Nuclear Engineering and Design 195 (2000) 211 – 215
www.elsevier.com/locate/nucengdes
Risk-informed inservice inspection program
Michael E. Mayfield a,b,*, Deborah A. Jackson a,b, Jack Guttman a,b,
Mark Cunningham a,b
a
US Nuclear Regulatory Commission, Electrical Materials and Mechanical Engineering Branch, Washington, DC 20555 0001, USA
b
Smirt 14, PCS 2, Lyon, France
Received 28 November 1997
Abstract
The Nuclear Regulatory Commission (NRC) has developed draft guidance for power reactor licenses on acceptable
methods for using probabilistic risk assessment (PRA) information and insights in support of plant-specific
applications to change the current licensing basis (CLB) for inservice inspection (ISI) of piping. This process is also
known as risk-informed inservice inspection programs (RI-ISI). The risk-informed inservice inspection process for
operating nuclear power plants provides an alternative method for selecting and categorizing piping components that
are inspected for the purposes of meeting the requirements of ASME Section XI. A RI-ISI approach will incorporate
probabilistic techniques to help define the scope, type and frequency of inservice inspection. The risk-informed
process may support a decrease in the number of inspection and inspection intervals but will also identify areas where
increased resources should be allocated to enhance safety. The approach discussed in this paper follows the method
developed by NRC staff. © 2000 Published by Elsevier Science S.A. All rights reserved.
1. Introduction
Current inservice inspection (ISI) programs are
based on past experience and engineering judgement. Service experience of over 2000 reactor
operating years indicates that failures are dominated by corrosion or fatigue mechanisms and
typically occur in areas not included in licensees’
ISI programs. This results in significant resources
being spent on inspection programs that provide
minimum benefit. Use of RI-ISI could reduce
worker exposure by eliminating inspection locations that are not subject to active degradation.
RI-ISI applies new technology and experience to
better focus inspection locations.
2. General background
* Corresponding author. Tel.: +1-301-4156690; fax: +1301-4155074.
On August 1995, the NRC published a policy
statement on the use of probabilistic risk assess-
0029-5493/00/$ - see front matter © 2000 Published by Elsevier Science S.A. All rights reserved.
PII: S 0 0 2 9 - 5 4 9 3 ( 9 9 ) 0 0 2 5 0 - 2
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M.E. Mayfield et al. / Nuclear Engineering and Design 195 (2000) 211–215
ment (PRA) methods in nuclear regulatory activities. It was the Commission’s intent that implementation of this policy statement would improve
the regulatory process in three areas: (1) enhancement of safety decision making by the use of PRA
insights; (2) more efficient use of agency resources; and (3) reduction in unnecessary burdens
on licensees. Subsequently, the NRC staff developed draft regulatory guidance and standard review plans to implement the Commission’s policy
statement. In June 1997, the Commission approved the publication of four draft regulatory
guides (RG) and draft standard review plans
(SRP) for public comments. The regulatory activities addressed by the documents include:
1. General guidance that sets the framework and
acceptance guidelines for all risk-informed licensing activities (DG RG-1061 & SRP Chapter 16)
2. Guidance on incorporating the framework and
acceptance guidelines into the following specific activities: inservice testing (DG-1062 &
SRP Section), graded quality assurance programs (DG-1064), and technical specifications
(DG-1065 & SRP Section).
The draft regulatory guidance document for
RI-ISI programs (US Nuclear Regulatory Commission, 1997) which focuses on the regulatory
programs to incorporate risk insights into inservice inspection programs, will be published in the
fall of 1997 These documents serve three purposes, (1) to help implement the Commission’s
1995 policy on the use of risk-information in the
regulatory process; (2) to provide an acceptable
approach for power reactor licensees to prepare
and submit applications for proposed plant-specific changes to the CLB that utilize risk information and (3) to provide guidance to the NRC staff
on the review of licensing submittals.
