E-Journal of Advanced Maintenance Vol.9-3 (2017) 168-172
Japan Society of Maintenology

Toward Effective Nuclear Power Plant Ageing Management- On the Second Revision of the AESJ Code Masahide Suzuki1, Nobuyuki Nakagawa2, Shingo Ito2, Takayuki Matsufuji3,
Masao Uesaka4, Takashi Hojo5, Ichiro Tamura6, Naoto Sekimura7
1Nagaoka University of Technology, 1603-1 Kamitomioka-cho, Nagaoka, Niigata 940-2188,
Nuclear Engineering, Ltd, 3Tokyo Electric Power Services Co., Ltd, 4Tokyo Electric Power CompanyHoldings, Inc., 5The Kansai Electric Power Co., Inc., 6The Chugoku Electric Power Co., Inc., 7TheUniversity of Tokyo

2ABSTRACTIn order to ensure safe and stable Nuclear Power Plant (NPP) operation, effective ageing management (AM)program and implementation of the program is necessary. For this purpose, the “Code on Implementation andReview of Nuclear Power Plant Ageing Management Programs”, which was prepared by the Atomic Energy
Society of Japan (AESJ), is used in Japan. The Code has been endorsed and used in the regulation by Japanese
regulatory body. Since the first edition was issued in 2007, continuous effort for improvement was made, andthe second revision was issued quite recently in 2015.
The Code has the following features;- Specify ageing management actions according to plant operational years from the early stage of plant
operation.
- Utilize “Summary Sheet of Ageing Phenomena”, which has been made based on the results of aging
management technical evaluation and the actual operation experience of Nuclear Power Plants (NPPs) in Japan.
- 8 ageing phenomena which should be selected and subject to AM technical evaluation are listed, and theevaluation procedures including the seismic safety evaluation are enacted in the Attachment as a part of theCode.
In the revision, not only new information and data are added reflecting the further operating experience andthe latest knowledge including IAEA IGALL, but also the evaluation procedures and “Summary Sheet ofAgeing Phenomena” are enriched in terms of seismic effects.
The approach to effective AM program and their implementation are explained and discussed.
KEYWORDS
ARTICLE INFORMATION
Ageing Management, Ageing Phenomena, Ageing Mechanisms, Nuclear
Power Plant, Seismic Effects

1.

Article history:
Received 12 October 2016
Accepted 29 May 2017

General Description of Ageing Management (AM) Activities in Japan

Nuclear power plants (NPPs） in Japan have to meet the new requirements formulated by
Nuclear Regulation Authority（NRA）for their restarting after the Accident at TEPCO’s Fukushima
Daiichi Nuclear Power Station. Various efforts are being conducted particularly for the safetymeasures. The number of those reactors which are expected for the operation are thought to be about
40 units. The ratio of the NPPs which are over 30 years in plant operation would be about 40%.
For the ageing degradation countermeasure, all structures, systems and components (SSCs)
which have safety functions have to be evaluated before the plant operation years of 30 years. Thisaction is called ageing management technical evaluation (AMTE).
For the effective ageing management, the “Code on Implementation and Review of Nuclear
Power Plant Ageing Management Programs”, which was prepared by the Atomic Energy Society ofJapan (AESJ), is used in Japan. The Code has been endorsed and used in the regulation by Japanese
regulatory body. Since the first edition was issued in 2007, continuous efforts for improvement weremade. The first revision was made in February 2008, thereby the ageing management was much
more harmonized with the maintenance activities. The second revised edition was issued in 2015,

ISSN-1883-9894/10 © 2010 – JSM and the authors. All rights reserved.

168

M. Suzuki, et al./ Toward Effective Nuclear Power Plant Ageing Management

reflecting the knowledge learned from the Great East Japan Earthquake. Also in the meantime, theknowledge and information obtained in the AMTE were reflected as the Supplement of the Code,which were issued almost every year.
Here the concept, contents and features of the Code of the second revision were explained andthe way towards the continuous improvement is discussed.

2. AESJ Code on Implementation and Review of Nuclear Power Plant Aging
Management Programs
2.1. Characteristic Feature of the Code
The Code has various characteristic features, such as the followings since the 2008 edition[1]-[2];
1) Effective ageing management can be continually performed from the early stage of operation.
Therefore, ageing management actions are specified in the Code according to the plant operational
years, namely, the period from the early stage of plant operation, the intermediate time at every 10years, and the time when AMTE is needed before the operation of 30 years and following every 10years.
2) In the early stage of the plant operation,“Summary Sheet of Ageing Phenomena”is utilized forthe maintenance management activities. The example of the Sheet is shown in Fig.1 for a pump, bywhich possible ageing phenomena could be extracted. The database for the Sheet has been establishedbased on the actual operation experience, in particular, by reviewing the final report of AMTE ofNuclear Power Plants (NPPs) in Japan. Through the Summary Sheet of Ageing Phenomena, more
harmonized action for maintenance and ageing management activities has become possible from plantcommissioning.
3) For the extraction of ageing phenomena which should be subject to AMTE, identification andclassification of ageing phenomena were necessary, from the point of the following consideration,
namely,- On degradation modes of each ageing phenomenon,- On necessary inspections and/or evaluations to monitor the degradation modes,- On feasibility of trend monitoring.

