NRC INSPECTION MANUAL	VPO

INSPECTION MANUAL CHAPTER 2506

CONSTRUCTION REACTOR OVERSIGHT PROCESS

GENERAL GUIDANCE AND BASIS DOCUMENT

Issue Date: 10/29/09	i	2506

2506-01	PURPOSE	1

2506-02	OBJECTIVES	1

2506-03	APPLICABILITY	1

03.01	cROP Programs	1

03.02	cROP Implementation	2

2506-04	DEFINITIONS	2

04.01	General	2

04.02	Terms Associated with Safety Culture.	9

04.03	Enforcement Terms.	10

2506-05	RESPONSIBILITIES AND AUTHORITIES	14

05.01	Director, Office of Nuclear Reactor Regulation (NRR).	14

05.02	Director, Vogtle Project Office (VPO).	14

05.03	Directors, Technical Divisions, NRR	14

05.04	Director, Division of Construction Oversight, Region II.	15

05.05	Regional Administrators, Host Regions	15

2506-06	REQUIREMENTS	15

06.01	Baseline Inspection Program	15

06.02	Baseline Inspection Program Completion	18

06.03	Plant Specific Supplemental and Reactive Inspections	19

2506-07	GUIDANCE	19

07.01	Construction Regulatory Oversight Framework	19

07.02	Construction Inspection Program (CIP)	21

07.03	Early Site Permit Audits/Inspections	21

07.04	Pre-Combined License (Pre-COL) Audits/Inspections	21

07.05	Inspections Subsequent to LWA/COL Issuance	22

07.06	Vendor Inspection Program	23

07.07	Inspection Planning	24

07.08	Staff Evaluation of non-ITAAC Combined License Conditions	25

07.09	Staff Review of Regulatory Commitments	25

07.10	Staff Review of Licensee Actions to Address Orders	25

07.11	Changes during Construction	26

07.12	Regulatory Treatment of Non-Safety Systems	26

07.13	Reliability Assurance Program	27

07.14	Reportability Under 10 CFR Part 50.55(e) and 10 CFR Part 21	27

07.15	Documenting Inspection Results	28

07.16	Construction Project Resource Estimate	28

07.17	ITAAC Closeout Process	29

07.18	ITAAC Maintenance	31

07.19	10 CFR 52.103(g) Finding	31

07.20	ITAAC Hearings and Interim Operations	31

07.21	Construction Assessment Program	32

07.22	Construction Enforcement Program	33

07.23	NRC Allegation Program	33

07.24	Construction Experience Program (ConE)	33

07.25	Annual cROP Self-Assessment	34

07.26	Transition from cROP to ROP	34

2506-08	REFERENCES	35

Issue Date:  11/25/20	i	2506

Exhibit 1 - Construction Reactor Oversight Process Overview........................................E1-1

Exhibit 2 - Construction Regulatory Oversight Framework.............................................E2-1

Appendix A - Construction Inspection Program Guidance..........................................AppA-1

Appendix B - Construction Inspection and Assessment Program Bases........................AppB-1

Attachment 1 - Acronyms......................................................................................Att1-1

Attachment 2 - Revision History for IMC 2506 ..........................................................Att2-1

Issue Date:  11/25/20	ii	2506

2506-01	PURPOSE	

01.01	This Inspection Manual Chapter (IMC) describes the U.S. Nuclear Regulatory Commission (NRC) Construction Reactor Oversight Process (cROP) for commercial nuclear power plants under construction.

01.02	This IMC describes the basis for the NRC’s development of the cROP. 

01.03	This IMC provides information for program documents such as IMCs and inspection procedures (IPs).

2506-02	OBJECTIVES	

02.01	To describe cROP programs and processes and provide cROP implementation requirements.

02.02	To provide the bases for the cROP. 

2506-03	APPLICABILITY	

03.01	cROP Programs.  The cROP consists of the following programs to oversee applicants and licensees of new reactor construction:

a. Construction Inspection Program (CIP)

b. Construction Assessment Program

c. Construction Enforcement Program

Several additional programs and processes interact with the cROP, including: 

a. Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) Closure Verification Process 

b. Vendor Inspection Program

c. NRC Allegation Program

d. Construction Experience Program (ConE)

e. NRC Open Government Plan (Communications)

Exhibit 1, “Construction Reactor Oversight Process Flowchart,” provides an overview of the cROP and how each of the individual programs and processes interact. 

The cROP is expected to be dynamic and to respond to changes in the nuclear power industry and construction experience.  Therefore, the program office expects the regions and inspectors to identify challenges in implementing the program, and to recommend improvements to the program for consideration by the program office.

NRC shall conduct its management involvement and oversight of site activities at reactor facilities that are under construction in accordance with IMC 0102, “Oversight and Objectivity of Inspectors and Examiners at Reactor Facilities.”  

Issue Date:  11/25/20	18	2506

03.02	cROP Implementation.  The NRC implements the cROP when an applicant announces its intent to submit an application for an early site permit (ESP), a limited work authorization (LWA), a construction permit or a combined license (COL).  The cROP will remain in effect until regulatory oversight for the plant is transitioned to the Reactor Oversight Process (ROP).

The degree to which the NRC implements the cROP depends on the application or license status and the amount of ongoing activities that are associated with applications or licenses.  For instance, if an applicant only requests an ESP, then only inspections pursuant to IMC 2501, "Construction Inspection Program:  Early Site Permit" may be necessary.  On the other hand, if the NRC issues a COL, then the NRC will implement all aspects of the cROP.

2506-04	DEFINITIONS	

04.01	General.	

a.	Act.  The Atomic Energy Act of 1954 (68 Stat. 919) including any amendments thereto.

b.		Annual Assessment Cycle.  The assessment period from January 1st through December 31st of each year.

c.	Assessment Inputs.  Information considered in the assessment process to determine appropriate NRC actions.

d.	Assessment Letter.  A letter from the NRC to a licensee that communicates assessment-related information.  Assessment letters include assessment follow-up letters and annual assessment letters.

e.	Assessment Period.  A period that contains four full consecutive calendar quarters.  For example, the first quarter assessment period is from April 1 of the previous year until March 31 of the current year (end of the first quarter); the second quarter assessment period is from July 1 of the previous year until June 31 of the current year (end of the second quarter); the third quarter assessment period is from October 1 of the previous year until September 30 of the current year (end of the third quarter) and the end-of-cycle assessment period is from January 1 until December 31 of the same calendar year.  Note that, in lieu of formal quarterly assessments, regional staff may use routine engagement with a licensee (inspection planning and scheduling meetings, project readiness group meetings, drop-ins and public meetings) and inspection follow up activities of previously identified inspection findings to gage and evaluate licensee performance.  Findings (FIN) that are held open longer than two quarters and have not been closed are also considered in the assessment period, even if they were identified prior to start of the respective assessment period.

f.	Applicant.  A person or an entity applying for a license, permit, or other form of Commission permission or approval under Title 10 *Code Federal Regulations* (10 CFR) Part 50 or Part 52.  

g.	Basic Component.  See definition in 10 CFR Part 21, “Reporting of Defects and Noncompliance.”

h.	Combined license (COL).  A combined construction permit and operating license with conditions for a nuclear power facility issued under Subpart C of Part 52.

i.	Construction.  See definitions in 10 CFR 50.2 and 10 CFR 50.10.  Applicable IMCs describe the application of these definitions to NRC construction assessment and enforcement programs.

j.	Construction Action Matrix (CAM).  A table that categorizes various levels of licensee construction performance and identifies the range of NRC and licensee actions and the appropriate level of communication for these various levels of performance.

k.	Construction Action Matrix Deviation.  Any regulatory action taken that is inconsistent with those discussed in Inspection Manual Chapter 2505, Section 07.02.

l.	Construction Action Matrix Inputs.  Inspection findings that the NRC uses to determine a plant’s CAM column.

m.	Construction Action Matrix Summary.  A description of a plant’s CAM column assignment, the basis for a plant being in Columns 2, 3, 4, or 5, and a brief description of the NRC’s current level of regulatory oversight at the plant.

n.	Construction Activities.  See definition in 10 CFR 50.10.

