POLICY ISSUE 

(Information) 

August 14, 2009 SECY-09-0113

FOR: The Commissioners 

FROM: Michael R. Johnson, Director 

Office of New Reactors 

SUBJECT: UPDATE ON THE DEVELOPMENT OF CONSTRUCTION 

ASSESSMENT PROCESS POLICY OPTIONS AND THE 

CONSTRUCTION INSPECTION PROGRAM INFORMATION 

MANAGEMENT SYSTEM 

PURPOSE: 

In response to Staff Requirements Memorandum (SRM) M081022, “Staff Requirements - 

Periodic Briefing on New Reactor Issues,” dated December 5, 2008, this paper updates the 

Commission on the progress by the U.S. Nuclear Regulatory Commission (NRC) staff towards 

the development of the construction assessment program policy options. This paper also 

provides an update regarding the development and embodiment of safety culture for the 

construction inspection program (CIP), including inputs and thresholds for the construction 

response table (CRT). In addition, this paper updates the Commission on the staff’s progress in 

developing the software enhancements to the Construction Inspection Program Information 

Management System (CIPIMS). 

BACKGROUND: 

In the SRM associated with SECY-07-0047, “Staff Approach to Verifying the Closure of 

Inspections, Tests, Analyses, and Acceptance Criteria Through a Sample-based Inspection 

Program,” dated May 16, 2007, the Commission directed the staff to engage industry and 

interest group stakeholders to obtain their views on the CIP using a public meeting approach 

similar to that used during the development of the Reactor Oversight Process (ROP). The 

Division of Construction Inspection and Operational Programs in the Office of New Reactors 

CONTACT: Robert Pascarelli, NRO/DCIP 

 (301) 415-1245 
The Commissioners - 2 - 

(NRO) and the Region II Center for Construction Inspection have conducted numerous public 

meetings with the industry and other NRC stakeholders for the past 2 years to discuss the 

development of programs and procedures related to new reactor construction under Title 10 of 

the Code of Federal Regulations (10 CFR) Part 52, “Licenses, Certifications, and Approvals for 

Nuclear Power Plants.” One of the primary purposes of these meetings has been to obtain 

stakeholder input and feedback for the construction assessment program. 

The staff considered stakeholder input received during these public meetings to develop the 

construction assessment program, which is described in Inspection Manual Chapter (IMC) 

2505, “Periodic Assessment of Construction Inspection Program Results,” dated 

October 20, 2008. The staff submitted SECY-08-0155, “Update on the Development of the 

Construction Inspection Program for New Reactor Construction under 10 CFR Part 52,” dated 

October 17, 2008, and subsequently briefed the Commission on October 22, 2008. Following 

the briefing, the Commission directed the staff in SRM M081022 to reconsider the construction 

assessment process as presented in IMC 2505 and propose policy options to the Commission. 

The Commission further directed that the staff proposal should address the inclusion in the 

construction oversight process of objective elements such as construction program performance 

indicators (PIs) and significance determination processes (SDPs) analogous to those used in 

the ROP. 

DISCUSSION: 

The staff has reconsidered the construction assessment process by engaging stakeholders in a 

series of meetings intended to identify alternative means of assessing licensee performance to 

those described in SECY-08-0155 and IMC 2505, and to develop construction assessment 

program options for Commission consideration. The staff re-reviewed ROP basis documents in 

an effort to identify additional features that could be applied in the development of the 

construction assessment process. 

The staff presented assessment program concepts, including options to develop and implement 

construction PIs and SDPs, to external stakeholders during numerous public workshops and 

meetings. The staff considered the feedback received from these meetings with stakeholders in 

further development of the options. The staff also considered feedback contained in a 

December 5, 2008, Nuclear Energy Institute (NEI) letter to the Commission entitled “NRC 

Oversight of Construction Activities” (Agencywide Documents Access and Management System 

(ADAMS) Accession No. ML092080426). Specifically, NEI did not support the development and 

use of PIs and SDPs for construction. 

As the staff was concluding its work to develop construction assessment program options 

through public workshops, the industry developed an independent proposal for the construction 

assessment process. On July 2, 2009, NEI submitted a letter to the NRC, entitled “Proposed 

Construction Inspection Assessment Process” (ADAMS Accession No. ML091831352). This 

effort, led and coordinated by NEI, presents a concept that is similar in structure to the ROP 

regulatory framework. This concept does not include the use of PIs and proposes that the 

significance of findings be determined using a method other than traditional enforcement. The 

concept lacks detail and requires developmental work. 
The Commissioners - 3 - 

The staff has considered NEI’s framework and supporting text. In light of the timing and content 

of NEI’s proposal, the staff believes that further collaboration with all stakeholders through 

public workshops is necessary to determine the most comprehensive set of construction 

assessment program policy options. NRC consideration and further development of NEI’s 

recent proposal will necessitate the allocation of additional staff resources to provide the 

Commission with policy options. 

During these meetings, the staff plans to thoroughly explore stakeholder proposals and to 

develop policy options for Commission consideration. With this in mind, the staff plans to 

provide construction assessment program policy options to the Commission in November 2010. 

