged
plant.
If an IVMR strategy has been successful (RPV failure prevented when entering LTM),
knowledge of long-term RPV resilience considering wall thinning by contact with corium,
induced fragilities in some designs (e.g. penetrations welds attacks by corium) and coolingconfigurations (cooling from inside and outside RPV) is important for LTM as RPV failure on the
LT could result in a new unstable state. A large international research initiative is ongoing in
the Horizon 2020 Framework Programme (H2020) (IVMR project [Fichot and Carénini, 2016]) to
consolidate methods to demonstrate the safety of IVMR strategies for different reactor designs.
However, the project will hardly address the recognised issue of RPV resilience on the LT
considering in particular the effect of erosion-corrosion reactions on a RPV weakened by an SAtransient.
Though management of IVMR on the long term should pose less challenge with the gradualdecrease of residual power, thinking about optimised strategies for efficient cooling on the longterm and limiting challenges (e.g. erosion-corrosion effects) to the RPV and reactor coolant
system (RCS) could also be conducted.
If IVMR has failed (failed RPV when entering LTM), maintaining a coolable configuration may
even be more challenging as degraded fuel and debris may be distributed in and ex-vessel,
possibly in low accessibility areas. The corium dispersion in the containment might be worse
when steam explosion occurs after the RPV failure.

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In all cases, fixed or mobile emergency core/containment cooling systems recirculating highly
contaminated waters will be involved for maintaining in the long term a coolable configuration.
Whether these systems are placed inside or partly outside the containment depending on reactor
design, their failure would be critical and should be avoided since possibly resulting in a loss ofcooling and in a new unstable state or possibly releasing contaminated liquid in the environment(if a failure occurs on a system outside the containment).
For investigation of clogging issues, a prerequisite is to review existing knowledge on debris
sources in an SA (e.g. insulation material, paint chips, latent debris, debris from degraded corematerials, etc.) and their expected evolution under SA conditions (chemical effect withformation of precipitates). A significant amount of knowledge exists for loss-of-coolant accident
conditions (NEA, 2013) and approaches to assessment of clogging risks of emergency corecooling system for such conditions have been reviewed, but much less knowledge exists for SA
conditions.
For investigation of corrosion reactions which may be catalysed in the presence of radiation,
main corrosion mechanisms for SA conditions would have to be identified and characterised toassess failure risks of RPV and of circuits recirculating contaminated liquids.

Maintaining sub-criticality
Develop, both for short-term management and LTM of an SA, methods and tools to assess criticality
safety margins for damaged fuel considering uncertainties on damaged fuel characteristics (shape, size,
distribution, porosity, composition, etc.) and assess the potential consequences of re-criticality eventsin terms of damage to safety equipment and structures and radioactive material remobilisation.
This recommendation is addressed to criticality experts but requiring support of SA material andmechanical experts. Establishing harmonised criticality risk analyses would also be beneficial.

This recommendation comes in addition to recommendations related to investigate thedevelopment of instrumentation to monitor re-criticality risks.
There are presently persistent questions about the possibility of re-criticality events bothduring short-term and long-term phases of an SA. This is mostly due to uncertainties of thedamaged core configuration in the short term and on damaged fuel characteristics and theirevolution upon ageing in the long term. It is also linked to the fact that no sub-criticality
monitoring exists for damaged configurations. The risk could be more significant in the shortterm for in-vessel configurations following neutron absorber rods or blade degradation withrelocation of neutron absorber material and in the long term in solidified corium and debris due
to radio-induced ageing processes. In both situations, water with no or little amount of neutron
absorber material can come in contact with degraded fuel material.
There are also some questions about the possibility of re-criticality events in some scenarios
of SFP accidents warranting further investigation.
Further assessment of the consequences related to possible re-criticality events in LTM is
also important in view of establishing if they could result in new unstable states.

Limiting radioactive releases
Knowledge of the phenomena of remobilisation of radioactive products in the long term, both incontaminated waters and in the atmosphere should be developed. These phenomena include leachingof debris, dust formation from degraded fuel and remobilisation from aerosol deposits.
These recommendations come in addition or complete recommendations to consolidate knowledge
on containment behaviour and on long-term phenomena (cf. Section 5.1).

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When the containment has failed or its leak tightness deteriorated during the accident,
limiting further liquid or atmospheric releases of radioactivity to the environment faces in thefollowing two challenges: i) lack of knowledge of leakage paths and leakage rates and theirevolution the LT, and ii) lack of knowledge of transfer and remobilisation processes ofradioactive products on the LT in contaminated waters and in the atmosphere.
Having the capacity to assess containment leakages and their evolution in the LT is of
critical importance as earlier discussed (cf. Section 5.1).
In order to limit contaminated liquid transfer to the environment, strategies for flooding andcooling the corium should avoid conveying contaminated waters outside the confinement as far
as feasible. Systems outside the containment should be robust enough to prevent leakages.
Relative to remobilisation of radioactive products in the LT, knowledge of ageing processes
both in contaminated water and in air, particularly to assess the importance and kinetics of fuelfragmentation and dust formation from degraded fuel/debris in the long term (radio-induced
damage and oxidation) is today very limited. Also, in the LT some fission products (e.g. Sr) andactinides (e.g. Pu) may contribute more significantly to the radiological hazards by transfer tocontaminated waters. For LTM, such knowledge is important for many aspects: management ofliquid releases from the damaged plant, of liquid wastes and of fuel retrieval operations.
Atmospheric releases from the damaged plant in the long term up to fuel retrieval can be
minimised as long as cooling of degraded fuel/fuel debris can be maintained and energetic
events (e.g. hydrogen combustion, re-criticality events) can be avoided. However, knowledge isrequired to assess remobilisation risks particularly during fuel retrieval operations.
Whatever the state of the containment, if cooling is maintained in the long term,
containment pressure control should not be an issue for LTM. Provisions to manage hydrogen
combustion risk should be implemented considering its production in the long term byradiolysis and radio-catalysed corrosion reactions.

Site clean-up, decontamination and wastes management
Knowledge should be capitalised (with feedback from past accidents), shared and guidance providedon:- best approaches and techniques (cleaning, fixing or implementing protection and confinement
measures) for the treatment of various contamination types;- multi-nuclide analysis methods in highly active SA solid and liquid wastes (detection, qualitativeand quantitative analyses of major isotopes);- purification technologies and related processes for highly contaminated SA liquid wastes;- approaches and techniques for handling, routine monitoring and storage of highly active SA
wastes and for defining criteria for selection of mode of management of SA wastes.
An important aspect in implementing these recommendations is the sharing of knowledge between
the SA and waste management communities.