o.	Construction Assessment Program.  The program that the NRC implements at each plant that is under construction to allow for the NRC to arrive at objective conclusions about a licensee’s effectiveness in assuring construction quality, provide for predictable responses to performance issues, and to clearly communicate performance assessment results to the public.

p.	Construction Deficiency Report.  As described in 10 CFR 50.55(e), an official notification to the NRC of a construction defect or failure to comply that could create a substantial safety hazard, were it to remain uncorrected.  A “substantial safety hazard” means a loss of safety function to the extent that there is a major reduction in the degree of protection provided to public health and safety from the facility.

q.	Construction Inspection Program (CIP).  The program comprised of inspections that the NRC conducts in accordance with IMCs 2501, 2502, 2503, 2504, and 2507.

r.	Construction Inspection Program Information Management System (CIPIMS).  The database that provides the means to plan, document, report, and track NRC construction inspections and their results.

s.	Contractor.  Any organization or individual under contract to furnish items or services to a licensee engaging in an NRC-regulated activity.  It includes the terms consultant, vendor, manufacturer, supplier, fabricator, constructor, and sub-tier levels of these organizations.

t.		Degraded Cornerstone.  A cornerstone that has three or more white inputs or one yellow input.

u.	Design Acceptance Criteria (DAC).  A set of prescribed limits, parameters, procedures, and attributes upon which the NRC relies, in a limited number of technical areas, in 

	making a final safety determination to support a design certification.  DAC are part of the ITAAC inventory for a given design.

v.	Design Control Document (DCD).  A repository of information comprising the Standard Plant Design.  The DCD also provides the design-related information to be incorporated by reference into the 10 CFR Part 52 Appendices containing the design certification rules (i.e., Appendices A, B, C and D).

w.	Documentation.  Any written, pictorial, or electronic information describing, defining, specifying, reporting, or certifying activities, requirements, procedures, or results.

x.	Early site permit (ESP).  Commission approval, issued under subpart A of Part 52, for a site or sites for one or more nuclear power facilities.  An ESP is a partial construction permit.

y.	Engineering Design Verification (EDV) Inspection.  An inspection to:  (1) verify that the design authority (e.g., the organizations contracted by an NRC applicant to provide engineering, procurement, and construction support) has developed processes that allow for the complete and accurate transfer of the high level design information and performance requirements specified in the Final Safety Analysis Report (FSAR) into detailed procedures, specifications, calculations, drawings, procurement, and construction documents, in a manner consistent with the requirements of Appendix B to 10 CFR Part 50; (2) verify that the design authority has developed processes to ensure changes to the design are adequately controlled; and (3) verify, through a detailed technical review of selected systems, that the design authority’s implementation of its design and design control processes has produced detailed procedures, specifications, calculations, drawings, procurement, and construction documents that are consistent with NRC regulations, the FSAR, and the NRC’s Safety Evaluation Report (if issued).

z.	Family of ITAAC.  A grouping of ITAAC that are related through similar construction processes, resulting products, and general inspection attributes.

aa.	Final Safety Analysis Report.  A report that is included in an application for an operating license that presents information describing the facility, presents the design bases and the limits on its operation, and presents a safety analysis of the structures, systems, and components (SSCs) and of the facility as a whole.

ab.	Held-Open Finding.  A safety-significant finding that is considered a Construction Action Matrix input for more than two quarters.

ac.	Inspection.  (1) An NRC activity consisting of examination, observation, or measurement to determine if an applicant, licensee, contractor, or vendor conforms with requirements and standards.  (2) Applicant, licensee, contractor, or vendor activity consisting of examination, observation, or measurements to determine the conformance of materials, supplies, components, items, parts, systems, processes or structures to pre-determined quality requirements.

ad.	Inspection Document.  Any material obtained or developed during an inspection that is considered to be an NRC record.  Inspectors should review IMC 0620, “Inspection Documents and Records,” for clarification on how materials become agency records.

ae.	Integrated Inspection Report.  A construction inspection report that combines inspection items from multiple inspections (resident, regional, etc.) conducted during a specific time period.

af.	Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC).  Those ITAAC identified in the combined license that if met by the licensee are necessary and sufficient to provide reasonable assurance that the facility has been constructed and will operate in conformity with the license, the provisions of the Atomic Energy Act, as amended, and the Commission’s rules and regulations.

ag.	ITAAC Attributes.  A number of common, descriptive characteristics for each ITAAC that can be analyzed and weighted by a methodology that allows the ITAAC to be prioritized for inspection planning. 

ah.	ITAAC Closeout.  The process by which the licensee affirms that an ITAAC has been satisfactorily completed.

ai.	ITAAC Closeout Verification.  The NRC process which evaluates the licensee’s affirmation of satisfactory ITAAC completion.

aj.	ITAAC Matrix.  An inspection planning tool that identifies groups (i.e., “families”) of ITAAC, based upon common characteristics, which facilitate the ITAAC sampling process and provide a consistent model for the targeting of ITAAC at plants of a similar design.

ak.	Licensee.  A person or entity authorized to conduct activities under a license (e.g., early site permit, construction permit, combined license, or limited work authorization) issued by the Commission.

al.	Licensee Agent.  An entity to which a licensee has delegated the work of establishing and executing its Quality Assurance (QA) program, or parts thereof.  For example, a contractor who does not have a license who conducts construction activities does so as an agent of the licensee who holds the applicable license.  This term also applies to applicants prior to the issuance of a COL.  Per 10 CFR Part 50, Appendix B, Criterion I, the applicant or licensee may delegate to others, such as agents, the work of establishing and executing the QA program, or any part thereof, but shall retain responsibility for the QA program. 

am.	Limited Work Authorization.  The authorization provided by the NRC under 10 CFR 50.10 allowing the driving of piles, subsurface preparation, placement of backfill, concrete, or permanent retaining walls within an excavation, installation of the foundation, including placement of concrete, any of which are for an SSC of the facility for which either a construction permit or combined license is otherwise required.

an.	Long-standing Cross-Cutting Issue (CCI).  A CCI that has been open for six or more consecutive quarters.

ao.	Multiple Degraded Cornerstones.  Two or more cornerstones that are degraded in any one quarter.

ap.	NRC Quality Assurance Guidance.  Guidance developed or endorsed by the NRC through issuance of regulatory guides, review standards, or national standard documents that discusses acceptable methods of implementing a QA program consistent with Appendix B to 10 CFR Part 50 requirements.  Standard Review Plan (SRP) 17.5, “Quality Assurance Program Description - Design Certification, Early Site Permit and New License Applicants,” provides QA guidance for COL application reviews.

aq.	NRC Record.  Any written, electronic, or photographic record under legal NRC control that documents the policy or activities of the NRC or an NRC licensee.  Also see the definition in 10 CFR Part 9.

ar.	Objective Evidence.  Any documented statement of fact, other information, or record, either quantitative or qualitative, pertaining to the quality of an item or activity, based on direct observations, measurements, or tests that can be verified.

as.	Observation.  For the cROP, a factual detail noted during a power reactor construction inspection.  Observations not directly related to a finding may only be documented if prescribed by an appendix to IMC 0613, “Power Reactor Construction Inspection Reports,” or by a specific inspection procedure.

at.	Plant Performance Summary (PPS).  A document prepared by Region II and used during the end-of-cycle, and Agency Action (if applicable) review meetings that describes assessment inputs and other pertinent information used to develop a conclusion about a plant’s safety performance.

au.	Pre-construction activity.  Any activity conducted prior to issuance of a COL or LWA by the applicant or contracted suppliers on behalf of the applicant associated with a proposed ITAAC for safety-related components or portions of the proposed facility and occurring at other than the final, in-place location at the facility.

av.	Pre-operational Tests.  Tests performed by or under the direction of the licensee to demonstrate the proper functioning and conformance to design requirements of SSCs.  Containment leak rate tests may fall in this category or may be combined with the containment integrity test.  Completion of preoperational testing frequently forms the contractual basis for custody transfer from the constructor to the operator.

aw.	Program element.  The means that exist to implement elements (e.g., procedures, facilities, equipment, or training) of the licensee’s emergency preparedness program.