This schedule will support the development of policy options prior to the first unit combined 

operating license and the full onset of construction oversight activities. To support the first 

limited work authorization, staff will revise IMC 2505 and IMC 0613 “Documenting 10 CFR 52 

Construction and Test Programs” to incorporate a safety culture approach, such that a 

construction assessment program can be implemented on an interim basis. 

As the staff considers the NEI proposal, it will build on the extensive work already completed in 

developing policy options. The enclosures to this document describe the results of work 

conducted to date by the staff in developing alternatives for the construction assessment 

program. Enclosure 1 describes the features that the staff believes the construction 

assessment program policy options should include, such as periodic assessments of licensee 

construction performance and areas important to safety culture, trending and reacting to issues 

to prevent larger issues from developing, a CRT that provides for a range of NRC actions 

commensurate with the significance of issues identified during the CIP, and routine and clear 

communication of assessment results to stakeholders. Consistent with the ROP, traditional 

enforcement should be used to disposition issues that have an impact on the regulatory 

process, are willful, or have actual safety consequences. 

Enclosure 2 details the methods the staff has considered, to date, to evaluate the significance of 

findings. Enclosure 3 describes staff efforts to develop construction PIs. The NRC could 

implement traditional enforcement processes to assess CIP findings. While traditional 

enforcement could be readily used, the development of a SDP is potentially challenging and will 

require additional work. Alternatively, the staff could develop a construction SDP to assess the 

significance of CIP findings. The establishment of a risk-informed threshold for construction PIs 

has proven to be particularly challenging because of the lack of performance data to benchmark 

new construction PI thresholds. 

The staff and our stakeholders recognize that a strong safety culture during new reactor 

construction is important to ensuring that the newly constructed plant is in compliance with the 

design and will be capable of operating safely following construction. In order to remain aligned 

with agency safety culture developments, NRO representatives have participated in NRC staff 

efforts pursuant to the Commission direction provided in COMSECY COMGBJ-08-0001, “A 

Commission Policy Statement on Safety Culture,” dated February 25, 2008. NRO will continue 

to participate in these NRC staff activities and will develop a long-term approach to safety 

culture that is consistent with that developed for the ROP. 
The Commissioners - 4 - 

The staff recognizes the need to assess areas important to safety culture at the beginning of 

NRC inspections associated with licensed construction activities. The staff reviewed domestic 

and international construction events and evaluated the applicability of the existing 13 safety 

culture components to these events. As a result of this review, it was determined that all 13 of 

these components were integral to a comprehensive NRC oversight of safety culture as applied 

to new reactor construction activities. Therefore, the staff intends to implement a near-term 

approach to safety culture which closely resembles that already being implemented in the 

operating reactor assessment program. This oversight role includes documenting cross-cutting 

aspects during the course of NRC inspections, evaluating these findings against a pre-defined 

set of criteria to determine if a significant concern exists, and conducting appropriate follow-up 

actions applied using a graded approach. Significant concerns will be treated in a manner 

analogous to the ROP’s substantive cross-cutting issues. 

On July 23, 2009, the staff conducted a public meeting with stakeholders and informed them of 

the planned near-term approach to incorporate safety culture into construction oversight, in a 

manner that closely resembles the ROP. The staff also discussed long-term options under 

consideration for incorporating safety culture into the construction assessment process. The 

staff has three public meetings planned for the remainder of calendar year 2009 and has 

tentatively scheduled seven public meetings for 2010. 

The staff has also continued development of CIPIMS, which will be used for documenting 

results from 10 CFR Part 52 construction inspections. CIPIMS is currently available as part of 

the Reactor Programs System, but work is in progress to migrate CIPIMS to the Enterprise 

Project Management (EPM) platform. The move to the EPM environment will provide 

inspectors with tools such as spell check and rich text editing. This transition is expected to be 

completed by the end of fiscal year (FY) 2009. Continued enhancements will be made to 

CIPIMS as experience with its use and capabilities is gained. 

COMMITMENTS: 

The staff plans to issue revisions to IMC 2505 and IMC 0613 by November 30, 2009, including 

an approach to safety culture similar to the existing ROP, so that a complete assessment 

process can be implemented on an interim basis to support the issuance of the first limited work 

authorization. 

The staff intends to continue working with industry and other stakeholders on the development 

of assessment program policy options, including long-term options that are being considered for 

incorporating safety culture into the construction assessment process. The staff plans to submit 

construction assessment program policy options to the Commission by November 2010. 

RESOURCES: 

The staff estimates that 3 FTE in FY 2010 and 1 FTE in FY 2011 will be required to complete 

the construction assessment program policy options. The staff intends to redirect FTE from 

other lower priority program development work. NRO will work with the Region II Center for 

Construction Inspection to ensure the redirected FTE will not impact timely implementation of 

the construction inspection program. 
The Commissioners - 5 - 

COORDINATION: 

The Office of the General Counsel has reviewed this paper and has no legal objection. The 

Office of the Chief Financial Officer has reviewed this paper for resource implications and has 

no objections. 