Two main aspects have been discussed by the task group in relation to site clean-up anddecontamination: i) technics and methods to better characterise contamination – including in
not easily accessible areas – would be most helpful to support the management of site clean-up
and decontamination operations and could also provide valuable information on contaminants
distribution in premises of the damaged plant, and ii) feedback from past accidents should befurther capitalised to provide guidance on best technics and approaches for the treatment ofvarious contamination types expected in an SA.
Research to have at hand mapping technics and methods providing data on dose – or major
radioactive isotopes distribution – in highly contaminated spaces following an SA, supporting
LTMA implementation, should be encouraged.

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Sharing of experience from past accidents on existing approaches and technics to treat SA
contamination should be organised to identify possible progress in the field and provide
guidance on most effective approaches and technics for different types of contamination andfor minimising workers exposure.
In relation to waste management, challenges are associated to limiting of volumes, treatingand handling highly active wastes and waste characterisation for its handling and storage.
Waste volumes will of course be reduced if mitigation measures have been effective inmaintaining containment and reducing the extent of radioactive releases outside the damaged
plant and if an optimised management of cooling waters for all LTM phase has been followed(cf. Section 5.4).
On-site solid waste volumes depend on the extent of contamination in the damaged plantand of operations to clean and decontaminate the plant and treat wastes. These operations
should be managed so as to limit as much as possible production of contaminated materials in
addition to existing ones in particular by avoiding as much as possible dispersion ofcontamination. Limitation of liquid waste volumes will also reduce solid wastes resulting fromtheir treatment (the filters used for purification).
Relative to treatment and characterisation of highly active wastes, methods, technics andlarge facilities were developed for past accidents (e.g. for Three Mile Island unit 2 (TMI-2),
Chernobyl and Fukushima Daiichi). It seems however that the related experience, notably on
multi-nuclide analyses in highly active waste samples and purification processes forcontaminated liquids produced during an SA has not fully been capitalised and shared. This
would be of interest to ease implementation of waste treatment facilities and improve waste
characterisation and monitoring, should another major accident occur. Also, quantitative
analysis of major nuclides in waste would provide useful information about radioactive releases
from the damaged plants.
The task group also noted that the management of tritium in water wastes may be an issue
particularly for large volumes of water and that no technology is presently available to eliminate
it when present at low concentrations. No specific recommendation was formulated but it could
be valuable to internationally reassess and harmonise criteria for the management of tritiumcontaining waters.

Fuel retrieval and disposal
Knowledge should be capitalised and developed on:- damaged fuel distribution and characteristics and their evolution over time, including ageing andleaching effects;- cutting and recovery methods and technics for damaged fuel minimising contamination
dispersion;- risks related to damaged fuel retrieval (re-criticality, fuel dust formation and dispersion).
These recommendations come in addition to those related to consolidation of SA calculation tools
and to enhancement of accident monitoring that should provide more reliable information onradioactive material distribution and properties in a damaged plant with fuel degradation.
Both recommendations are of special importance for the preparation of fuel retrieval and disposal at
Fukushima Daiichi and relevant actions have been engaged in the NEA PreADES and TCOFF projects.

As seen at TMI-2 and Fukushima Daiichi, one major issue to manage these operations is thelimited knowledge of degraded fuel/debris distribution and characteristics, until access ispossible (which could be years after the accident as at TMI-2 and Fukushima Daiichi). One issue
is the mechanical resistance of damaged and aged fuel as debris, as solidified corium, as
remnants of fuel rods or as visibly intact fuel assemblies. Retrieval may necessitate cutting formost resistant materials but some material may collapse or form dust particles with ageingthereby possibly affecting significantly material resistance.

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For an accident in a SFP, similar challenges may result (with in some cases, the additional
issue that no barrier against radiation exists). One possible additional challenge for fuelhandling and evacuation could be to identify assemblies that are mechanically weakened if theaccident extent was limited to part of the pool. Apparently intact fuel assemblies could collapse
during handling.
Consolidation of calculation SA tools relative to material aspects with the support ofdedicated experimental work (some performed with simulant materials and some with
prototypic materials, particularly as preparation for the decommissioning of the Fukushima
Daiichi plant) and knowledge capitalisation from detailed analyses made on degraded fuelsamples from past accidents (e.g. TMI-2, Chernobyl) and relevant experiments (e.g. Phebus FP,
VULCANO, etc.) should be performed as recommended in Section 5.1. These actions are engaged
in the NEA Pre-ADES and TCOFF projects.
The task group noted that more predictive tools could be supportive for fuel retrieval
operations. However, enhanced monitoring of the accident progression in and ex-vessel should
also be developed as it would directly provide information on the damaged fuel distribution.
This could involve both SA instrumentation providing information during the transient toguide severe accident management (e.g. demonstrating the success of in-vessel or ex-vessel
melt retention strategies) and also remote techniques after the accident (distant monitoring offuel material with sufficient resolution, sampling).
In relation to the limitation of contamination dispersion during the fuel retrieval operations,
it has already been suggested to design and conduct research on degraded fuel ageing in waterand in air. In addition, it is thought that research for optimised cutting technics should continue.
The objective is the selection of appropriate techniques in view of minimising fuel dispersion
and workers’ occupational exposure during fuel retrieval operations.

5.4. Recommendations related to provisions developments for specific long-term
management and actions
Provisions should be developed for optimisation of management of cooling waters, e.g.:-

During the emergency phase, closed-loop cooling should be implemented as early as possible tolimit transfer of contaminated waters outside the confinement.

-

Strategies for flooding and cooling the corium should as far as feasible avoid transfer ofcontaminated waters outside the confinement or use qualified and tight systems to prevent
contamination spread on-site.

-

Methods/criteria to optimise the mode of cooling (such as active cooling with water, natural
cooling with water, cooling in air) in the long term in order to minimise contaminated water
volumes should be developed.

-

Provisions for ensuring the injection at short and long term of water with “controlled” chemistry
should be further studied with the following potential objectives: limit re-criticality risks up to theLT, limit possibilities of fission products remobilisation (particularly iodine at short term), limit
long-term corrosion reactions, limit possibilities of precipitation and clogging in cooling loops, andfacilitate liquid waste management in the long term.

As the Fukushima Daiichi accident has shown LTM, limiting contaminated water volumes
and their transfer to the environment could reduce significantly LTM challenges. Analyses ofrequired cooling modes for foreseen plant damaged states and corresponding decay heat
distributions could provide guidance for optimised cooling in LTM.
Injections of untreated raw or sea water could potentially increase criticality risks for somedegraded configurations (in the absence of neutron absorber additives). The effect in the shortterm of some impurities on degraded fuel coolability and fission products chemistry(e.g. chlorine affecting iodine chemistry) is not adequately known. The pH control in the days

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following the accident may be of interest to limit gaseous iodine releases. Kinetics of radiolytic
and corrosion reactions as well as precipitation processes may be accelerated due to thepresence of specific impurities. This may increase the risk of failure or clogging of circuits andequipment in contact with highly contaminated liquids. The presence of some impurities insignificant concentrations (e.g. chlorine) may make the separation processes in waste
management more complex.