P01-01
No.
12

Pump (vertical axial flow turbo pump/seawater/SS)

Issues required toachieve intended
functions
Assurance of pumpcapacity (head)・1314Maintenanceboundaryof・
23

Support of component

PartSS

Wear

Main shaft

SS

Corrosion (pitting, etc.)

・・・

・・・

・・・

Discharge elbow

Cast ironCorrosion (pitting, etc.)

Discharge pipe

Cast ironCorrosion (pitting, etc.)

・・・・・・・・・

Support plate

SS

（N/A）

SS

（N/A）

・・・

・・・

・・・

Fig.1

Ageing phenomena

Main shaft

24
・

MaterialRemarks

・・・

・・・

・・・

Schematic diagram of Summary Sheet of Ageing Phenomena

As a result of the consideration, ageing degradations were classified to three modes. The firstone ① corresponds to the degradation mode for which planned inspections should be implemented
from the early stage of operation. The second one ② is the mode for which continuous inspections
should be implemented from the early stage of operation. The last one ③ is the mode for which
trend monitoring should be performed at the predetermined timing. For example, IGSCC(Intergranular stress corrosion cracking) and PWSCC (Primary water stress corrosion cracking) were169

M. Suzuki, et al./ Toward Effective Nuclear Power Plant Ageing Managementclassified to ①, while pipe wall thinning due to FAC (flow accelerated corrosion) was classified to②. The ageing phenomena classified to the mode ③ were considered to be subject the AMTE.
Finally from these consideration, eight ageing degradation phenomena necessary for theAMTE are identified. These are as follows;- Neutron irradiation embrittlement,- Irradiation assisted stress corrosion cracking (IASCC) including under irradiation creep andirradiation swelling,- Low cycle fatigue,- High cycle thermal fatigue (fluctuations in temperature),- Insulation degradation of electrical /instrumentation equipment,- Thermal ageing of cast stainless steel,- Degradation of strength and shielding function of concrete structures,- Fretting fatigue.
After AMTE implementation, the seismic safety are evaluated considering the degradation
phenomena which could affect the safety of the plant. These phenomena are not only theabove-mentioned 8 ageing phenomena but also other ageing degradation like flow accelerated
corrosion.
For these ageing degradation phenomena, the evaluation procedures are enacted in theAttachment as a part of the Code.
2.2. The Main Revised Points of the 2015 Edition
In principle, the revision has been conducted reflecting not only the latest knowledge which wasestablished after the last edition and expanded operation experience, but also taking into consideration
the various changes including the new regulatory requirements and lessons learned from the・Example ; Pump, Heat Exchanger
necessary

No. function
Location1ensure Main shaft
2
capacity
3 and pump Impeller
head
4
Impeller Ring

MaterialSUS
cast SUS，Cu alloy (cast)－5
6
7
8ensureheattransfer

Heat exchanger tube

Inconell 690 alloy

Ageing degradation
phenomenawear
fatigue crack（highcycle）
corrosion（cabitation）（consumables・periodicalyexchange）
fatigue crack（flettingfatigue)
crevice SCC
dentingscale adhesion

Seismic Safety
effects
static dynamic▼
☆

▼★
▼▼

9

anti-vibration metal fitting

SUSwear★

▼10

coolant outlet/inlet nozzle stubsafe end （SUS）

SUS（Inconell overlay）

SCC★

◎manhole (primary side)

low alloy steel（insert plate isSUS）（not atnticipated）

★

12

13
14 maintain tube plate
15 boundary

low alloy steel（Inconell overlay）

16

gasket－partition plate

Inconell 600 alloy
Inconell 690 alloy

17
18

fatigue crack
overlay SCC
lower part overlay crack（consumables・periodicalyexchange）
SCCSCC★

◎▼

★
★

▼

From seismic safety point（ based on JEAC4601 ）・static function― ★：evaluation target、／：non evaluation target
・dynamic function― ★： evaluation target、☆： evaluation target when not rigid、＊：cope with generalseismic design、／：non evaluation target
・effects― ◎：significant、■：little、▼：depending on the ageing management, ◎ or ■ is judged、／： non evaluation target