 /RA/ 

Michael R. Johnson, Director 

Office of New Reactors 

Enclosures: 

1. Construction Assessment Program Attributes 

2. Evaluation of CIP Findings 

3. PI Considerations 
The Commissioners - 5 - 

COORDINATION: 

The Office of the General Counsel has reviewed this paper and has no legal objection. The 

Office of the Chief Financial Officer has reviewed this paper for resource implications and has 

no objections. 

 /RA/ 

Michael R. Johnson, Director 

Office of New Reactors 

Enclosures: 

1. Construction Assessment Program Attributes 

2. Evaluation of CIP Findings 

3. PI Considerations 

WITS 200800443, 200800444, 200800445/EDATS: SECY-2008-0678, SECY-2008-0679, SECY-2008-0680 

ADAMS Accession Number: ML091970152 *via e-mail

OFFICE NRO/DCIP/CCIB DCIP/CCIB:TL TECH ED * DCIP/CCIB:BC NRR/DIRS * NRO/PMDA 

NAME TKozak 

(RPascarelli for) RPascarelli KAzariah-Kribbs JStarefos FBrown (NSanfilippo for) BGusack 

DATE 07/21/09 07/22/09 07/22/09 07/22/09 07/29/09 07/29/09 

OFFICE RII * OE * OCFO * OGC NLO* NRO/DCIP * NRO:OD 

NAME LPlisco CCarpenter 

(JWray for) 

RMitchell 

(JRand for) KWinsberg GTracy MJohnson 

DATE 07/27/09 07/29/09 07/30/09 07/31/09 07/23/09 8/14/09 

OFFICIAL RECORD COPY 
 

Enclosure 1 

Construction Assessment Program Attributes 

This enclosure provides background information on the development of the construction 

inspection program in light of lessons learned from NUREG-1055, “Improving Quality and the 

Assurance of Quality in the Design and Construction of Nuclear Power Plants: A Report to 

Congress,” and changes to the construction licensing framework as reflected in Title 10 of the 

Code of Federal Regulations (10 CFR) 52. Additionally, this enclosure details the construction 

assessment program as outlined in IMC 2505, “Periodic Assessment of Construction Inspection 

Program Results” including the near-term safety culture approach that closely resembles that 

being implemented for operating reactors as well as a longer term approach. 

 

Background

NUREG-1055 was issued May 1984 and detailed lessons learned during the early days of 

construction under 10 CFR Part 50, “Domestic Licensing of Production and Utilization Facilities”. 

This report concluded that the U.S. Nuclear Regulatory Commission (NRC) was slow to detect 

and take strong action on significant quality problems that developed during nuclear power plant 

construction projects. In addition, the NRC did not have a formal assessment process in place 

to evaluate the performance of construction permit holders. Following the accident at Three 

Mile Island, the NRC initiated an effort to better address licensee performance through the 

Systematic Assessment of Licensee Performance (SALP) program. Under the SALP program, 

the NRC periodically reviewed the overall performance of each nuclear power plant licensee 

(both construction permit holders and operating license holders) in a number of different 

functional areas. Each functional area evaluated was assigned to one of three categories to 

indicate whether more, less, or about the same level of NRC inspection and licensee attention 

was appropriate for the coming period. The SALP assessment was intended to be sufficiently 

diagnostic to provide a rational basis for assessing licensee performance, allocating NRC 

inspection resources, and providing meaningful guidance to licensee management. 

In 1991, the NRC began work to revise the construction inspection program (CIP) to address 

programmatic weaknesses that had been identified during the inspection and licensing of plants 

in the 1980s. This project was suspended in late 1994 because of the lack of nuclear power 

plant construction activities. Before that project was suspended, the staff had worked to 

document the lessons learned from previous NRC construction inspections and from reviews of 

inspection practices overseas and modular construction techniques used in the 

U.S. shipbuilding industry. During this period, the staff continued to use the SALP program to 

assess licensee performance. In 1998, the NRC suspended the SALP program and, for 

operating reactors, eventually replaced it with the Reactor Oversight Process (ROP) in 2000. 

The NRC renewed work to revise the CIP in 2001 when it formed the Construction Inspection 

Team, composed of representatives from each region, new reactor licensing staff, and 

inspection program management, and tasked it with updating the inspection and assessment 
- 2 - 

program for use in inspecting reactors to be licensed and constructed under 10 CFR Part 52, 

“Licenses, Certifications, and Approvals for Nuclear Power Plants.” 

The current effort to develop the CIP has focused on ensuring that the inspection program 

collects the information to support the Commission determination under 10 CFR 52.103(g), on 

whether the acceptance criteria in the combined license are met. It also focuses on addressing 

the various lessons learned from the NRC’s previous construction inspection experience (e.g., 

trend evaluations and early identification of problems and other lessons learned as detailed in 

NUREG-1055, “Improving Quality and the Assurance of Quality in the Design and Construction 

of Nuclear Power Plants: A Report to Congress,” issued May 1984). 