ax.	Quality Assurance.  Quality Assurance (QA) comprises all those planned and systematic actions necessary to provide adequate confidence that an SSC will perform satisfactorily in service.  QA includes quality control (QC).

ay.	Quality Assurance Manual.  A compilation of quality assurance documents that defines the QA policy and program, describes the methods by which the policy will be implemented through procedures and instructions, and identifies the parties responsible for implementation.

az.	Quality Assurance Program Description.  A description of the QA program to be applied to the design, fabrication, construction, and testing of the SSCs of the facility.  10 CFR Part 50, Appendix B (Appendix B), "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," sets forth the requirements for QA programs for nuclear power plants and fuel reprocessing plants.  The description of the QA program for a nuclear power plant or a fuel reprocessing plant shall include a discussion of how the applicable requirements of Appendix B will be satisfied.

ba.	Quality Assurance Program/Quality Assurance Commitments.  These terms relate to the description of the QA program, or any part thereof, as required by 10 CFR 52.79(a)(25) in each application for a COL for a nuclear power facility.  The description of the QA program must include a discussion of how the applicable requirements of Appendix B to 10 CFR Part 50 have been and will be satisfied, including a discussion of how the QA program will be implemented.

bb.	Quality Control (QC).  QC comprises QA actions related to the physical characteristics of an SSC.  This provides a means to control the quality of the SSC to applicant-predetermined requirements.

bc.	Reactive Vendor Inspection.  Inspections performed for the purpose of obtaining additional information and verifying adequate corrective actions on reported problems or deficiencies involving vendor supplied products or services.  Reactive inspections are typically performed in response to a specific problem identified by any group within the NRC (e.g., including headquarters, the regional offices), or in response to allegations or other identified problems (e.g., 10 CFR Part 21 or 10 CFR 50.55(e) reports).

bd.	Regulatory Performance Meeting.  A meeting held between a licensee and the NRC to discuss corrective actions associated with safety-significant CAM inputs.

be.	Repetitive Degraded Cornerstone.  Three open white inputs or one open yellow input in a single cornerstone for more than five consecutive quarters with at least one of the quarters having:  (1) four or more white inputs (the additional white inputs can be from any cornerstone), or (2) one yellow and one white input.  The additional white input can be from any cornerstone.

bf.	Routine Vendor Inspection.  Inspections performed to verify effective implementation of a facility’s QA program used to supply basic components to the nuclear industry.

bg.	Safety Evaluation Report.  The safety evaluation report (SER) provides the technical, safety, and legal basis for the NRC’s disposition of a license request (i.e., COL, early site permit, and design certification) or license amendment request.

bh.	Safety-related structures, systems and components (SSC).  Those SSCs that are relied upon to remain functional during and following design basis events to assure:

1.	The integrity of the reactor coolant pressure boundary

2.	The capability to shut down the reactor and maintain it in a safe shutdown condition; or

3.	The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the applicable guideline exposures set forth in 10 CFR 50.34(a)(1) or 10 CFR 100.11.

bi.	Standard Design.  A design that is sufficiently detailed and complete to support certification in accordance with Subpart B of 10 CFR Part 52 and that is usable for a multiple number of units or at a multiple number of sites without reopening or repeating the review.

bj.	Standard Design Certification.  Standard design certification, design certification, or certification means a Commission approval, issued under Subpart B of 10 CFR Part 52, of a final standard design for a nuclear power facility.  This design may be referred to as a certified standard design.

bk.	Startup Testing.  The testing program conducted after the authorization to load fuel.  It includes initial fuel loading and pre-criticality tests, and continues until the plant reaches commercial operating status at or near its licensed power rating.  The Startup Test Program includes low power, physics, and power ascension testing.

bl.	Supplier.  For the purposes of this manual chapter, any organization that supplies basic components to a vendor, applicant, or holder of a 10 CFR Part 52 license.

bm.	Surveillance.  Applicant and contractor activities such as reviews, observations, inspections, and audits to determine if an item or activity conforms to QA Program commitments.

bn.	Tier 1 Information.  The portion of the design-related information contained in the generic DCD that is approved and certified by the applicable 10 CFR Part 52 appendix.  The design descriptions, interface requirements, and site parameters are derived from Tier 2 information.  Tier 1 information includes:

1.	Definitions and general provisions;

2.	Design descriptions;

3.	Inspections, tests, analyses, and acceptance criteria (ITAAC);

4.	Significant site parameters; and

5.	Significant interface requirements.

bo.	Tier 2 Information.  The portion of the design-related information contained in the generic DCD that is approved but not certified by the applicable 10 CFR Part 52 appendix.  Compliance with Tier 2 is required, but generic changes to and plant-specific departures from Tier 2 are governed by Section VIII of the applicable 10 CFR Part 52 appendix.  Compliance with Tier 2 provides a sufficient, but not the only acceptable, method for complying with Tier 1.  Compliance methods differing from Tier 2 must satisfy the change process in Section VIII of the applicable 10 CFR Part 52 appendix.  Regardless of these differences, an applicant or licensee must meet the requirement in Section III.B of the applicable 10 CFR Part 52 appendix to reference Tier 2 when referencing Tier 1.  Tier 2 information includes: 

1.	Information required by 10 CFR Parts 52.47(a) and 52.47(c), with the exception of generic technical specifications and conceptual design information;

2.	Supporting information on the inspections, tests, and analyses that will be performed to demonstrate that the acceptance criteria in the ITAAC have been met; and

3.	Combined license (COL) action items (COL license information), which identify certain matters that must be addressed in the site-specific portion of the FSAR by an applicant who references the applicable 10 CFR Part 52 appendix.  These items constitute information requirements but are not the only acceptable set of information in the FSAR.  An applicant may depart from or omit these items, provided that the departure or omission is identified and justified in the FSAR.  After issuance of a construction permit or COL, these items are not requirements for the licensee unless such items are restated in the FSAR.

4.	The investment protection short-term availability controls in Section 16.3 of the DCD.

bp.	Tier 2* means the portion of the Tier 2 information, designated as such in the generic DCD, which is subject to the change process in Section VIII.B.6 of the applicable 10 CFR Part 52 appendix.  This designation expires for some Tier 2* information under paragraph VIII.B.6 of the applicable 10 CFR Part 52 appendix.

bq.	Type Test.  A test on one or more sample components of the same type and manufacturer to qualify other components of the same type and manufacturer.  A type test is not necessarily a test of the as-built SSCs.

br.	Unannounced Inspection.  An NRC inspection where the organization is not notified by the NRC until the NRC staff arrives at the organization’s facility or site.

bs.	Vendor.  Any company or organization that provides products such as material, equipment, components or services to be used in an NRC-licensed facility or activity.  In certain cases the vendor may be an NRC licensee (e.g., a nuclear fuel fabricator) or the product may have NRC certificates (e.g., a transportation cask).

bt.	Verification of ITAAC Closure, Evaluation and Status (VOICES).  The database that provides the means to verify, evaluate and track ITAAC closure request reviews.

04.02	Terms Associated with Safety Culture.	

a.	Cross-Cutting Area.  Fundamental performance attributes that extend across all of the cROP cornerstones of safety.  These areas are human performance, problem identification and resolution, and safety conscious work environment (SCWE).

b.	Cross-Cutting Aspect.  A performance characteristic of a finding that is the most significant causal factor of the performance deficiency as described in IMC 0613.

c.	Cross-Cutting Theme.  For the cross-cutting areas of problem identification and resolution and human performance, a cross-cutting theme exists when at least six inspection findings are assigned the same cross-cutting aspect during the second quarter or end-of-cycle assessment period.  The findings should be representative of more than one cornerstone; however, given the significant inspection effort applied to the Construction/Installation Cornerstone, a cross-cutting theme can exist consisting of inspection findings associated with only this one cornerstone.  A cross-cutting theme also exists when there are at least 20 findings in the Human Performance area or at least 12 findings in the Problem Identification and Resolution area during a second quarter or end-of-cycle assessment period.