Fig.2

Schematic diagram of Summary Sheet of Ageing Phenomena for Seismic170

M. Suzuki, et al./ Toward Effective Nuclear Power Plant Ageing Management

Fukushima Daiichi Nuclear Accident. Main new and/or modified points are as follows [1];1) One of the most important revised points is seen in the sophistication of seismic safety
evaluation relating to ageing. Data and information for seismic evaluation are added in the form ofSummary Sheet of Ageing Phenomena for Seismic. The example is shown in Fig.2. Degradation
phenomena which could affect the seismic safety are picked out, from not only the above-mentioned 8ageing phenomena but also other degradation phenomena.
The evaluation procedures are also enacted in the Attachment as a part of the Code. As anexample, seismic evaluation method for the local wall thinning of the piping prescribed in the Code isexplained next.
2) The effect of local wall thinning on seismic response can be evaluated using either theminimum required thickness piping model or the predicted wall thinning piping model. Because ingeneral seismic response analysis on nuclear power plant requires a lot of time and efforts, thedevelopment of simplified and reliable method was essential for the effective ageing management. In
the present Code of the 2015 edition, for the seismic response analysis of the thinning piping of theclass B and C in terms of the seismic classification, the simplified calculation method which uses theresults of sound piping is prescribed. This method is thought to be one of the characteristic point of
the present Code.
The method first assumes the degraded configuration by use of, for instance, the minimum
required thickness piping model shown in Fig.3. Then, the primary stress of the wall thinning piping
is calculated based on either the wall thickness or floor response spectrum by comparing the result ofdegraded piping with that of sound piping. For this purpose, we use such ratios like section modulus,
stress factor and response acceleration ratios. For adopting to the Code, careful attention had to be
paid so as to the make the evaluation conservative. Hence, the selection of the evaluation points or thecorrection of floor response spectrum configuration, etc. are important, and these are also prescribedin the Code.

Wall thinning modelling range
Drift region
Downstream of drift region
Measurement region by JSME Codeanchor

restraint

Fig.3

Example of the minimum required wall thickness piping model3) For the plant which is stopped for a long time, ageing evaluation has also become possible taking
account the reason that caused the prolonged stop as well as the plant condition in the meantime.
4) Evaluation on the tsunami-resistant safety from the point of ageing for cold shutdown plants is171

M. Suzuki, et al./ Toward Effective Nuclear Power Plant Ageing Management

incorporated to the 2015 edition. This corresponds to the new regulatory requirements after theFukushima Daiichi Nuclear Accident.
5) The knowledge from the IAEA IGALL (International Generic Ageing Lesson Learned) [3] wasalso reviewed whether there existed any domestic plant issue. As a result, ageing phenomena relating
to wear, boric acid corrosion, tensioning force relaxation of the tendon, etc. are reflected in therevision.
6) Other points revised are like the followings:- Target evaluation period for AMTE is changed to have more flexibility, not necessarily fixed to60 years considering the various plant specific situation.
- Reflecting the latest version of various civil codes and standards, including evaluation guides
which were prepared based on the results of national projects,- Evaluation procedure on the assessment of ageing management measures which was made atprevious AMTE is prescribed in the Code. The verification is considered important to confirm
whether there is a gap between the prediction and actual ageing progression.

3.

FutureDirection andConclusion

The continuous improvement reflecting the latest knowledge of both domestic and overseas,taking every surrounding circumstance relating to plant safety into consideration, steady accumulationof operating experiences and ensuring the path for the feedback of new knowledge to standards andcodes, etc. should be the key to effective and useful AMP and hence that is our philosophy to improve
the Code. Industry-academic-government cooperation as well as international cooperation and/orinternational contribution will continue to be essentially important [4].
From the technical point of view, the lessons learned from the Fukushima Daiichi Nuclear
Accident are reflected in the revised Code as stated in the Section 2.2. If further facts are made clear
as the investigation proceeds, proper revision will also be made accordingly.
Technical development would also take place rapidly in the future in such area as inspection
technology. For example, many innovative technologies are proposed in the area of wall thinning
detection, because wall thinning phenomenon requires the technologies about nondestructive sensing
and anti-corrosion measures. It should be the role of code engineer to incorporate the new method
which is important to safety to code and standard timely.
On the other hand, further efforts have to be made especially in the area of stakeholder
communication. For example, extending the plant operation over 40 years should require theunderstanding of local residents. The role of the academic society should become even more
important, and hence it is also our policy to make a clear explanation of the Code at every opportunity.
The communication would also contribute to the continuous improvement.

Acknowledgement
Special thanks are due to the member and staff of PLM Subcommittee, Atomic Energy Society ofJapan.

References[1] Atomic Energy Society of Japan: AESJ Standards “Code on Implementation and Review of Nuclear PowerPlant Ageing Management Program:2015”.
[2] Atomic Energy Society of Japan: AESJ Standards “Code on Implementation and Review of Nuclear Power
Plant Ageing Management Program:2008”.
[3] IAEA International Generic Ageing Lessons Learned (IGALL) for Nuclear Power Plants (2013)[4] Masahide Suzuki and Toshiaki Nishiyama: Maintenology, Vol.9. Number 4, pp.2-8(2011) (in Japanese)172