Construction Assessment Process

The construction assessment process will begin after the NRC issues a limited work 

authorization and/or a combined license; the NRC has implemented either Inspection Manual 

Chapter (IMC) 2503, “Construction Inspection Program: Inspections of Inspections, Tests, 

Analyses, and Acceptance Criteria (ITAAC),” or IMC 2504, “Construction Inspection Program—

Non-ITAAC Inspections,” both dated October 3, 2007; and there has been sufficient activity 

occurring for any assessment to be meaningful. The construction assessment process will 

continue until oversight transitions to the ROP. The agency will base the assessment period on 

the anniversary date on which the construction assessment process began. Once a plant 

transitions to the ROP, the assessment process will revert to the calendar year process outlined 

in IMC 0305, “Operating Reactor Assessment Program,” dated April 9, 2009. 

The staff plans to implement a construction response table (CRT) analogous to the ROP action 

matrix. The staff developed the CRT according to the philosophy that, within a certain level of 

performance (i.e., the Baseline Program column), licensees would address their performance 

issues without additional NRC engagement beyond the baseline inspection program. Agency 

action beyond the baseline inspection program will normally occur only if assessment input 

thresholds are exceeded. The CRT identifies the range of NRC and licensee actions and the 

appropriate level of communication for varying levels of performance. 

The overall response to licensee performance will be determined by the inputs specified in the 

option approved for implementation by the Commission. The NRC will use a graded approach 

to determine the response to the identified issues. This graded approach will result in an 

increase in sampling in the area(s) of concern; an increase in the inspections, tests, analyses, 

and acceptance criteria (ITAAC) being inspected; and/or the issuance of a confirmatory action 

letter, demand for information, and/or order. Increased inspection, whether increased sampling 

or the selection of additional ITAAC, will be conducted through the use of expanded inspections. 

The construction assessment process includes three basic parts: continuous assessment, 

quarterly assessment, and semiannual performance review (SPR) assessment. The site 

construction team will continuously assess licensee performance by evaluating violations as 

they are identified and adjust the inspection plan as necessary. Additionally, the staff will 

conduct assessments on a quarterly basis and take appropriate actions in accordance with the 

CRT. The licensee and the public will be informed of the results of the Agency’s assessment on 

a semi-annual basis. Additionally, the Agency will hold an annual public meeting in the vicinity 

of the construction site to communicate performance assessment results. 
- 3 - 

Regardless of which assessment option is approved, consistent with the Enforcement Policy, 

several types of violations will continue to require the use of traditional enforcement, including 

the possible issuance of fines. Examples include the following: 

• discrimination against workers for raising safety issues or other willful violations 

• actions that may adversely affect the NRC’s ability to monitor utility activities, including 

failure to report required information, failure to maintain accurate records, or failure to 

provide the NRC with complete and accurate information 

• incidents with actual safety consequences 

Safety Culture Approach

In order to remain aligned with agency safety culture developments, NRO representatives have 

been appropriately involved with NRC staff activities pursuant to the Commission direction 

provided in COMSECY COMGBJ-08-0001, “A Commission Policy Statement on Safety Culture,” 

dated February 25, 2008. The formulation of a long-term agency level approach to safety 

culture is ongoing. Much work remains to be accomplished on an agency level in the 

development of a long-term approach to safety culture. The staff recognizes the need to assess 

areas important to safety culture at the beginning of NRC inspections associated with licensed 

construction activities. Therefore, the staff intends to implement a near-term safety culture 

approach which closely resembles that already being implemented in the operating reactor 

assessment program. This oversight role includes documenting cross-cutting aspects during 

the course of NRC inspections, evaluating these findings against a pre-defined set of criteria to 

determine if a significant concern exists, and conducting appropriate follow-up actions applied 

using a graded approach. 

The staff determined that all 13 of the safety culture components were integral to a 

comprehensive NRC oversight of safety culture as applied to new reactor construction activities 

and would reveal themselves in an observable manner. Similar to the ROP, the staff plans to 

evaluate inspection findings against nine of the components in the baseline inspection program 

and against all 13 components in the supplemental inspection program. While the 13 safety 

culture components are applicable to construction environments, some of the aspects 

associated with each of the components are not applicable. The staff, with stakeholder input, is 

developing construction program safety culture aspects for each of the 13 components. 

Similar to the operating reactor assessment program, during periodic assessments of licensee 

performance, the staff plans to analyze CIP findings to determine if a theme existed for 

violations identified during the assessment period (analogous to the ROP’s substantive crosscutting issues). Themes for violations would be referred to as construction safety focus issues. 

To remain consistent with the ROP, the staff plans to remove construction safety focus issues 

as an input to the CRT. Thresholds for the identification of construction safety focus issues are 

being developed by the staff. 

The staff will modify IMC 2505, “Periodic Assessment of Construction Inspection Program 

Results,” and IMC 0613, “Documenting 10 CFR Part 52 Construction and Test Inspections,” 
- 4 - 

both dated October 20, 2008, to incorporate the safety culture approach outlined above. The 

staff expects to issue revisions to IMCs 2505 and 0613 by November 30, 2009. 