A cross-cutting theme exists in the area of SCWE if at least one of the following three conditions exists in an 18-month period (i.e., the current second quarter or end-of-cycle assessment period and the two quarters preceding that period):  (1) a finding with a documented cross-cutting aspect in SCWE and the impact on SCWE was not isolated, or (2) the licensee has received a chilling effect letter, or (3) the licensee has received correspondence from the NRC that transmitted an enforcement action with a Severity Level (SL) I, II, or III, and that involved discrimination, or a confirmatory order that involved discrimination.

d.	Cross-Cutting Issue (CCI).  A CCI is a cross-cutting theme which has been identified in at least three consecutive assessment or assessment follow-up letters.  

e.	Safety-Conscious Work Environment (SCWE).  An environment in which personnel feel free to raise safety concerns without fear of retaliation, intimidation, harassment, or discrimination.

f.	Nuclear Safety Culture.  The set of core values and behaviors resulting from a collective commitment by leaders and individuals to emphasize safety over competing goals to ensure protection of people and the environment.

g.	Safety Culture Assessment.  A comprehensive evaluation of the assembly of characteristics and attitudes related to all of the cross-cutting aspects described in IMC 0613.  Individuals performing the evaluation can be qualified through experience and formal training.

1. An independent safety culture assessment is one performed by qualified individuals that have no direct authority and have not been responsible for any of the areas being evaluated.  For example, staff from another of the licensee’s facilities, or corporate staff who have no direct authority or direct responsibility for the areas being evaluated.

2.	A third-party safety culture assessment is one performed by qualified individuals who are not members of the licensee’s organization or utility operators of the plant.  Licensee team liaison and support activities are not team membership.

04.03	Enforcement Terms.	

a. Accept-as-is.  A hardware disposition which may be imposed for a nonconformance when it can be established that the discrepancy will result in no adverse condition and that the item under consideration will continue to meet all engineering functional requirements including performance, maintainability, fit, and safety.  A design change may be required as a result of the accept-as-is disposition.  

For the purposes of this definition, a nonconformance is a deficiency in characteristics, documentation, or procedures which renders the quality of an item unacceptable or indeterminate.  Examples of a nonconformance include physical defects, test failures, incorrect or inadequate documentation or deviation from prescribed manufacturing processing, inspection, or test procedures.

b.	Apparent Violation (AV).  A violation of regulatory requirements that is being considered for potential escalated enforcement action.

c.	Closed Item.  A matter previously reported as an inspection finding, a deviation, a nonconformance, an item reported by the licensee (e.g., 10 CFR Part 21 report, an ITAAC maintenance item, 10 CFR Part 50.55(e) construction deficiency report or licensee event report), or an unresolved item that the inspector concludes has been satisfactorily resolved based on information obtained during the current inspection.

d.	Common Cause.  Multiple failures (i.e., two or more) of proper installation of equipment, construction of structures or processes attributable to a shared cause.

e.	Consequence.  The actual or potential outcome of an identified problem or condition.  

f.	Construction Issue.  An inspection result that is dispositioned in accordance with the guidance in IMC 0613, also referred to as Issue of Concern.

g.	Contributing Causes.  The causes that by themselves would not create the problem but are important enough to be recognized as needing corrective action.  Contributing causes are sometimes referred to as causal factors.  Causal factors are those actions, conditions, or events which directly or indirectly influence the outcome of a situation or problem.

h.	Escalated Enforcement Action.  Severity Level I, II, and III Notice of Violation (NOV); White, yellow or red findings; civil penalties; NOVs to individuals; Orders to modify, suspend, or revoke NRC licenses or the authority to engage in NRC-licensed activities; and Orders issued to impose civil penalties.

i.	Extent of Cause.  The extent to which the root causes of an identified problem have impacted other plant construction processes, equipment, or human performance.

j.	Extent of Condition.  The extent to which the actual condition exists with other plant construction processes, equipment, or human performance.

k.	Finding (FIN).  A performance deficiency of more than minor significance.  A finding may or may not be associated with regulatory noncompliance and, therefore, may or may not result in a violation.  There are two types of findings that can be identified through the implementation of the CIP:  ITAAC Finding and Construction Finding.

1.	An ITAAC Finding is a finding that is identified through the implementation of the CIP that is associated with a specific ITAAC and is material to the ITAAC acceptance criteria.

Issue Date:  11/25/20	18	2506

2.	A Construction Finding is a finding that is identified through implementation of the CIP that is not an ITAAC finding.

l.	Issue of Concern.  An inspection result that is dispositioned in accordance with the guidance in IMC 0613.

m.	Licensee-Identified.  For cROP, licensee-identified findings are those findings that are not NRC-identified or self-revealing.  Most, but not all, licensee-identified findings are discovered through a licensee program or process.  Some examples of licensee programs or processes that will likely result in such findings are the identification and documentation of findings (e.g., procedural violations, procedure inadequacies, etc.) by craft workers or licensee or contractor supervision during routine construction activities, construction QA activities, self-assessments, independent assessments, audits and surveillances.  Additional examples may include preoperational testing, start-up testing, hydrostatic testing, nondestructive testing, EP drills, and critiques conducted by or for the licensee.

n.	Minor Violation.  A violation that is less significant than a SL IV violation.  Minor violations do not warrant enforcement action and are not normally documented in inspection reports.  However, minor violations must be corrected.

o.	Non-Cited Violation (NCV).  A non-recurring, typically non-willful, Severity Level IV violation that is not subject to formal enforcement action if, for a reactor licensee, the licensee places the violation in a corrective action program to address recurrence and restores compliance within a reasonable period of time and, for all other licensees, the licensee corrects or commits to correcting the violation within a reasonable period of time.

The use of NCVs for self-revealing and NRC-identified violations as part of the enforcement process is predicated on a licensee having an adequate CAP into which identified issues are entered and effectively resolved in a timely manner.  Because the CAP at construction sites will be new and implemented initially by individuals with limited experience with the new program and because construction will involve program implementation by contractors, the NRC will delay the use of NCVs for self-revealing and NRC-identified violations pending confirmation that the new program is adequate and being effectively implemented.

p.	Notice of Deviation (NOD).  A written notice describing a licensee’s failure to satisfy a commitment where the commitment involved has not been made a legally binding requirement.  An NOD requests that a licensee provide a written explanation or statement describing corrective steps taken or planned, the results achieved, and the date when corrective action will be completed.  

q.	Notice of Nonconformance (NON).  A written notice describing the failure of a licensee’s contractor to meet commitments that have not been made legally binding requirements by the NRC (e.g., a commitment made in a procurement contract with a licensee or applicant as required by 10 CFR Part 50, Appendix B).  If the contractor deliberately fails to meet the terms of a procurement contract, the NRC may issue a violation under the Deliberate Misconduct Rule in 10 CFR 50.5.  NONs request that non-licensees provide written explanations or statements describing:  (1) the reason for the 

Issue Date:  11/25/20	18	2506

	noncompliance, or if contested, the basis for disputing the noncompliance; (2) the corrective steps that have been taken and the results achieved; (3) the corrective steps that will be taken; and (4) the date when corrective actions will be completed.

r.	Notice of Violation.  A formal, written citation in accordance with 10 CFR 2.201 that sets forth one or more violations of a regulatory requirement.

s.	NRC-Identified.  NRC-Identified findings are those that are found by NRC inspectors that the licensee was not previously aware of or had not been previously documented in the licensee’s corrective action program.  NRC-identified findings also include previously documented licensee findings to which the inspector has significantly added value.  Added value means that the inspector has identified a previously unknown significant weakness in the licensee’s classification, evaluation, or corrective actions associated with the licensee’s correction of a finding.

t.	Performance Deficiency (PD).  An issue that is the result of a licensee not meeting a requirement or standard where the cause was reasonably within the licensee’s ability to foresee and correct, and therefore should have been prevented.  A performance deficiency can exist if a licensee fails to meet a self-imposed standard or a standard required by regulation, thus a performance deficiency may exist independently of whether a regulatory requirement was violated.  Additional discussion can be found in Appendix B, “Issue Screening,” of IMC 0613.

u.	Program critical attribute.  An element of a program that is established to ensure that a regulatory requirement is met.  Program descriptions are contained in the FSAR.

v.	Regulatory Commitment.  An explicit statement of “intent” or “agreement” to take a specific action agreed to or volunteered by a licensee, where the statement has been submitted in writing on the docket to the NRC.  This may include a commitment in the licensee’s application, a response to a NOV, etc.