In addition to the NRC staff efforts, the US
industry has been actively pursuing risk-informed
initiatives. The nuclear industry, under the umbrella of Nuclear Energy Institute (NEI), has
submitted two topical reports which describe different methodologies for the implementation of
the RI-ISI. One methodology has been jointly
developed by ASME Research and the Westinghouse Owners Group (WOG). The ASME/WOG
report addresses the details of a quantitative RIISI program. The other methodology is being
developed by Electric Power Research Institute,
(EPRI), and describes a qualitative approach to
RI-ISI programs. The proposed change includes
an allocation of piping resources commensurate
with the importance of the piping and the potential for degradation.
The industry has proposed pilot plant studies to
evaluate each of these methods. The pilot plant
for the ASME/WOG methodology is Surry I,
while the pilot plants for the EPRI methodology
are ANO-2 and Fitzpatrick.
In addition to the industry’s efforts to develop
alternative methodologies, the ASME Code has
been working to provide risk-informed alternatives through the development of Code Cases.
Three Code Cases for alternate examination requirements to ASME Section XL, Division I for
piping welds are being developed. Code Case N560, which provides alternate examination requirements for Class I, Category B-J piping welds,
based on the EPRI methodology, was approved
in November 1996. That Code Case was revised
to include the WOG methodology, as an option.
Code Case N-577, which provides alternate examination requirements for Risk-Based Selection
Rules for Class 1, 2, and 3 piping welds, is based
on the WOG methodology. Finally, Code Case
N-578, which provides alternate examination requirements for Risk-Based Selection Rules for
Class 1, 2, and 3 piping welds, is based on the
EPRI methodology. The NRC staff is working
with the industry and ASME to ensure integration and consistency of the various approaches
being proposed.
2.1. Key principles of risk-informed regulations
For the licensee who elects to incorporate risk
insights into its ISI programs, it is anticipated that
the licensee will build upon its existing probabilistic risk analysis (PRA) activities. Draft Reg Guide
1061 ‘An approach for plant-specific risk-informed decision making: general guidance’ describes a general approach to risk-informed
regulatory decision making and identifies five key
principles that encompass a risk-informed philos-
M.E. Mayfield et al. / Nuclear Engineering and Design 195 (2000) 211–215
ophy. These are: (1) the proposed change meets
the current regulations; (2) defence-in-depth is
maintained; (3) sufficient safety margins are maintained; (4) proposed increases in risk and their
cumulative effect are small and do not exceed the
NRC Safety Goals; (5) performance based implementation and monitoring strategies are proposed
that address uncertainties in analysis models and
data and provide for timely feedback and corrective action.
3. Description of the RI-ISI process
The RI-ISI process has four major elements, (1)
define the proposed changes to the inservice inspection program; (2) perform an engineering
analysis that is composed of traditional engineering analyses and incorporates PRA insights in the
final decision making; (3) define an implementation and monitoring program that addresses uncertainties in analysis models and data, and
provides for timely feedback and corrective actions; and (4) document evaluations and submit
request for proposed changes. Addressing the
four-elements in a licensing submittal is expected
to improve the license review process of a licensee’s submittal by ensuring that issues of regulatory importance are addressed (Fig. 1).
Fig. 1. Four principal elements of risk-informed, plant-specific
decision making.
In element I the proposed changed to the ISI
program would be defined with a full description
of the proposed change that would include:
1. An identification of the aspects of the plant’s
CLB that would be affected by the proposed
RI-ISI program.
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2. An identification of the specific revision to
existing inspection schedules, locations and
methods that would result from implementation of the proposed program.
3. Any piping not presently covered in the plant’s
ISI program, but which are determined to be
categorized as high-safety-significant (e.g.
through PRA insights) should be identified
and appropriately addressed. In addition, the
particular systems that are affected by the
proposed changes should be identified since
that information is an aid in planning the
supporting engineering analysis.
4. An identification of the information that will
be used to support the changes. This will
include performance data, traditional engineering analyses and PRA information.
5. A brief statement describing the way the proposed changes meet the objectives of the Commission’s PRA policy statement.
The proposed changes include an allocation of
piping resources commensurate with the importance of the piping and the potential for
degradation.