NRO will continue to participate in NRC staff efforts and will develop a long-term approach to 

safety culture that is consistent with the approach developed for the ROP. In accordance with 

the Commission’s direction in Staff Requirements Memorandum M081022, “Staff Requirements 

- Periodic Briefing on New Reactor Issues,” dated December 5, 2008, the NRO staff will 

continue to work with industry and other stakeholders on the development and embodiment of 

safety culture within the CIP, and it will keep the Commission informed as the approach to 

safety culture evolves. 
Enclosure 2 

Evaluation of Construction Inspection Program Findings

Adequate protection of public health and safety and the assurance of the common defense and 

security are fundamental regulatory objectives. Licensee compliance with U.S. Nuclear 

Regulatory Commission (NRC) requirements plays an important role in giving the NRC 

confidence that safety is being maintained, in that adequate protection is predicated on 

compliance with NRC requirements. When a licensee does not comply with NRC requirements, 

the NRC implements its Enforcement Policy. Once the agency identifies a noncompliance, the 

staff assesses its significance or severity. The significance of a construction inspection program 

(CIP) violation can be determined through several different methods. This enclosure describes 

the various methods the staff considered for use in determining the significance of CIP 

violations. 

Traditional Enforcement

The staff of the Office of New Reactors (NRO) has worked closely with the Office of 

Enforcement in developing a revision to the Enforcement Policy to address circumstances 

associated with new reactor construction under Title 10 of the Code of Federal Regulations

(10 CFR) Part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants.” The staff, 

with stakeholder input, has developed draft examples of violations in each of the four severity 

levels as guidance in determining the appropriate severity level for violations identified through 

the CIP, including violations that are associated with the inspections, tests, analyses, and 

acceptance criteria (ITAAC) closure and maintenance process. The staff has issued a 

proposed revision to the Enforcement Policy, including the examples of violations, for public 

comment. The staff is in the process of incorporating these comments and will forward the 

proposed policy to the Commission for approval. Using examples as guidance in dispositioning 

findings should provide transparency and predictability to the oversight of licensee performance. 

The CIP will review the adequacy of the development and implementation of construction and 

operational programs. Traditional enforcement is well suited to disposition findings associated 

with inadequate program development and implementation. The inclusion in 10 CFR Part 52 of 

inspections, tests and analyses with specific acceptance criteria as requirements presents 

objective performance criteria that do not exist in the Reactor Oversight Process (ROP) or in 

previous construction projects licensed under 10 CFR Part 50, “Domestic Licensing of 

Production and Utilization Facilities.” Thus, findings associated with ITAAC closure and 

maintenance can be objectively dispositioned using traditional enforcement. 

As discussed in Enclosure 1, regardless of which assessment option is approved, the following 

violations will continue to require the use of traditional enforcement, including the possible 

issuance of fines. Examples include the following: 

• discrimination against workers for raising safety issues or other willful violations 
- 2 - 

• actions that may adversely affect the NRC’s ability to monitor utility activities, including 

failure to report required information, failure to maintain accurate records, or failure to 

provide the NRC with complete and accurate information 

• incidents with actual safety consequences 

Significance Determination Process

In the ROP, the staff developed a method for assigning a qualitative or quantitative probabilistic 

risk characterization to inspection findings related to reactor safety. This risk characterization 

tool was the first of a set of tools that became central elements of the significance determination 

process (SDP) to determine the significance of reactor inspection findings consistent with the 

thresholds used for plant performance indicators. This allowed the staff to use both inspection 

findings and performance indicators consistently as inputs to the plant performance assessment 

portion of the ROP. Subsequently, the staff developed other SDP tools to characterize the 

safety significance of issues associated with emergency preparedness, occupational and public 

radiation safety, physical protection, fire protection, shutdown operations, containment integrity, 

operator requalification, and steam generator tube integrity. These SDP tools used qualitative 

or quantitative risk evaluation methods when possible. SDPs that could not be related to core 

damage or containment failure risk used other rationale for assigning significance. Historically, 

such other factors included those listed in Regulatory Guide 1.174, “An Approach for Using 

Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the 

Licensing Basis,” Revision 1, issued November 2002, such as maintaining defense in depth, 

compliance with regulations, engineered safety margins, and expert staff judgment. 

Given that fuel will not be loaded until construction is complete, core damage is not a direct 

consideration and findings identified during plant construction have no actual radiological public 

health and safety risk at that time. It is possible to project the risk significance finding as if the 

findings were to remain undiscovered until after plant operations began. However, NRC 

practice has been to determine the risk of findings based on the conditions that existed when 

the deficiency occurred. For instance, the ROP only uses full-power SDPs when a plant is 

above residual heat removal system entry conditions. The staff has developed a shutdown 

operation SDP for use when the plant meets residual heat removal entry conditions. 