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Appendix A. Questionnaire on long-term management and actions
for a severe accident in a nuclear power plant

Background information
The kick-off meeting of the NEA Task Group on Long-Term Management and Actions for a
Severe Accident in a Nuclear Power Plant was held on 23-24 February 2015 at the NEA, Issy Les
Moulineaux. The work scope of the task group is:•

to review existing regulatory requirements, guidance, practices, issues underconsideration and existing technical bases in NEA member countries with respect tolong-term management and actions for a severe accident in a nuclear power plant (NPP);•

to identify, describe and review in an exhaustive way the issues to be tackled;•

to potentially propose recommendations and areas of investigations (studies and research)
to improve the management and actions for a severe accident in an NPP in the long term.

Following the kick-off meeting, the following definition was proposed for the long-term
management of a severe accident:
Long-term management 1 of severe accidents refers to accident management actions implemented
after a plant reaches a safe stable state following a degraded core accident and up to defuelling of
the reactor and removal of spent fuel off-site to permanent or intermediate storage. The timing ofa plant reaching a stable state depends on, among other things, the initiating event, plant
operating conditions and any and all prevention and mitigation measures to terminate theaccident.
A plant is considered in a safe stable state when all components of the degraded core are ina coolable configuration, either still in place and/or relocated in-vessel and/or ex-vessel, and any
stored spent fuel is also in a coolable configuration. The degraded core, if retained in-vessel, isconsidered to have reached a coolable configuration when there is no further hydrogenproduction from water-metal (clad and structural materials) interaction, the release rate offission products is exceedingly low and there is no risk of re-criticality or of corium rupturing
the vessel. Similarly, the degraded core, if ex-vessel, is considered to have reached a coolable
configuration when there is no further incondensable gas generation from molten core-concreteinteraction, release of fission products from core-concrete interactions is exceedingly small,there is no risk of re-criticality and the ex-vessel core debris is retained in the containment
without breaching the containment integrity. The spent fuel inventory is considered in a
coolable configuration if all the spent fuel rods, degraded or not, are confined in the pool without
the risk of a runaway oxidation reaction, there is no significant production of hydrogen and norisk of criticality.
Accident management actions for long-term management of severe accidents comprise theactions taken to: i) ensure the plant remains in a coolable state as defined above until complete
defuelling; ii) mitigate unintended off-site consequences; iii) ensure that there are nounintended consequences during the safe stable state of the plant; and iv) prepare for the1.

The definition of “long-term management” in the context of this project is not to be interpreted as theso-called long-term management action traditionally considered by the industry as part of the severeaccident management. The latter definition still relates to bringing a plant to a safe stable state
following a degraded core accident, and involves implementation of mitigation measures employing
mobile equipment, among other things, brought from off-site locations.

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defuelling to ensure that there are no unintended consequences during defuelling of the plant
and transfer of spent fuel to a permanent or intermediate storage location.
For more information about the project, the kick-off meeting, participating organisations
and the report outline, please visit: www.oecd-nea.org/download/wgama/ltmnpp.
To complete Action 1.7 recorded in the kick-off meeting action list, a set of survey questions
are provided below in order to solicit inputs from the task group participants. The input will be
used for the preparation of a draft report to achieve the project objectives.
Survey completed by state country, organisation(s).

184SurveyIntention of the survey

a. Are there any existing and planned
regulatory requirements and/or
guidance specifically for long-term
management and actions for a severe
accident in your country? If yes, what are
they?
b. What could be the most importantsafety objectives associated to the longterm management and actions
following a severe accident in yourcountry?

Input to be used for the Status Report:• To summarise the current regulatory
requirements and guidance in the field• To list important safety objectives for the
long-term management of a severe
accident (emphasis may be on otherobjectives than for short-term managementof the accident up to the emergency
phase), e.g. linking severe accident
management and radiation protection andwaste management aspects

Are there any commendable
practices/strategies pertinent to longterm management and actions fromyour experience? If yes, describe them.

Input to be used for the Status Report:• To summarise possible practices pertinent
to long-term management and actions(e.g. use of available on-site and off-siteresources not primarily dedicated to severe
accident management, etc.)• To provide recommendations regarding
practices for long-term management andactions

What kind of measures and actions tomaintain the plant in a long-term safe
stable state are planned, envisaged orexisting for long-term management of asevere accident?

Input to be used for the Status Report:• To summarise existing, planned orenvisaged measures and actions in the field(may be generic or plant/site specific)• Strategies• Procedures and guidelines• Equipment, infrastructure andinstrumentation• Human and organisational resources

Which factors/criteria are/were/should
be considered for thedesign/implementation of suchmeasures/actions?

Input to be used for the Status Report:

Response (or provide writeup in separate pages)

Potential factors:• Site/plant specificities• Multi-units accidents• Accidents involving spent fuel pool (SFP)• Degradation of equipment, infrastructure
and instrumentation• Positive/negative impact of short-term
prevention/mitigation actions (more or lesscontrolled during the emergency phase)• Extent of radioactive material releases• Existing on-site and off-site organisation
and resources• Etc.

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SurveyIntention of the survey

Which factors/criteria
are/were/should be considered for thedesign/implementation of suchmeasures/actions?

Potential criteria:• Fulfilment of safety objectives (cooling core
and SFP, mitigate further radioactive
releases, others targeted to recovery
phases, etc.)• Prevention/mitigation of risks (gascombustion, further radioactive releases –gas and liquid, other industrial risks)• Easing recovery, limitation of workers
exposure and further environment
contamination• Etc.

Which issues/challenges should beconsidered? What is the associatedknowledge base?
Are there any technical issues thatdeserve further investigations in viewof designing efficient long-term
management measures and actions?

Input to be used for the Status Report:• Identify, describe and review issues to be
tackled• Review existing technical bases andidentified gaps
• Provide recommendations for further
research and studies• Potential technical issues:– assessment of plant conditions after asevere accident (infrastructure,components, systems, instrumentation,
radiological conditions, etc.)– risks assessment after an SA– assessment of positive/negative impact of
emergency actions (feedback on existing
severe accident management strategy?)– corium and debris location and behaviour
after an SA– liquid effluent chemistry and treatment– recovery and mitigation techniques andsystems– etc.