w.	Repetitive Violation.  See definition in the NRC Enforcement Policy.

x.	Requirement.  A legally binding obligation such as a statute, regulation, license condition, technical specification, or an order.  

y.	Root Causes.  The basic reasons (i.e., hardware, process, or human performance) for a problem, which if corrected, will prevent recurrence of that problem.

z.	Safety-Significant.  Having greater than very low (i.e., green) safety significance.

aa.	Self-Revealing.  For the cROP, self-revealing findings are those that become selfevident and require no active and deliberate observation by the licensee or NRC inspectors to determine whether a change in process or equipment capability or function has occurred.  Self-revealing findings become readily apparent to either NRC or licensee personnel through a readily detectable degradation in the material condition, capability, or functionality of equipment and require minimal analysis to detect.  Some examples of self-revealing findings include failure of equipment or instrumentation to operate properly during testing that was not related to the purpose of the test (e.g., inadequate foreign material controls cause the failure) and violation of radiography 

Issue Date:  11/25/20	18	2506

	exclusion area requirements that are subsequently identified through an electronic dosimeter alarm.

ab.	Significance Determination Process (SDP).  A characterization process that is applied to inspection findings to determine their safety significance.

ac.	Unresolved Item (URI).  An issue of concern about which more information is required to determine (a) if a performance deficiency exists, (b) if the performance deficiency is More-than-Minor, or (c) if the issue of concern constitutes a violation.  Such a matter may require additional information from the licensee or cannot be resolved without additional guidance or clarification/interpretation of the existing guidance.

ad.	Violation.  The failure to comply with a legally binding requirement, such as a statute, regulation, order, license condition, or technical specification.

ae.	Work activity.  Processes implemented during the construction of the facility in areas such as but not limited to structural, piping, electrical, and foundations.

2506-05	RESPONSIBILITIES AND AUTHORITIES.	  

05.01	Director, Office of Nuclear Reactor Regulation (NRR).	  

a. 	Provides overall program direction for the cROP.

b.	Develops and directs the implementation of policies, programs, and procedures for inspecting applicants, licensees, and other entities subject to NRC jurisdiction.

c.	Assesses the effectiveness, uniformity, and completeness of implementation of the cROP.

d.	Provides overall direction for the NRC vendor inspection program.

e.	In the event of a pandemic, concurs on the regions’ recommendations to the modification to the inspection program in accordance with the direction provided under Appendix A of this IMC.

05.02	Director, Vogtle Project Office (VPO).

Manages inspection program development within NRR, develops and prepares revisions to the cROP, oversees regional implementation, and serves as the NRR contact with the regional offices for program development and implementation.

05.03	Directors, Technical Divisions, NRR	.

a. Assists the Director, VPO in developing the technical content of and reviewing periodic revisions to the requirements and guidance contained in IPs related to their areas of technical expertise.

Issue Date:  11/25/20	18	2506

b.	Ensures their staff provides technical assistance in support of ITAAC closure and other inspection activities.

05.04	Director, Division of Construction Oversight, Region II.	

a.		Provides program direction for implementation of the cROP elements performed by Region II.

b.	Ensures, within budget limitations, that the regional office staff includes adequate numbers of inspectors in the various disciplines necessary to carry out the inspection program described in this chapter, including that which may be needed for regional supplemental and reactive inspections.

c.	Directs the implementation of the supplemental inspection program.

d.	Applies inspection resources, as necessary, to deal with significant issues and problems at specific plants.

e.	Ensures that line managers assign inspectors who are appropriately trained and have the necessary knowledge and skills to successfully implement IPs. 

f. Determines that a pandemic situation which affects inspection resource availability has occurred and recommends modification to the inspection program.

g.	Determines the scope and frequency of operational program implementation inspections that are conducted prior to the 10 CFR 52.103(g) finding.

05.05	Regional Administrators, Host Regions	.

a.	Provides assistance with construction inspections to Region II for plants in their respective region within budgeted resources.

b.	Ensures, within budgeted resources, that their staff leads inspections of select operational program inspections at facilities under construction in their respective region as assigned by this IMC.

2506-06	REQUIREMENTS	

06.01	Baseline Inspection Program	.  The NRC will complete the baseline inspection program at all reactors under construction prior to the Commission’s affirmative 10 CFR 52.103(g) decision.  The program requires inspections of licensee performance in the six cornerstones of safety.  Region II has the responsibility to complete the baseline inspection program.

The objectives of the baseline inspection program are (1) to provide a sufficient basis to support the finding, in accordance with 10 CFR 52.103(g), that the acceptance criteria in a combined license have been met; and (2) to develop confidence in the licensee’s programmatic controls.

Issue Date:  11/25/20	18	2506

Thus, the baseline inspection program consists of ITAAC inspections and construction and operational program inspections.

ITAAC inspections are conducted to provide confidence that licensee’s ITAAC completion and verification processes are effective and provide reasonable assurance that licensee ITAAC completion notifications are sufficient and accurate.  Construction program inspections confirm that construction products provide an adequate level of quality.  Operational program inspections verify that operational programs are consistent with their description in the FSAR.

In implementing these objectives, the program allows for flexible scheduling to permit the adjustment, including expansion or reduction of inspection scope, and includes ITAAC across a full range of significance with effort being weighted toward those with higher significance.

a. ITAAC Inspections.

There are two key elements to ITAAC inspections.  The first element is inspection of a broad range of ITAAC-related activities.  This includes inspection of activities and SSCs associated with the following ITAAC:

1.	Targeted DCD ITAAC

2.	DAC ITAAC

3.	Targeted Emergency Preparedness ITAAC

4.	Targeted Security ITAAC

5.	Targeted Site Specific ITAAC.  A separate panel will select the Site Specific Targeted ITAAC after the COL is issued.

The second element of ITAAC inspections is inspection of ITAAC-related construction processes.  This is accomplished through implementation the ITAAC IPs.  The staff developed IPs for each of the rows and columns in the ITAAC matrix.  These procedures constitute the construction baseline IPs applicable to ITAAC inspections.  They are written to provide inspection requirements and guidance for a wide range of SSCs from all reactor types.  Therefore, not every step will apply to every SSC nor will every step apply to each reactor type.  During the planning for ITAAC inspections, inspectors will identify those steps related to a given ITAAC and include them in their inspections.

Each ITAAC will be assigned to a lead Region II inspection branch.  The assigned branch is required to plan inspections of targeted ITAAC and verify that these inspections are completed. 

Identification of targeted ITAAC is not intended to restrict the inspection of non-targeted construction activities.  Although not required for completion of the baseline inspection program, the staff may inspect non-targeted ITAAC when appropriate, or as a substitute for targeted SSCs, as a result of construction schedule changes where otherwise no inspections would be performed.

Issue Date:  11/25/20	18	2506

To maximize efficiency, the staff should consider bundling ITAAC whenever there are opportunities to witness multiple ITAAC during the same inspection or trip, especially if the activities are being conducted at a foreign location.

b. Construction Program Inspections (including Preoperational Testing Inspections).

IMC 2504, “Construction Inspection Program ‐ Inspection of Construction and Operational Programs,” contains the guidance for construction program inspections and describes the governing IPs to be used in the construction baseline inspection program.  Region II staff will coordinate and conduct these inspections.

c. Operational Program Inspections.

Appendix B of IMC 2504 lists the operational programs that are required to be implemented by a license condition or regulation.  A licensee must implement each operational program on or before the program’s milestone date.  Operational program milestones are listed in the attachment to the SRP, Section 13.4 “Operational Programs.”  The technical staff review the scope and content of the required operational programs during the license application review process.  When the license is approved, the staff approves the operational programs.  The NRC is required to conduct inspections to ensure the approved scope and content of the required operational programs are contained in the licensee’s operational program documents and procedures.  Operational program inspections completed prior to the 10 CFR 52.103(g) finding are a part of the cROP baseline inspection program.  Those operational program inspections completed after the 10 CFR 52.103(g) finding are part of the ROP inspection program.