An additional challenge in attempting to project the risk of findings into the period of plant 

operations would be the large uncertainty introduced into the process when considering the 

frequency of the initiating event and operator actions that could be taken to mitigate the 

condition. With a large uncertainty, an SDP result is less accurate and provides less value to 

the assessment process. Nonetheless, the staff considered methods such as a risk matrix, a 

construction experience risk determination model, and reliability growth models to project the 

risk of construction findings into plant operations. It is worth noting that the ITAAC prioritization 

methodology provided a risk-informed selection of ITAAC for inspection. The inclusion of 

ITAAC in combined licenses provides objective measures that the staff can use to determine the 

quality of construction activities. 
- 3 - 

Given current risk practices and the large uncertainty involved with projecting the risk of 

construction findings into the period of operations, the staff believes that if a construction SDP is 

developed, it is most practical to develop one that assesses the effect of findings on plant 

construction. Such an SDP would be deterministic in nature and would be structured similarly to 

radiation protection, emergency preparedness, security, and operator licensing SDPs developed 

and used in the ROP. The construction SDP goals should be based on successful completion 

of ITAAC. 

Risk Matrix 

The SDP for inspection findings of degraded performance of structures, systems, and 

components (SSCs) in operating reactors uses as input the estimated impact on core damage 

frequency and/or large early release frequency, along with the estimated duration of the 

degraded condition. This formulation is highly quantitative in nature and reasonably objective 

given the inputs that are assumed. For new reactor construction, it is not possible to replicate 

these elements. However, the concept of a two-dimensional risk matrix that includes a measure 

of the risk importance of the SSC in question, along with the degree of nonconformance, could 

provide elements of a somewhat objective risk-informed and performance-based construction 

oversight program. 

One dimension would be the risk importance of the SSC in question. In the Maintenance Rule 

(10 CFR 50.65, “Requirements for Monitoring the Effectiveness of Maintenance at Nuclear 

Power Plants”), as well as in 10 CFR 50.69, “Risk-Informed Categorization and Treatment of 

Structures, Systems, and Components for Nuclear Power Reactors,” probabilistic risk 

assessment importance measures such as Fussell-Vesely and risk achievement worth take on 

special meaning. These risk importance measures are available for all the new standardized 

reactor designs. Either these values or the ITAAC ranking of the SSC could be used to 

determine the risk importance of the SSC in question. A three-group categorization of risk 

importance (low, high, very high) lends itself to deriving a four-tier finding significance 

classification (green, white, yellow, and red). The three-group categorization could be increased 

to four groups to provide a more expanded scale for risk importance. 

The second dimension of the matrix would be some measure of the degree of nonconformance. 

The details of this element could involve such attributes as the following: 

• the severity of the nonconformance 

• repetition of the nonconformance 

• the number of opportunities for self-discovery that were missed 

• the duration of the nonconformance 

The staff would evaluate the findings and assign them a color corresponding with placement in 

the matrix. 
- 4 - 

Construction Experience Risk Determination Model 

NRO Office Instruction NRO-REG-112, “New Reactor Construction Experience,” dated 

March 31, 2009, includes a methodology for screening construction or operating events for 

safety significance and applicability to new reactor designs. The results of this determination 

are used to identify appropriate NRC followup actions, including trending. The screening 

process measures the safety significance of construction or operational events, because of 

design or construction errors, based on two main factors: (1) the degradation of barriers (i.e., 

reduction in defense in depth), and (2) the likelihood that the failure would not be detected 

before operation or the period of time it remained undetected during operation. Additionally, the 

staff considers other factors that may increase the safety significance level on a scale of 1 to 5. 

These additional factors include common-cause failures, programmatic deficiencies, and 

occurrences of similar events. For screening purposes, all defects are assumed to stay hidden 

until after operations begin. 

The staff considered this approach as a possible model for an “SDP-like” process for new 

reactors. A key weakness in this approach is that the process requires a deficiency to be 

discovered during operations. Therefore, this approach would not be applicable to the first built 

plant and would be of very limited use for the next few plants. Another challenge is the number 

of deficiencies that the staff could not process through this model. Although this approach 

would apply to hardware deficiencies, it is not applicable to programmatic deficiencies. 

A possible use of this process may be as an input to a potential performance indicator that 

measures appropriate licensee response to NRC generic communications, if the agency issued 

one as result of the deficiency. Another possible use may be as a collection or database of 

information that could be used as a basis for a longer term project for developing an “SDP-like” 

process. 
Enclosure 3 

Performance Indicator Considerations 

This enclosure provides a discussion of staff activities and considerations in developing 

performance indictors for construction inspection. The objective of the Reactor Oversight 

Process (ROP) for operating reactors is to monitor performance in three broad strategic areas: 

reactor safety (avoiding accidents and reducing the consequences of accidents if they occur), 

radiation safety for both plant workers and the public during routine operations, and protection of 

the plant against sabotage or other security threats. To measure plant performance, the 

operating reactor oversight program focuses on seven specific cornerstones that support the 

safety of plant operations in the three broad strategic areas. 