What approaches (method or
combination of methods, tools, dataand computational aids, evaluation
criteria, etc.) are, or could be, used fordesigning long-term management
and actions?
Based on your experience, what kindof new methods or tools would be
convenient or necessary to better
address the issue?

Input to be used for the Status Report:• To provide a general description of existingor under development methods, tools,
data, criteria (laboratory capacities,
monitoring equipment, advanced analysis
and evaluation tools, SA codes,
prognosis/diagnosis tools, probabilistic
safety assessment, etc.) that could be used
to assess the degraded core state and theradiological situation, to plan and optimise
recovery actions and design well focusedlong-term management and actions• To discuss the merits and shortcomings(limitations) of these methods and tools forthe purpose of assuring long-term
management and actions effectiveness• To provide recommendations for future
developments devoted to increasing
confidence in methods to design long-term
management and actions (e.g. integrated
accident management approaches, etc.)

Response (or provide writeup in separate pages)

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Appendix B. Simplified application of SA-LT categorisationfrom level 2 probabilistic risk assessment releasecategory figures of merit

Level 2 probabilistic risk assessment release categories as a starting point
Conduct of long-term operations, decision making and preparedness before loss of safe stablestate strongly depends on plant configuration-specific issues. Even though large number ofvariables contributes to plant damaged state (PDS) characterisation, only a few of them turns tobe significant from a back-end approach hence making the long-term PDS identification process
subject to logic-binning-based classification.
The overarching objective of such classification targets towards improving the entire set ofmeans for managing the severe accident long term by helping highlighting main risks specificof the plant configuration. Therefore, plant state characterisation can be seen as the preliminary
task for addressing long-term safety issues.
Level 2 probabilistic risk assessment (PRA) release categories constitute groups of coredamage accident sequences featuring similar source term release characterisation in terms ofmagnitude, composition and timing as a consequence of similar containment evolution andfailure location and type. Figure B.1 depicts a generic level 2 flowchart where main tasks areplaced along the top row and connected below to user-dependent subtasks.

Figure B.1. Level 2 probabilistic risk assessment flowchart

Source: OECD/NEA.

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SIMPLIFIED APPLICATION OF SA-LT CATEGORISATION•First stage: level 2 starts from coupling sequences leading to core damage withcontainment systems performance through so-called bridge trees, whose outputs areclassified into so-called level 2 plant PDSs through the use of the PDS logic tree (PDS LT)
according to similar evolutions in containment.

•

Second stage: single PDS evolves following different sequences as a result of high level of
uncertainties related to severe accident phenomena through containment event
tree/accident progression event trees application for uncertainty propagation.

•

Third stage: release category logic tree (RCs LT) application to group containment event
tree/accident progression event trees outputs according to similar release term
characterisation: magnitude, composition and timing 1.

Transformation process from level 2 PRA RC into SA-LT PDS
Generic features dealing with release categories are briefly set forth to help carrying out
transformation into SA-LT PDS:•

release category sequences lead either into containment/bypass failure, i.e. source term
release to environment, or into safe state;•

release category logic tree classification criteria basically attend to the following issues:•–

containment isolated or bypassed;–

containment type of failure;–

related source term characterisation (mostly timing and event);–

probability issues since release category sequences are also risk oriented,
i.e. frequency values assigned to each accident sequence must be tracked throughout
the entire sequence hence treated distinctly.

Most of level 2 PRA applications limit mitigating systems availability to design basis
accident-related, hence alternative safety equipment of any kind, such portable or nonconventional, are not credited.

According to the former considerations, steps carried out to convert release categories into
SA-LT PDS are the followings:
1.

Combine release categories by removing above criteria classification on source term
characterisation related phenomena and probabilistic issues 2: From the perspective of
SA-LT, plant configuration after attaining safe stable state will not be influenced bywhether severe accident sequence yields x or 2x release values, and whether
containment failure occurs y or 2y hours after initiating event.
Nonetheless, several particular accident management actions belonging to the“accident analysis” category and “accident management” will be affected by related
source term characterisation criteria, e.g. outside radiation levels linked with accident
management action 1 on ensuring the plant remains in a safe stable state through“working conditions” issue, action 2 on mitigating releases. Such differences will betaken up in further step 4 by considering different scenarios each of which corresponds
to particular boundary conditions as specified in removed top events. Last, the majority

1.

2.

188

Fourth stage dealing with RC characterisation through system code calculations such as Accident
Source Term Evaluation Code (ASTEC), MELCOR or Modular Accident Analysis Programme (MAAP)
might be of interest when performing accident analysis and boundary conditions evolution. Fifth stage
for regulatory purposes to identify RCs whose source term exceeds certain magnitude threshold at aparticular time, i.e. large early release frequency identification, goes beyond our goals.
Different risk contributions of each sequence will turn out to be very useful if thermal-hydraulicevolutions need to be simulated, for instance to characterise containment environmental evolution or
to identify cliff-edge effects. Therefore, it should be possible to trace back each SA-LT PDS to level 2 PRA
release categories, and from here to level 1 interfacing level 2 PRA PDS within which each sequenceleading to core damage is associated with a specific frequency.
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of level 2 PRA applications includes top event questioning on reactor pressure vessel(RPV) integrity. Since this aspect of severe accident evolution is fundamental for SA-LT
PDS characterisation, this top event should be kept.
SA-LT PDS characterisation task may be finalised at this point. However, since many
crucial aspects dealing with improving severe accident management depend on fullplant configuration, i.e. state of the plant linked with mitigating system alignment,each of the resulting scenarios will hence be completed through coupling with
mitigating systems fulfilling safety functions as stated in Section 3.2.
2.

Safety functions addressed in each sequence will be identified through a check list
wherein only those minimal configurations to meet with minimal safety functions will
be pointed out and subsequently addressed.

3.

Each release category is branched into and coupled with mitigating systems alignment
to fully meet with SA-LT safety functions, where each alignment corresponds to a
specific plant configuration (including backup systems). With this aim, a parallel taskof identifying all means of carrying out each safety functions will be performed.

4.

For each obtained PDS configuration according to previous steps 1-3, distinctions
previously neglected on source term releases (magnitude, time, leakage receiverbuilding/area) will be here taken into account within each PDS category as long as
leading to a different thermal-hydraulic evolution or radioactive spreading map hence
affecting mitigating release related actions.

5.

The entire list of SA-LT PDSs will then be the result of i) RC sequences coupling withii) mitigating systems alignment within which iii) different thermal-hydraulic andradioactive spread boundaries affecting accident management issues. It is worth
noticing that particular i) mitigating releases or ii) defuelling actions may apply to morethan one PDS as classified.