IMC 2504, Appendix B also includes the IPs used to ensure that the approved scope and content of the required operational programs are contained in the licensee’s operational program documents and procedures.  NRR Office Instruction LIC-114, “Title 10 of the *Code of Federal Regulations*(10 CFR) Section 52.103(g) Finding and Communication Process” describes the process to provide the status of operational programs to the Commission just prior to the issuance of the 10 CFR 52.103(g) finding.

When a licensee implements an operational program prior to the 10 CFR 52.103(g) finding, the region may, at its discretion, use the applicable reactor oversight process (ROP) baseline IPs in addition to or in lieu of IMC 2504 IPs.  The staff should disposition findings identified during these inspections, prior to the 10 CFR 52.103(g) finding, using the cROP SDP in IMC 2519, “Construction Significance Determination Process.”  The findings will be assigned to the Operational Programs cornerstone in the cROP Action Matrix.

Some operational programs have implementation milestones just prior to, or after, the 10 CFR 52.103(g) finding.  Therefore, licensees may not have developed and implemented some operational programs in time for inspection prior to the 10 CFR 52.103(g) finding.  Those programs will become the responsibility of the host region.  The staff should disposition findings identified during these inspections, after the 10 CFR 52.103(g) finding, in accordance with IMC 0609, “Significance Determination Process,” and IMC 0612, “Issue Screening.”  The staff will conduct inspections of other licensee activities, after the 10 CFR 52.103(g) finding, as directed by IMC 2515, “Light-Water Reactor Inspection Program - Operations Phase,” and IMC 2514, “AP1000 Reactor Inspection Program - Startup Testing Phase.”

At sites with an operating unit, the licensee may opt to implement similar operational programs as those that are already in use (e.g., corrective action program, radiation protection program, etc.).  In this case, the region may, at its discretion, choose to not inspect to the same level as sites with no operational units.

d. License Conditions

Each licensee must comply with the conditions specified and incorporated into its license.  In accordance with Section 07.08 of this IMC, the staff conducts an evaluation to determine if inspection is needed to verify the licensee’s conformance with the license conditions.  The staff must inspect each license condition that is identified during the evaluation conducted in accordance with Section 07.08, as part of the baseline inspection program.

06.02	Baseline Inspection Program Completion.  The baseline inspection program is complete when the ITAAC inspections and the construction and operational program inspections meet the following criteria:

a. ITAAC-related Work Inspections

The staff has inspected all targeted ITAAC.  For each targeted ITAAC the lead Region II branch chief will make the determination that:

1. The staff has inspected a sufficient number of SSCs related to the ITAAC.

2. No Regional or Headquarters (HQ) vendor branch ITAAC-related open inspection items (e.g., URIs, FIN, Violation (VIO), NCV, NON) exist.  

Region II management will review the conclusion of this assessment and the staff will document completion of the ITAAC inspections in CIPIMS.  The assessment process in IMC 2505 will consider these conclusions as part of making the final recommendation that the acceptance criteria in the COL have been met.

b. Construction Program Inspections 

The staff conducts construction program inspections with a cyclic frequency during construction.  The construction program inspections provide insight into the construction processes and procedures and are ongoing up until the 10 CFR 52.103(g) finding.  Open issues will be transferred to the ROP for further action.

c. Operational Program Inspections

Operational program inspections will be completed as part of the cROP if they are performed prior to the 10 CFR 52.103(g) finding or as part of the ROP inspection program if they are completed after the 10 CFR 52.103(g) finding. 

Issue Date:  11/25/20	18	2506

d. License Conditions

License conditions inspections (if required) will be completed as part of the cROP if they are performed prior to the 10 CFR 52.103(g) finding or completed as part of the ROP inspection program if they are performed after the 10 CFR 52.103(g) finding. 

06.03	Plant Specific Supplemental and Reactive Inspections.  The staff will assess plant performance using IMC 2505.  Licensees whose performance is outside the licensee response column in the CAM will receive plant specific supplemental inspections based on its assessed performance.  The depth and breadth of specific supplemental inspections chosen for implementation will depend upon the significance of the identified issues and will be conducted pursuant to the inspection procedure specified in the CAM.  Region II has the lead for conducting plant specific supplemental inspections.  In addition, Region II staff may conduct reactive inspections in response to events and issues that occur at the facility.  Reactive Inspections include inspections required for allegation response and event follow-up.  IMC 2504, Appendix C, “Response to Significant Issues or Events" contains guidance for reactive inspections.

2506-07	GUIDANCE	

07.01	Construction Regulatory Oversight Framework.  The staff used a hierarchical approach to develop a construction regulatory oversight framework that addresses the agency’s regulatory principles, as shown in Exhibit 2.  This framework starts with the NRC’s overall mission to license and regulate the Nation’s civilian use of byproduct, source, and special nuclear materials to ensure adequate protection of public health and safety.  The staff then identified those aspects of licensee performance that are important to the mission and therefore merit regulatory oversight.

Six cornerstones of safety form the framework for the cROP:  design/engineering, procurement/fabrication, construction/installation, inspection/testing, operational programs, and security programs for construction inspection and operations.  These cornerstones have been grouped into three strategic performance areas:  construction reactor safety, operational readiness, and safeguards programs.

For the construction reactor safety area, the objectives of the cornerstones of safety are defined as follows:

Design/Engineering:  The objective of this cornerstone is to ensure that licensees adequately develop and implement programs and processes for design and engineering controls.

Procurement/Fabrication:  The objective of this cornerstone is to ensure that licensees adequately develop and implement programs and processes for procurement and fabrication activities.

Construction/Installation:  The objective of this cornerstone is to ensure that licensees adequately develop and implement programs and processes to ensure the construction and installation of facilities and structures, systems, and components are in accordance with the design.

Issue Date:  11/25/20	18	2506

Inspection/Testing:  The objective of this cornerstone is to ensure that licensees adequately develop and implement programs and processes to inspect and test programs, facilities, and structures, systems, and components.

For the operational readiness area, the objective of the cornerstone of safety is defined as follows:

Operational Programs:  The objective of this cornerstone is to ensure that licensees adequately develop and implement the operational programs required by a license condition or regulation.

For the safeguards programs area, the objective of the cornerstone of safety is defined as follows:

Security programs for construction inspection and operations:  The objective of this cornerstone is to provide assurance that (1) fitness-for-duty issues do not adversely impact construction activities; and (2) the licensee’s security programs use a defenseindepth approach and can protect against the design basis threat of radiological sabotage from internal and external threats.

In addition to the cornerstones, the cROP features three “cross-cutting” areas, so named because they affect and are therefore part of each of the cornerstones.  The cross-cutting areas are Human Performance, Problem Identification and Resolution, and Safety Conscious Work Environment.  Cross-cutting aspects are defined for each of the cross-cutting areas.  This framework is based on the principle that the agency’s mission of assuring public health and safety is met when the agency has reasonable assurance that licensees are meeting the objectives of the six cornerstones of safety.  The construction inspection program is an integral part, along with assessment, and enforcement, of the cROP.  Along with the verification that all ITAAC have been completed, acceptable performance in the cornerstones, as measured by the risk-informed baseline inspection program, provides reasonable assurance that the facility has been constructed and will be operated in conformity with the license.

Another principle of the framework is that there is a level of licensee performance in the cornerstones above which the NRC does not need to engage the licensee beyond some minimum level.  When this level of licensee performance is reached, the risk-informed baseline inspection is sufficient to provide reasonable assurance of public health and safety.

The supplemental portion of the inspection program provides more diagnostic inspections of identified problems and issues beyond the baseline.  Supplemental inspections will be planned in response to issues that result in crossing a CAM threshold.  These changes to the inspection program are factored into the inspection program through the assessment program as further discussed in Section 2506-09.

10 CFR Part 50, Appendix B, Criterion 1, “Organization,” allows an applicant or licensee to delegate to others, such as contractors, agents, or consultants, the work of establishing and executing the quality assurance program.  If the licensee has contracted all or portions of the construction of the nuclear power plant, then the licensee has delegated the work of establishing and executing the QA program, or parts thereof.  In these cases, the contractor is acting as an agent of the licensee.  The NRC policy holds licensees and applicants responsible for the acts of their employees, contractors, or vendors and their employees, and the NRC may 

Issue Date:  11/25/20	18	2506

cite the licensee or applicant for violations committed by its employees, contractors, or vendors and their employees.  Consequently, citations against the agent of the licensee instead of the licensee would be very unusual.  As such, situations where this approach is considered should be reviewed by senior agency management during a Significance and Enforcement Review Panel (SERP) or other similar process.