For each cornerstone, the U.S. Nuclear Regulatory Commission (NRC) develops findings from 

inspections and licensees collect performance indicator (PI) data. The staff evaluates 

inspection findings for safety significance using a significance determination process and 

compares PI data against prescribed thresholds. The staff then assesses the resulting 

information and determines an appropriate NRC response using the guidelines in an action 

matrix, which typically includes supplemental inspections for selected issues. The agency takes 

enforcement action on significant inspection findings, as appropriate. 

The NRC developed the ROP with the benefit of four decades of operational experience and, 

generally speaking, steadily improving plant performance. Before implementing the ROP, 

industry (Nuclear Management and Resources Council/Nuclear Energy Institute/Institute of 

Nuclear Power Operations) and the NRC (Office for Analysis and Evaluation of Operational 

Data, Office of Nuclear Regulatory Research) maintained plant operational PIs that were used 

in various ways to assess overall industry and individual plant performance. These PIs were 

used as a basis to develop the PIs that were implemented as part of the ROP assessment 

process. PIs are a means of obtaining information related to the performance of certain key 

attributes in each of the cornerstone areas. They indicate problems that, if uncorrected, may 

increase the probability and/or the consequences of an off-normal event. The staff uses data 

submitted by each licensee to calculate PI values and then compares these values to objective 

thresholds to determine the performance band associated with the values. 

The NRC developed the significance determination process to help inspectors determine the 

safety significance of inspection findings. In developing the operating reactor performance 

assessment process, one of the tasks was to establish thresholds for PIs and corresponding 

thresholds for inspection findings, so that indications of performance degradation obtained from 

inspection findings and from changes in PI values could be put on an equal footing. The 

concept for setting these performance thresholds included the consideration of risk and the 

regulatory response to different levels of licensee performance. Regulatory Guide 1.174, “An 

Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific 

Changes to the Licensing Basis,” served as the basis for establishing these performance 

thresholds and adopted core damage frequency and large early release frequency as the 

metrics for the characterization of risk. 
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The nuclear power construction industry has been dormant and, as a result, is much less 

mature than the operating reactor industry. While it is possible to develop PIs for construction 

performance, there will be no history for the PIs, and establishing appropriate thresholds will be 

very challenging without an accumulated history of data. 

In response to Commission direction in Staff Requirements Memorandum M081022, “Staff 

Requirements - Periodic Briefing on New Reactor Issues,” dated December 5, 2008, the staff 

attempted to identify viable construction PIs. The staff first analyzed ROP PIs to determine if 

any could be applicable to a construction environment. Table 1 includes the results of this 

analysis. The staff concluded that it may be possible to implement a modified version of the PI 

for occupational exposure control effectiveness. No other ROP PIs appeared to apply in a 

construction environment. 

The staff then attempted to identify viable construction PIs. The staff used the following basis, 

the same as that used for initial ROP PI development: 

• capable of being measured objectively 

• thresholds to guide NRC and licensee actions 

• reasonable sample of performance 

• valid and verifiable indication of performance in the area being measured 

• encouragement of appropriate licensee and NRC actions 

• provision of sufficient time for the NRC and the licensee to correct performance 

deficiencies before the deficiency poses an undue risk to public health and safety 

The staff identified approximately 60 potential construction PIs and binned them into three 

categories, with Category A having the most likelihood of development. PIs in Category B may 

be implemented with fundamental changes, and those in Category C were not likely to be 

implemented. Generally, an answer of “no” to three or more of the above criteria resulted in the 

assignment of a PI to Category C; an answer of “no” to two criteria resulted in the assignment of 

a PI to Category B; and an answer of “no” to one criterion resulted in the assignment of the PI to 

Category A. Of the approximately 60 potential construction PIs considered by the staff, 13 were 

determined to be either Category A or B. The staff sought stakeholder feedback regarding 

possible PIs in several public meetings and, after further deliberations, determined that it may 

be possible to develop the following PIs: 

• failure to adequately assess or respond to generic communications 

• overdue pending design changes 

• overdue safety-significant corrective action program issues 

• due date extensions to safety-significant corrective action program issues 
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• ineffective corrective actions 

• number of re-opened inspections, tests, analyses, and acceptance criteria (ITAAC) 

• number of errors resulting from inadequate training 

• extent of condition review 

• ratio of NRC-identified findings to licensee-identified findings 

• occupational exposure control 

During development of the ROP PIs, the staff conducted extensive benchmarking activities. 

The NRC used 17 plants as case studies to validate the selected equipment PI and thresholds. 

It performed similar benchmarking for emergency planning, radiological effluents, and 

occupational exposure controls. To determine the green/white threshold (for operational PIs), 

the staff needed to determine the acceptable level of performance. It applied a statistical 

approach for plants with an acceptable performance (a generically achievable level of 

performance) while developing the threshold. For drill and exercise performance, it determined 

the white band by using two standard deviations below the average, and the yellow band by 

using three standard deviations below the average. The evaluation was a measure of the 

successful classifications/protective action recommendations divided by the total opportunities. 