Application
The method set forth above will be applied to generic Western large-dry containment pressurised
water reactor (PWR) as plant reference and level 2 PRA release category logic tree classification
criteria as depicted in Table B.1. Top row collects the binning criteria and bottom row the possible
answers to each of the criteria. Aside from design basis accident standard safety systems, backfitting system includes active ex-vessel cooling by reactor cavity flooding once reached
649 Celsius degrees at Core Exit Thermocouples. Back-fitting passive safety systems include
passive autocatalytic recombiners installed inside containment. In order to simplify theapplication, and given its high reliability degree, passive autocatalytic recombiners will not be
part of the containment event tree/release category logic tree, i.e. they will not be put into
question.

Table B.1. Release category logic tree
CONTAINMENT IN-VESSEL CONTAINMENT IN-VESSEL EX-VESSEL EARLY CONTAINMENTINTEGRITY
STEAM
ISOLATION COOLING COOLING
FAILUREEXPLOSIONSSTATE(IVMR)(DCH)

Not bypassed
SGTR
ISLOCA-V
ISLOCA-VR

Yes
No

Isolated
Large failure
Small failure

Yes
No

Yes
No

Yes
No

MCCI

No MCCI
FloodedNot flooded

SG ISOLATION WATER-COVERED CONTAINMENT CONTAINMENTSTATEBREACHCOOLING STATESPRAYS

Yes
No

Yes
No

Yes
No

Yes
No

CONTAINMENTFAILURE TYPE

ALPHA
DCH
Success
H2 Exp.
Basemat erosion
Overpressure

Source: OECD/NEA.

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Removal of source term release related classification criteria
In order to identify the reasons underlying potential screening out of release category logic treeclassification criteria, let us proceed analysing one by one:

190•

Containment integrity: bypass alternatives comprise steam generator tube rupture as
initiating or induced event, interfacing system loss-of-coolant accident as initiating event(ISLOCA-V) or at recirculation switch (ISLOCA-VR), e.g. through refuelling water storagetank. Plant configuration will largely be affected depending on containment failure
location.

•

In-vessel steam explosions: this particular phenomenon stands for alpha containment
failure mode where RPV lid thrust by in-vessel steam explosion at low pressure causes
containment failure. Contribution from this category to the spectrum of plantconfigurations only deals with source term release time hence should only besubsequently taken into account in step 4 within early containment failure thermalhydraulic sequence characterisation.

•

Containment isolation state: this classification criterion points out at source term
magnitude of the release. Alongside “isolated”, alternatives are “large”, meaning failure
size between 5 and 20 inch equivalent diameter, and “small”, meaning isolation failure
size between 3 and 5 inch equivalent diameter. In this regard, it will only be taken into
account within thermal-hydraulic sequence characterisation. From the perspective ofplant configuration, e.g. water flooding from one building to another, it may be
subsumed within containment failure condition.

•

In-vessel cooling (IVC): plant configuration will obviously be affected in case of
preventing RPV failure by flooding the core. Core reflooding without success mightinstead be considered as a different PDS or included within failure event sequenceswherein it will be separately treated for thermal-hydraulic issues. For our application,
failure to succeed in IVC will be taken as a different PDS.

•

Ex-vessel cooling (in-vessel melt retention, IVMR): IVMR will be treated as IVC.

•

Early containment failure (direct containment heating): Early failures will be further
taken into account within thermal-hydraulic characterisation.

•

Molten corium-concrete interaction (MCCI): MCCI is classified according to quenchingdegree in “no MCCI” for very rapid corium quenching by high efficient corium to water heattransfer; “flooded” for situations where water is available but high MCCI likely leading tocontainment failure whether because of basemat melt-through, overpressure or hydrogen
explosion; in case no water is present to flood the reactor cavity, “not flooded” alternative
should be selected. On one hand, plant configuration will not be affected by the extent of
MCCI as long as not jeopardising the containment barrier; on the other hand, thesecategories allow distinguishing between whether RPV is intact (“no MCCI”) or breached(other two alternatives). In order to keep this crucial criterion for PDS characterisation, thistop event will thus be transformed in question on RPV status.

•

Steam generator isolation state: this criterion addresses specifically steam generator
tube rupture scenarios and distinguishes between steam generator relief valves stuck
open or cycling. This criterion only looks at source term magnitude hence it will be
neglected in terms of PDS configuration.

•

Water-covered breach: this criterion only looks at the source term magnitude (andcomposition) of the release so it will not be taken into account for SA-LT plantconfiguration spectrum characterisation.

•

Containment cooling state and containment sprays: the rationale underlying bothcriteria rests on different containment failure frequencies since depending on gas
concentration in the containment atmosphere, probability linked to hydrogen
explosions or over-pressurisation will be likewise different. This criterion also affects tosource term magnitude of the release through containment spray scrubbing. Therefore,
it will be neglected since dealing with probabilistic and magnitude issues. Moreover,mitigating system alignment will further be taken into account in the next step.
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Containment failure type: “alpha” and “direct containment heating” early kind of failures
already treated. “Success” was initially limited to core reflooding or IVMR and it will be
accounted for. Basemat erosion will also be taken into account. Regarding “hydrogen
explosion” – which actually refers to late containment failures –, and over-pressurisation,
both categories, even if featuring different phenomenological signatures, do notcorrespond to different SA-LT plant configurations so that they will be grouped as“containment late failure” 3.
Figure B.2 depicts the filtered classification criteria acting as top events of an event tree
where each sequence represents SA-LT PDS characterisation, which constitutes the first step if
coupling with mitigating system alignment is to be performed. Depicted sequences only accountfor containment integrity “not-bypassed” option.

Figure B.2. Not-bypassed release category event tree
CONTAINMENTINTEGRITY(NBP/SGTR/V/VR)

IN-VESSELCOOLING(IVC)

EX-VESSELCOOLING(IVMR)RPVINTEGRITY(RPV)

Non-bypassed
SGTR
ISLOCA-V
ISLOCA-VR

Yes
No

Yes
No

Yes
No

CONTAINMENTFAILURE TYPE
(S/LF/BM)

No.

Code

1

NBP-S

2

NBP-RPV-S

3

NBP-RPV-LF

4

NBP-RPV-BM

5

NBP-IVMR-S

6NBP-IVMR-RPV-S

7

NBP-IVMR-RPV-LF

8

NBP-IVMR-RPV-BM

9

NBP-IVC-S

10

NBP-IVC-RPV-S

11

NBP-IVC-RPV-LF

12

NBP-IVC-RPV-BM

13

NBP-IVC-IVMR-RPV-S

14

NBP-IVC-IVMR-RPV-LF

15

NBP-IVC-IVMR-RPV-BM

ALPHA (TH)DCH (TH)
Success
H2 Exp. (late cont. fail.)
Overpressure (late cont. fail.)
Basemat melt-throughNotes: NBP – non-bypassed; RPV – reactor pressure vessel; IVMR – in-vessel melt retention; IVC – Invessel cooling; S – success; LF – late failure; BM – basemat melt-through failure.
Source: OECD/NEA.