Engineering analysis, element 2, comprises two
strategies The first classification is the traditional
engineering analysis. For this analysis, basic material sciences, engineering fracture mechanics analysis, failure mode and effects analyses, etc., are
performed. The second classification of engineering analysis is the probabilistic risk assessment,
from which risk-insights are obtained. In this
analysis, most, if not all of the traditional analysis
are input to the PRA model of the nuclear power
plant. This integration is the state-of-the-art in
RI-ISI. The PRA and supporting analyses are
used for two specific purposes. The first is to
categorize piping segments into high and low
safety significance. The second purpose is to identify the change in safety that results from the new
inspection program. The integration between the
traditional and PRA analyses is an iterative process. Once the new inspection program is defined,
the change in risk to the general public is evaluated. If the change meets the acceptance guidelines identified in draft Reg Guide 1061, then the
proposed inspection strategy could be found acceptable, conditional on an acceptable implementation and monitoring strategy (Fig. 2).
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M.E. Mayfield et al. / Nuclear Engineering and Design 195 (2000) 211–215
the ISI plan without modification and (2) appropriate modifications are developed if new or unexpected degradation mechanisms occur.
Element 4, documentation should describe the
proposed RI-ISI program in sufficient detail for
the reviewer. The description should discuss the
five items in element 1 so that the reviewer will
comprehend how the program would be
implemented.
4. Results
Fig. 2. Engineering evaluation process overview.
The number of pipe segments defined for an ISI
analysis will be plant specific. Given that system
boundaries involve system functions and may also
involve interactions between different systems, the
definition of these boundaries requires a careful,
logical approach. All interfaces must be identified
to ensure that there is consistency between the
defined boundaries, when viewed from the systems on either side of each boundary, and that no
safety functions are overlooked. In making its
safety findings, the NRC will assess the proposed
RI-ISI program stipulated in 10 CFR
50.55a(a)(3).
In element 3, plans are formulated to monitor
factors that reflect piping reliability commensurate with the pipe’s safety significance. The RI-ISI
program plan is developed using the information
obtained from elements 1 and 2. The plan should
include implementation, performance monitoring
and corrective action strategies. The requirements
for program implementation for a RI-ISI are the
same as the implementation requirements for ISI.
The program may begin at any point of the
inspection interval as long as the examinations are
in accordance with the licensee commitments
made in accordance with 10 CFR 50.55a. Performance monitoring of the RI-ISI is required. Monitoring ensures that, (1) piping reliability used in
the calculation of risk contributions from piping
remain valid and thereby justify continuation of
To date, the staff has not received an application for a RI-ISI program. As previously stated in
the background section of this paper, the industry
has submitted two topical reports on RI-ISI methods. NRC staff has also been working with Virginia Power on their pilot plant application of
RI-ISI development. The knowledge gained to
date with the pilot plant application has been used
to help define the content and guidance contained
in the draft RG and SRP.
As part of the NDE program sponsored by the
NRC, Pacific Northwest National Laboratory developed and applied a method using risk-informed
techniques for ISI plans of nuclear power plants.
NUREG/CR-6181, 1997, provides a methodology
which incorporates recent plant-specific information and improved risk-informed calculational
tools. The methodology used was a test of the
expert elicitation process, which is permitted by
the draft RG on RI-ISI. The results from the
expert elicitation process in NUREG/CR-6181
will be compared with the ASME/WOG pilot
plant application.
5. Conclusions
The development of the RI-ISI process has
required extensive interaction between the NRC
staff and the industry. NRC staff has been performing independent evaluations in parallel with
industrys’ initiatives, including evaluations of the
ASME/WOG topical report and the preliminary
information available from EPRI. Final results
from the pilot plant applications will provide ver-
M.E. Mayfield et al. / Nuclear Engineering and Design 195 (2000) 211–215
ification that the RI-ISI process does provide a
sound technical alternative for ISI and may also
provide economic benefits for operating licensees
by possibly reducing the number of inspection
locations and occupational exposure during the
inspectors. The reduction in inspections will be
accomplished without compromising existing
safety standards.
.
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References
US Nuclear Regulatory Commission, 1997. An approach for
plant-specific, risk-informed decision making: inservice inspection of piping. Draft Regulatory Guide 1063.
NUREG/CR-6181 Rev. I, 1997. A pilot application of risk-informed methods to establish inservice inspection priorities
for nuclear components at surry unit I nuclear power
station.