The following seven questions represent the extent to which the staff examined the overall 

impact of the ROP PIs: 

1) Do the PIs differentiate among superior, average, declining, and “watch list” plants, 

as determined at the Senior Management Meeting (SMM)? 

2) How effective are individual PIs at differentiating among plants with different levels of 

performance (from the SMM)? 

3) Do the PIs demonstrate timely response? 

4) Do the PIs show declining trends for plants identified as declining by the SMM? 

5) Do the PIs show declining trends before accident sequence precursor events? 

6) How well does the set of PIs conform to those selected by Arthur Andersen for use in 

the trending methodology that was used in the SMM process? 

7) Do small decreases in the green/white thresholds capture more of the watch list and 

declining plants? 

The establishment of a risk-informed threshold for construction PIs has proven to be particularly 

challenging. As mentioned previously, the staff initially developed the ROP PI thresholds based 

on years of historical licensee performance data collected by the NRC and industry. The staff 

has been unable to identify similar data for plants under construction. No performance data 

exists against which to benchmark new construction PI thresholds. The applicable regulatory 

requirements for construction activities call for the licensee to demonstrate that all ITAAC are 
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met. The staff will continue to explore PIs focused on the licensee’s successful completion of 

ITAAC. 

The staff understands that utilities planning new construction projects intend to develop and 

implement PIs to monitor construction activities. The staff plans to monitor industry’s use of PIs 

and will incorporate the experience gained during these activities as directed by the 

Commission. 
- 5 - 

Table 1 Comparison of Existing PIs for the ROP to the 

New Reactor Construction Environment 

PI Description New Construction 

Environment 

(1) Unplanned 

SCRAMS per 

7,000 Critical 

Hours 

Monitors the number of unplanned 

SCRAMS—measures Human Error, 

Procedure Quality, Design and Equipment 

Performance 

No critical operations 

(2) Unplanned 

SCRAMS with 

Complications 

Subset of unplanned SCRAMS while 

critical that require additional operator 

actions—measures Human Error, 

Procedure Quality, Design and Equipment 

Performance 

No power operations 

(3) Unplanned 

Power Changes 

per 7,000 

Critical Hours 

Monitors unplanned power changes 

(excluding SCRAMS) potential to 

challenge safety functions, considered to 

be a leading indicator of declining 

performance – measures Human Error, 

Procedure Quality, Design, and Equipment 

Performance 

No power operations 

(4) Mitigating 

System 

Performance 

Index 

Monitors readiness of important safety 

systems – measures Configuration 

Control, Equipment Performance, and 

Human Performance 

Systems will be placed in 

isolation or standby once 

ITAAC are completed, no 

operations in a condition when 

the safety-significant equipment 

is required to be available; no 

potential to affect CDF. 

(5) Safety 

System 

Functional 

Failures 

Equipment failures that could have 

prevented: 

(a) Ability to shutdown or maintain 

reactor shutdown 

(b) Remove residual heat 

(c) Control release of radioactive 

material 

(d) Mitigate the consequence of an 

accident 

No direct comparison; new 

definition can be a measure of 

ITAAC preservation for safetysignificant systems 

(6) Reactor 

Coolant System 

(RCS) Specific 

Activity 

Monitors the integrity of fuel cladding – 

measures Design Control, Configuration 

Control, Cladding Performance, Procedure 

Quality, and Human Performance 

No power operations, no source 

term production, no challenge 

to fuel 

(7) Reactor 

Coolant 

Leakage 

Monitors integrity of RCS pressure 

boundary – measures RCS equipment and 

Barrier Performance 

Only relevant once RCS 

boundary is established, 

construction cycle not impacted.
- 6 - 

(8) 

Drill/Exercise 

Performance 

Timely and accurate licensee performance 

in drills and exercises (notifications and 

protective action recommendation)—

measures facilities and equipment, 

procedure quality, and emergency 

response organization (ERO) 

Only one drill required before 

fuel load 

(9) ERO Drill 

Participation 

Percentage of key ERO members – 

measures facilities and equipment, 

procedure quality, and ERO performance 

Little demand for requalification 

before transition to ROP 

(10) Alert and 

Notification 

System (ANS) 

Reliability 

Monitors reliability of offsite ANS, link to 

public for implementation of protective 

action recommendation - measures 

facilities and equipment 

Not required before fuel load, 

not relevant during construction 

(11) 

Occupational 

Exposure 

Control 

Effectiveness 

Control of access to and work activities 

within radiologically significant areas of the 

plant – measures plant facilities/equipment 

and instrumentation, program/process, and 

human performance 

Radiological sources limited to 

radiography before reactor 

startup 

(12) 

RETS/ODCM 

Radiological 

Effluent 

Occurrence 

Assesses the performance of radiological 

effluent control program – measures plant 

facilities/equipment and instrumentation, 

program/process, and human performance 

No radiological effluents before 

fuel load 

(13) Protected 

Area Equipment 

Index that compares the amount of time 

closed circuit television cameras and 

intrusion detection system are unavailable, 

as measured by compensatory hours, to 

the total hours in the period 

No current security 

requirements for construction