3.

Instead, containment failure location becomes highly relevant for identifying the spectrum of SA-LT
plant configurations and subsequent actions related to accident management.

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SA-LT PDS minimal safety functions
According to Section 3.2, safety functions comprise reactivity control (A), generated heatremoval (B) either in-vessel (B1) or ex-vessel (B2), and containment protection within which
combustible gas control, basemat melt-through prevention and containment pressure control(C1, C2, C3, respectively).
Each non-bypassed SA-LT PDS sequence is coupled with requested safety functions throughcheck list as shown in Table B.2 to highlight which safety functions need to be addressed.

Table B.2. Minimal safety function configuration for non-bypassed SA-LT PDSs
SA-LT PDS
No.

Code

Safety functions
A

B1

B2

C1

C2

C3

1

NBP-S

X

X

X4

X

N/A

X

2

NBP-RPV-S

X

N/A

X

X

X

X

3

NBP-RPV-LF

X

N/A

X

X

X

N/A

4

NBP-RPV-BM

N/A 5

N/A

X

X

N/A

X

5

NBP-IVMR-S

X

X

N/A

X

N/A

N/A

6

NBP-IVMR-RPV-S

X

N/A

X

X

X

X

7

NBP-IVMR-RPV-LF

X

N/A

X

X

X

N/A

8

NBP-IVMR-RPV-BM

N/A

N/A

X

X

N/A

X

9

NBP-IVC-S

X

N/A

X

X

N/A

X

10

NBP-IVC-RPV-S

X

N/A

X

X

X

X

11

NBP-IVC-RPV-LF

X

N/A

X

X

X

N/A

12

NBP-IVC-RPV-BM

N/A

N/A

X

X

N/A

X

13

NBP-IVC-IVMR-RPV-S

X

N/A

X

X

X

X

14

NBP-IVC-IVMR-RPV-LF

X

N/A

X

X

X

N/A

15

NBP-IVC-IVMR-RPV-BM

N/A

N/A

X

X

N/A

X

Notes: NBP – non-bypassed; RPV – reactor pressure vessel; IVMR – in-vessel melt retention; IVC – In-vessel cooling; S – success LF – late
failure; BM – basemat melt-through failure.
Source: OECD/NEA (2018).

SA-LT PDS coupled with minimal safety functions configurations
Within each safety function, the entire list of combinations of hydraulic pumps, water/boric
acid/fuel reservoirs, heat exchangers, heat sinks and associated required instrumentation andcontrol including support systems will be developed. Whenever several safety functions might
be accomplished by one single mitigating system combination, minimal safety functions linked
to SA-LT PDS will hence be simplified. This might be for instance the case of functions B2 andC2, or in some cases of A and B1.
Each combination might be tagged according to the safety function representative letter
followed by a number so that each SA-LT PDS will then be linked and branched, for instance, as
1/B1-1/B2-1/C3-1; 3/B2-1; etc. In these two examples, there is only one option, i.e. one
combination of safety systems, support systems and instrumentation and control, to address
combustible gas control (C1) so that it could be removed from the sequence codification.

4.

Success in saving the RPV through IVC together with IVMR arise problems for the technical support centre
to know whether RPV integrity depends whether on RCS injection or reactor cavity flooding or on both.
Therefore, this configuration compels the technical support centre to keep operating both systems.

5.

Corium re-criticality may not be an issue in such a case as corium would have dissolved a significant
amount of concrete.

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Thermal-hydraulic sequences linked to each SA-LT PDS minimal safety functions
configuration
According to the current application as depicted in Figure B.2, thermal-hydraulic initial andboundary conditions will be characterised as the combination of the following aspects:•

most risk-significant sequence traced back upon sequences grouped under level 2 PRA
Release Categories included in each SA-LT PDS;•

Minimal safety functions alignment coupled with the selected sequence;•

combination of consistent thermal-hydraulic and radioactive different conditions as
identified in Table B.1, e.g. isolated/small failure/large failure dealing with containment
isolation system; direct containment heating (DCH) (application limited to high reactor
coolant system (RCS) pressure at RPV failure scenarios), etc.

Resultant sequences as codified are for instance 1/B1-1/B2-1/C3-1/Isolated/non-DCH and1/B1-1/B2-1/C3-1/Small_Isolation_Failure/DCH.

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Appendix C. Fault tree analysis and event tree analysis

Fault tree analysis has proven fundamental to identifying safety systems weak points. Design
extension conditions concept envisages the enhancement of plant capabilities to withstand or
mitigate severe accidents. After the events unfolded at Fukushima Daiichi, a huge number ofactions, both in back-fitting systems and guidelines, have been applied to such an extent thatbaseline capabilities in the field of beyond design basis accidents have been fully modified.
Available mitigating systems increasing complexity, including back-fitted or non-conventional,
fixed or portable, prompts parallel extension of the probabilistic risk assessment (PRA) scope tobetter address this new field of safety and help predict potential weak points hence improving
its long-term working performance with typical insights coming from PRA.
Aside from fault tree analysis implementation, event tree analysis tool application, throughan accident set of event trees and a baseline set of event trees (one per each plant damaged state(PDS) and each mitigating system failure/challenging or loss of critical safety function), could
shed light on predicting accident sequences in the long term.
Accident event tree analysis will follow traditional level 1 approach, i.e. a top-down analysisof all kind of initiating events leading to loss of so-called safety pillar. Accident event tree
initiating events should be identified through a twofold approach based on the initiating event
together with plant configuration 1, i.e. considering the challenging and/or loss of a critical safety
pillar for every type of severe accident long-term PDS.
In the baseline event trees, failures come from challenging or loss of critical safety functions
stemming only from intrinsic mitigating system failures given that working systems in the longterm will be practically limited to systems performing safety functions, i.e. mitigating systemthemselves.
Furthermore, additional benefits will derive from event tree application since many systems
will make use of same equipment and/or human actions and backup systems and infrastructure,
hence dependencies among them might be better highlighted by means of Boolean treatmentof the entire sequence rather than on a single-system basis.
Within fault tree analysis, each option available to meet one safety function will be madeup of a house event through which appropriately select the mitigating system configuration
meeting that safety function.
Accounted for systems should be all those having been included in severe accident
management guidelines (SAMGs) – whether conventional and non-conventional, fixed andportable, on-site and off-site – together with dedicated back-fitting mitigating systems, as longas their deployment to cope with the severe accident is foreseen.
Even if SA sequence event tree prognosis in the long term might be a difficult issue torealistically cope with and foreseen sequences might depart from reality, benefits will draw
more from qualitative prioritisation and identification results through relative ranking ofhighlighted potential relevant events and mitigating systems structure, system andcomponents than from quantitative results of selected figures of merit (i.e. the absolute result –the number – not as important as the relative result – how the component or minimum cut set,i.e. long-term accident sequence evolution, is ranked).