In Volume 72 of the *Federal Register*, page 49351, on August 28, 2008 (72 FR 49351), the agency made clear the difference between suppliers and contractors performing construction, or the functional equivalent of construction.  A supplier provides basic components and does not perform construction as defined in 10 CFR Part 50.10.  Suppliers are inspected via the vendor inspection program.  Contractors performing construction, or the functional equivalent of construction, do so as agents of a licensee by assembling or installing basic components that eventually are installed in their final resting place.  Inspections of licensee agents are conducted by Region II as part of the baseline inspection program and enforcement actions are taken against the licensee, who retains ultimate responsibility for the QA program.

07.02	Construction Inspection Program (CIP).  The objectives of the CIP are to determine if appropriate quality controls are implemented in the development of applications that will be or have been submitted to the NRC; and provide reasonable assurance that the facility has been constructed and will operate in conformity with the license, the provisions of the Act, and the Commission's rules and regulations.

The CIP is conducted to support a licensing decision for an ESP application; to support a licensing decision for a COL application; and to support construction activities and the preparations for operation.  In addition, prior to and during plant construction, the NRC will conduct inspections to review vendor activities and licensee oversight of these activities.

07.03	Early Site Permit Audits/Inspections	.  IMC 2501, “Construction Inspection Program:  Early Site Permit (ESP),” provides policy and guidance for the implementation of the inspection program during licensee preparation and NRC review of ESP applications submitted under 10 CFR Part 52.  IMC 2501 also provides guidance for the inspection, assessment, and documentation of QA program implementation during geotechnical and site characterization activities (e.g., site exploration and data collection and analysis) performed by the applicant and its contractors.  IMC 2501 takes effect when an applicant applies for an ESP and continues during the review process until the NRC issues the ESP.  The NRC will implement this IMC to inspect and assess the applicant’s implementation of applicable 10 CFR Part 50, Appendix B QA requirements by the applicant or contractors on behalf of the applicant during the performance of geotechnical and site characterization activities.

07.04	Pre-Combined License (Pre-COL) Inspections	.  IMC 2502, “Construction Inspection Program: PreCombined License (Pre-COL) Phase,” provides policy and guidance for the implementation of the inspection program during NRC review of COL applications submitted under 10 CFR Part 52 and guidance for the inspection, assessment, and documentation of pre-construction activities performed by the applicant and contracted suppliers of the applicant.  The NRC will conduct inspections of an applicant once the COL application is tendered.

These inspections will continue during the application review process until a COL is issued.  This timeframe is referred to as the pre-COL phase.  In addition to pre-COL inspections 

Issue Date:  11/25/20	18	2506

conducted in support of the COL licensing process, the NRC staff will inspect the applicant’s oversight of pre-construction activities that may support the NRC’s future verification of ITAAC completion.  The specific inspections required during the pre-COL phase are listed in IMC 2502.  

NRR vendor inspection staff has the lead for the inspections conducted:

a. To verify quality processes used in the development of the COL application are adequately described, and that technical, quality, and administrative requirements important to public health and safety are effectively implemented during the design and procurement phases of pre-COL activities.

b. To verify effective implementation of the QA program, as described in the application for a COL, to provide reasonable assurance of the integrity and reliability of the COL data or analyses that would affect the performance of safety-related SSCs.

c. To verify that the applicant’s and contracted suppliers’ offsite pre-construction activities are being effectively implemented in accordance with the applicable 10 CFR Part 50 Appendix B QA requirements.  The results of these inspections may support the NRC’s future closure verification of ITAAC.

Region II has the lead for the inspections conducted to verify that the applicant’s and contracted suppliers’ onsite pre-construction activities are being effectively implemented in accordance with the applicable 10 CFR Part 50 Appendix B QA requirements.  The results of these inspections may support the NRC’s future closure verification of ITAAC.

Exceptions to these inspection responsibilities should be rare and shall be approved by the responsible HQ program office and RII Division Directors.

07.05	Inspections Subsequent to LWA/COL Issuance	.  The development of the 10 CFR Part 52 COL regulatory and inspection framework introduced the concept of ITAAC as a codified, pre-approved set of performance standards that a COL licensee is required to certify as acceptable and complete.  Subsequent to LWA/COL issuance, the NRC staff conducts inspections to review the licensee’s construction activities as the licensee completes the ITAAC.  Guidance for these inspections is contained in IMC 2503, “Construction Inspection Program:  Inspections of Inspections, Tests, Analyses and Acceptance Criteria (ITAAC) Related Work.”  

Region II will lead most ITAAC-related inspections at licensee owned and operated facilities.  Coordination of inspection activities at licensee-controlled locations (e.g., onsite or corporate headquarters) with the program office is not necessary.  The vendor inspection branch will lead most construction-related inspections at offsite facilities not owned and operated by a licensee.  On rare occasions, Region II staff may lead ITAAC-related inspections at offsite facilities that are not owned and operated by a licensee.  For those inspections:

a. The inspectors should perform direct observation of work (i.e., more than receipt inspection and document review) to provide reasonable assurance that the ITAAC is met;

b. 	Any enforcement action should be attributed to a specific licensee; and

Issue Date:  11/25/20	18	2506

c. The inspection will be approved by the program office and the responsible Region II Division Directors.

The NRC will plan inspections at facilities that are not controlled by the licensee in coordination with the vendor inspection branch to ensure that inspections are properly staffed.

Region II will use the Technical Assistance Request (TAR) process to request support from NRR technical staff, if required, to resolve inspection-related questions.  In addition, on occasion, Region II may request the host region to conduct ITAAC-related work inspections; however, Region II will maintain overall lead for the inspection.

Subsequent to LWA/COL issuance, the NRC staff conducts inspections to review the development and implementation of construction and operational programs.  IMC 2504 contains guidance for these inspections.  Region II has the lead for the IMC 2504 construction program and operational program inspections.  Region II may request the host region to conduct certain construction or operational program inspections; however, Region II will maintain overall lead for the inspection and the host region will report the results to Region II for consideration in the overall assessment of licensee construction performance.

07.06	Vendor Inspection Program	.  The Vendor Inspection Program is implemented by the vendor inspection center of expertise (COE).  The purpose, organization, and responsibilities of the vendor inspection COE are provided in the COE’s charter (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12045A064).

IMC 2507, “Vendor Inspections,” establishes the inspection program for vendors providing safety-related materials, equipment, and services to the commercial nuclear industry.  The vendor inspection staff is responsible for implementing the vendor inspection program.  The staff conducts routine and reactive inspections to verify that vendors implement QA programs that comply with regulatory requirements (e.g., 10 CFR Part 21, Appendix B to 10 CFR Part 50.)

In addition to IMC 2507, the NRC developed the “Vendor Inspection Program Plan (VIP Plan),” which establishes an overall approach, including goals, priorities, performance metrics, and resource management strategies for vendor inspection program activities.

Region II inspection staff will occasionally support the vendor inspection staff during vendor inspections for new reactor construction.  NRC staff take the following steps to ensure that new reactor construction vendor inspections are properly coordinated:

a. The vendor inspection branches develop inspection targets following the guidance in the VIP Plan.  When a scope of supply includes ITAAC related items, vendor inspection branch chiefs will request support from Region II construction inspection branch chiefs.

b. Region II branch chiefs may provide a recommendation for the inspection of a vendor that is conducting ITAAC-related work.  

c. The vendor inspection schedule will be available on the vendor inspection SharePoint site.  Notes on this schedule will be included to provide inspection plans.

Issue Date:  11/25/20	18	2506

d. The vendor inspection and Region II branch chiefs will coordinate ITAAC related vendor inspections and resources.

Vendor inspection reports may support future closure verification of ITAAC.  To the extent possible, Region II will use the results of vendor inspections to inform its inspections at specific sites.  By maintaining awareness of vendors and their activities, Region II will improve its ability to effectively and efficiently conduct CIP inspections.