1.

Plant configuration will be sensitive not only to plant damaged state itself (accident sequence signature
affecting the long-term phase management and actions) but also to the mitigating system alignment.

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It is worth noting the following aspects concerning PRA applications:196•

Benefits drawn from PRA applications are not limited to quantifying the sources of risk
but to carrying out a comprehensive analysis of the plant performance as a whole. In thisregard, PRA is in the position of highlighting weak points thus allowing utilities
addressing them.

•

In terms of allocated efforts, level 1 PRA application mostly relies on an accuratequantification of the likelihood side of risk, whereas the consequence side, namely coredamage, is initially fixed according to a particular figure of merit, e.g. temperature
excursion, core exit thermocouples at 1 200 F, collapsed water level at TAF (top of activefuel), etc., and then simply calculated with the help of a thermal-hydraulic system code.
The underlying reason allowing not paying too much attention to the consequence sideis the simplification made when considering that all sequences coming into a hazard
situation share exactly a similar consequence, i.e. core damage, no matter whether suchdamage is more or less limited, i.e. more extended, more rapid, etc.

•

On the other hand, allocated efforts to develop a level 2 PRA – with the exception of thoseinterfacing event trees on containment safety systems – both rely on the probabilistic
and consequence side, the former being addressed by carefully computing the so-called
split fractions, the latter by carefully performing the necessary severe accident thermalhydraulic calculations.

•

As of the long-term phase of the accident, the consequence side of risk can be highly
difficult to compute since to have a clear picture of the radioactive releases resulting
from a loss of a safe stable state, or mishandling activities during decontamination or
fuel retrieval, presents so high level of uncertainties to prevent an accuratequantification.

•

This is why PRA application should apply to one and only field of activities, i.e. one andonly long-term management (LTM) goal, so that, just as in level 1 PRA, for all sources ofrisk only one consequence applies, i.e. all sources of risk end up with a similarconsequence. This is the way to simplify the PRA application by limiting the attention to
the likelihood side of risk.

•

Among the LTM top goals, PRA is well suited to cope with maintaining a coolable
configuration since the comprised actions follow a very similar nature than that of level
1 PRA: they all rely on interconnected safety systems interfacing with a source of heatand radiation.

•

Notwithstanding the above, the PRA tool might be applied to other LTM top goals
different than the one related to maintaining a coolable configuration, provided the goal
ultimately relies on systems constituted by mechanical, electrical and instrumentationand control components, subjected to human actions and environmental constraints.

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SA-LT MANAGEMENT SCHEMATIC GUIDING PROCEDURE

Appendix D. SA-LT management schematic guiding procedure

Assuming that i) a coolable situation has been reached ii) by the use of preconceived mitigating
systems, the goal is to provide managing staff with an integral comprehensive guideline linked
with specific mitigating system-oriented and safety function-oriented collection of guidelines.
This is a response for tackling with severe accident management guideline (SAMGs) extension
to cover the damaged plant long-term operation.
The SA-LT management integral process tool layout is depicted in Figure D.1, wherein around-shape box means “information source” and a rectangular-shape box means “action”.
The meaning of the most significant items is addressed below.

Figure D.1. Long-term management flowchart

Source: OECD/NEA.

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Mitigating systems in use: according to hypothesis (ii), coolability conditions have beenreached through the use of pre-existing systems, either located on-site or off-site,
conventional or non-conventional, fixed or portable. Mitigating system in use should be
updated upon surveillance criteria or system failures.

•

Selection of mitigating systems status trees: for each of the working mitigating systems
accomplishing a specific critical safety function, there is a dedicated flowchart toindicate the associated surveillance programme, alternative equipment and corrective
actions (please refer to the following section below).

•

Surveillance of updated mitigating system status trees: mitigating system status trees
will be updated (if possible) whenever available nuclear power plant (NPP) data in termsof the affected physical quantities such as inlet and outlet flowrates, liquid levels forkeeping with system performance such as water inventory and fuel refilling, room
ventilation, etc., is available. Updated values will be taken from critical safety pillars
trending values information. Alternatives for blind scenarios, i.e. without
instrumentation and control, might be considered. Surveillance criteria, component andhuman action risk ranking, etc., will be taken offline from the corresponding fault tree
analysis and (if needed) event trees.

•

Mitigating system switch: should mitigating systems in use be replaced for operating ormaintenance reasons, the process will be restarted.

•

Surveillance of long-term critical safety pillars status trees: long-term critical safety pillars aresafety issues dealing with either keeping the coolable state (inner pillars) or with mitigating
radioactive releases and effluents (outer pillars). Please refer to the section below.

Mitigating system status trees
Mitigating system status trees (MSST) should be developed for each mitigating system. The
MSST content shall be updated with available relevant in-plant data and fault tree analysis
results. MSSTs shall include the following aspects:•

Surveillance programme according to risk-informed (offline, pre-calculated) outputs
from probabilistic safety assessment fault tree analysis (in line for instance with US
Nuclear Regulatory Commission (NRC) maintenance rule) and potential recovery actions(on that particular mitigating system) derived from accident event trees.

•

Table of mandatory maintenance, preventive actions and related keep-going available
times (for instance, upon an inlet flowrate there is the need of refuelling water storagetank/CST replenishment after X hours). The fill-in form should be built up fromprobabilistic safety assessment outputs dealing with actions to maintain theperformance of the system.

•

Frequency of parameters reading should also be indicated.

•

Blind scenario performing values (conservative values).

Long-term critical safety pillars status trees
As said before, long-term critical safety pillars are those safety functions dealing with either
keeping the coolable state (inner pillars) or with mitigating radioactive releases and effluents(outer pillars).
Each pillar shall contain precursors to safety challenges and associated actions, including
relevant plant input data such as pressures, temperatures, liquid levels or flammable gases
concentration for tracking purposes. Alternatives for blind scenarios should be considered.
Significant parameter trends and potential corrective measures (embracing all kind of
pre-existing alternatives to meet that safety function aside from the current working mitigating
system) coming from accident event trees should be also taken into account (in the line of what
if responses challenging the safety pillar). Mitigating and generic surveillance actions,198

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alternative solutions, etc., specific for every plant (or accident sequence type) configuration (and,if applicable, to every particular mitigating alignment), should be part of the long-term criticalsafety pillars guideline (for instance, checking the stuck-open position of pressurised water
reactor (PWR) steam generator pilot operated relief valves in a steam generator tube rupture(induced or not) accident sequence type).
Each safety function-oriented pillar guideline will depend or will have subsections depending
on plant configuration (accident sequence type and coolability (mitigating) alignment) where
accidents would be classified according to a set of critical factors affecting success on pillar safety
goals (hence leading to different sequence evolutions): steam generator tube rupture,containment isolation failure, in-vessel melt retention (IVMR), containment mechanical failure,interfacing system loss-of-coolant accident (ISLOCA), etc. For instance, should an ISLOCA
occurred, an in connexion with radioactive releases mitigating pillar, specific instructions should
be provided to manage the auxiliary/reactor building flooding level to identify the appropriate
containment/CPV flooding flowrate in order to minimise the water being flowing out to theattached buildings to containment, simultaneously classifying the containment/CPV pressure
pillar as a non-important pillar; if the containment has not mechanically failed, provisions tocheck the state of potential leakages through the weakest seals such as personal or emergency
hatches should be addressed; etc.
First pillar, called diagnosis pillar, different in nature with respect to the safety functionoriented pillars, shall be specifically devoted to accident sequence diagnosis and SA-LT Plant
Damaged States since knowledge of these issues will affect surveillance implementation for
subsequent pillars:•

NPP general safety issues status (infrastructures, communication, operating staff) to feed
subsequent safety pillars and mitigating system fault tree analysis/event trees;•

SA-LT plant damaged state to feed subsequent safety pillars and mitigating system fault
tree analysis/event trees;•

status of defence in depth barriers;•

available expertise staff (management issues);•

available safety equipment (conventional/non-conventional, on-site/off-site, fixed/FLEX)
for each critical safety pillar;•

tracking of additional sources of risk (flooding, fire, habitability);•

SA-LT PDS-specific (including mitigating system alignment) event tree sequence results
to subsequently feed long-term critical safety pillars in terms of due surveillance actions,mitigating preparedness (alternative equipment), recovery actions, predictive risks andavailable time for associated corrective actions (outside the working mitigating systemboundaries), etc.

Each other SA-LT PDS-specific (including mitigating system alignment) long-term criticalsafety pillar shall contain:•

recovery actions derived from the diagnosis pillar dealing with mitigating releases(e.g. preparation for managing large continuous masses of radioactive liquid effluents;
closing of stuck-open valves), repairing unavailable equipment, etc.;•

precursors (quantitative settings) to anticipate safety challenges, each of which linked
with and leading to modifications of the mitigating system working values;•

baseline event tree results for available time before departing from a safe stable state as
defined in terms of critical safety functions/pillars, e.g. coolability, combustible gas
concentration, containment pressure (inside the active mitigating system boundaries), etc.;•

alternative systems able to accomplish with identical safety functions according to
mitigating system (offline, pre-calculated) fault tree analysis results (linked to currently
existing appropriate guidelines to perform preparedness actions on alternative systems);•

significant plant input data such as pressures, temperatures, liquid levels or flammablegases concentration values and trends for tracking purposes and feedback adjustmentof mitigating system performing values;

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corrective, recovery actions in case of unanticipated loss of the safety pillar both from theaccident event tree (if originated by an external event or internal different than mitigating
system failure) and baseline event tree (if originated by an active mitigating system failure);•

surveillance related actions (risk oriented).

•

specific managing staff in charge of addressing and monitoring each long-term
management of critical safety pillars.

The preliminary list includes the following pillars 1:•

criticality (inner action);•decay heat removal (inner action);•

flammable gases and oxygen (inner action);•containment over and low-atmospheric pressure (inner action);•gaseous radioactive releases (outer action);•

liquid or solid (effluents) radioactive releases (outer action);•status of defuelling (outer action).

Mitigating system fault tree analysis/event tree results
Fault tree analysis/event tree database should be fed with the following information (in linewith applications such as the monitor risk):•plant configuration (accident sequence type);•

available/unavailable equipment;•

available/unavailable infrastructures;•

available/unavailable technical (operator fields) and managing staff;•

time since the initiating event or scram (only for related severe accident system code
simulations).

Fault tree analysis database should feed mitigating system status trees with the following
information:•

surveillance programme hence risk-ranked components to be monitored (offline);•available times for addressing keep-going actions (online), entirely in case of blind
scenarios; partly in case of non-blind such as battery depletion time according to the SALT PDS (including mitigating system alignment).

Event tree database should feed long-term critical safety pillars with the following information:

1.

200

•

alternative systems available to meet the critical safety function (once the entire list of
SA-LT PDSs have been updated according to the supplied information as listed above);•available times in case of a sudden loss of the critical safety pillar (inside (departure timefrom coolability) and outside (available time for the corrective chain of actions) the
operating mitigating system boundaries), from associated deterministic simulations;offline but updated according to the elapsed time from the initiating event or scram;•corrective actions in case of a sudden loss of the critical safety pillar both in case of
external or internal event (common-cause failure for the entire mitigating systems
meeting with that safety function as identified in SA-LT PDS plus mitigating systemalignment characterisation).

Pillars shall focus on monitoring quantities rather than on accomplishable safety functions,i.e. containment/CPV flooding is not a pillar, but containment/CPV water level or flammable gases
grown-up (whether in-vessel or ex-vessel) is a pillar (the latter being a representative figure of merit of
the coolability state of the corium).
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REFERENCES

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Long-Term Management and Actions
for a Severe Accident in a Nuclear
Power Plant

As the Fukushima Daiichi nuclear power plant (NPP) accident illustrates, many challenges have to be faced
in maintaining safety over the long term in a damaged NPP following a severe accident. These comprisemaintaining and monitoring a stabilised and controlled state of the damaged plant; implementing
provisions against further failures; evaluating the plant damaged state from a physical and radiologicalstandpoint and ranking related risks; preparing and achieving fuel retrieval (either fuel assemblies stored
in spent fuel pools or fuel debris from damaged reactors); and managing safely plant recovery andaccident waste. All these actions are to be conducted protecting plant personnel from radiation exposure.
This status report reviews knowledge and experience gained through long-term management (LTM) of
the Three Mile Island, Chernobyl and Fukushima Daiichi accidents, by identifying and ranking main issues
and knowledge gaps. It also reviews the existing regulations and guidance, practices, technical bases andissues considered in member countries of the Nuclear Energy Agency regarding LTM of a severely damaged
nuclear site. Finally, it proposes recommendations and areas for future investigation to enhance LTM ofan NPP as regards necessary knowledge and provisions development, particularly for the optimisation ofmanagement of contaminated cooling waters.

Nuclear Energy Agency (NEA)
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NEA No. 7506