The vendor inspection staff will also conduct Engineering Design Verification (EDV) inspections.  These inspections verify that the design authority (1) has developed processes for complete and accurate transfer of high level design information and performance requirements specified in the FSAR in a manner consistent with the requirements of Appendix B to 10 CFR Part 50, (2) has developed processes to ensure changes to the design are adequately controlled, and (3) has produced detailed procedures, specifications, calculations, drawings, procurement, and construction documents that are consistent with NRC regulations, the FSAR, and the NRC’s SER (if issued).  The staff conducts EDV inspections pursuant to IMC 2507 and IP 37805, “Engineering Design Verification Inspections.”

07.07	Inspection Planning.  Region II has responsibility for developing an inspection plan for each unit under construction.  The inspection plan contains entries for all targeted ITAAC inspections, selected non-targeted ITAAC inspections, and construction and operational program inspections for the unit under construction.  The staff updates inspection plans as necessary to align with the construction activities ongoing at the site.

These plans should outline the following:

a.	Inspection activity’s description

b.	Explanation of when the inspection should take place

c.	The type of activity (e.g., pre-operational test, short term, long-term)

d.	Estimate of hours

e.	Desired inspection resource (e.g., Resident inspector, Regional inspector, HQ support, etc.)

f.	Intent of each inspection activity.  This can include quantitative or qualitative goals, a description of mandatory inspection aspects, etc.

g.	Bundling of ITAAC with similar design commitments and or inspections, tests, and analyses where inspection efficiencies may be achieved and supported by the licensee’s construction schedule.

The plans may recommend IPs for use during the inspection.  However, in general, inspectors should use the IPs that relate to the ITAAC family for which the inspection is being conducted (e.g., an inspection of an ITAAC in family 01A should be conducted using IPs 65001.01 and 65001.A).

Issue Date:  11/25/20	18	2506

The NRC staff can increase or decrease the number of construction activities and SSCs planned for inspection based on program results.  The NRC staff will review and adjust the number of SSCs to be inspected as part of the annual performance review assessment process described in IMC 2505, “Periodic Assessment of Construction Inspection Program Results.”  This can be done at any stage of the annual assessment process, including during continuous, quarterly, or end-of-cycle reviews.  The NRC staff should document reasons for deviations from the original plans in CIPIMS.

07.08	Staff Evaluation of non-ITAAC Combined License Conditions.  A combined license contains conditions that are part of the plant’s licensing basis.  Shortly after issuance of a combined license, the program office will perform an evaluation to determine if inspection is needed to verify the licensee’s conformance with the license conditions.  The team will accomplish the following actions:  

a. Determine if the staff should conduct an inspection to verify the licensee’s conformance with license conditions. 

b. Evaluate each license condition that is determined to require inspection to determine if it is covered by an existing ITAAC inspection (IMC 2503), construction or operational program inspection (IMC 2504), or start-up testing inspection (IMC 2514) and identify the applicable inspection procedure. 

c. Identify the applicable inspection procedure to be implemented for the inspection of any remaining license conditions that should be inspected or submit an IMC/IP change request if new or revised inspection guidance is necessary. 

d. The program office will document the license condition evaluation results and provide a copy of the report to Region II for incorporation into the site inspection plan.

07.09	Staff Review of Regulatory Commitments.  Regulatory commitments are explicit statements that describe specific actions agreed to, or volunteered by, a licensee and submitted in writing on the docket to the NRC.  Regulatory commitments are appropriate for matters that are of significant interest to the staff, but do not warrant either legally binding requirements or inclusion in Updated Final Safety Analysis Reports (UFSARs) or programs subject to a formal regulatory change control mechanism.  Nevertheless, the regulatory process appropriately uses commitments in many instances and the NRC expects licensees to honor these commitments.  The licensee is responsible for creating and maintaining configuration control of all regulatory commitments made to the NRC.  The NRC staff will manage regulatory commitments made by COL holders through the implementation of NRR Office Instruction LIC-105, “Managing Regulatory Commitments Made by Licensees to the NRC.”

07.10	Staff Review of Licensee Actions to Address Orders.  The NRC exercises the authority to issue an order when deemed necessary to either gain compliance with existing regulations (i.e., Enforcement Orders) or to further provide reasonable assurance of adequate protection of public health and safety, and the common defense and security (i.e., Non-Enforcement Orders).  The Enforcement Manual contains guidance for the issuance of orders.  NRR-COM-104, “NRR interfaces with the Office of Enforcement,” contains guidance for NRR roles and responsibilities

Issue Date:  11/25/20	18	2506

in the processing of enforcement actions, including orders.  An order may require follow-up inspection to verify completion of the specified licensee actions.

The staff will determine if an inspection is necessary to close an order on a case-by-case basis.  If the staff determines that inspection and closure of an order is necessary, then the responsible Region II branch chief should ensure that the action items are included in the appropriate tracking system (e.g., CIPIMS) for future inspection and closure.  The inspection results should be documented in an inspection report.

07.11	Changes during Construction.  Interim Staff Guidance on Changes during Construction (CdC) Under 10 CFR Part 52 (COLISG025) describes the license amendment request (LAR) preliminary amendment request (PAR) process.  The staff developed this process for maintaining licensing basis configuration control and in order to avoid unnecessary construction delays related to CdC arising after the issuance of the COL and before the 10 CFR 52.103(g) finding.

A Construction Operations Engineer will review PARs to determine if there is an impact to inspections or to an ITAAC and will communicate the description of the proposed change and impact on associated ITAAC to Region II to inform its inspection process.

07.12	Regulatory Treatment of Non-Safety Systems.  Unlike the current generation of light-water reactors or the evolutionary advanced light-water reactors (ALWRs), the AP1000 plant design uses passive safety systems that rely on natural forces, such as density differences, gravity, and stored energy, to supply safety injection water and provide core and containment cooling.  These passive systems do not include pumps.  They do include some active valves, but all the safety-related active valves require either direct current (dc) safety-related electric power (supplied by batteries), are air operated (and fail safe on loss of air), or are check valves.  All active systems (i.e., systems requiring alternating current (ac) power to operate) are designated as non-safety related, except for the instrumentation and control (I&C) systems, which use safety-related ac power converted from safety-related dc power.  Passive systems should be able to perform their safety functions, independent of operator action or offsite support, for 72 hours after an initiating event.  After 72 hours, non-safety or active systems may be required to replenish the passive systems or to perform core and containment heat removal duties directly.

The AP1000 includes active systems that provide defense-in-depth (or investment protection) capabilities for reactor coolant system makeup and decay heat removal.  In existing plants, as well as in the evolutionary ALWR designs, many of these active systems are designated as safety-related.  The residual uncertainties associated with passive safety system performance increase the importance of active systems in providing defense-in-depth functions to back up the passive systems.  Recognizing this, the NRC and the Electric Power Research Institute (EPRI) developed a process to identify important active systems and to maintain appropriate regulatory oversight of those systems, called the Regulatory Treatment of Non-Safety Systems (RTNSS).  This process does not require that the active systems brought under regulatory oversight meet all safety-related criteria, but rather that these controls provide an ITAAC inspection level of confidence that active systems having a significant safety role are available when they are challenged.

This section will be updated to address other designs (such as the Economical Simplified Boiling Water Reactor, a Small Modular Reactor, or an Advanced Non-Light Water Reactor) if a licensee decides to start construction of one of these designs.

Issue Date:  11/25/20	18	2506

07.13	Reliability Assurance Program.  The Reliability Assurance Program (RAP) applies to SSCs that are significant contributors to plant safety.  The COL holder uses the plant- and site-specific probabilistic risk analyses, industry operating experience, component failure data bases, expert panels and other methods to identify the SSCs to be included in RAP.  The RAP is implemented in two stages.  The first stage applies to reliability assurance activities that occur before the initial fuel load, known as the Design Reliability Assurance Program (D-RAP).  The second stage applies to reliability assurance activities for an operating plant.  The D-RAP ensures that the reliability of SSCs within the scope of the RAP is properly considered and designed into the plant and is implemented through the reactor design, procurement, fabrication, construction, and preoperational test activities and programs.  For passive reactor designs, SSCs scoped into the regulatory treatment of non-safety systems (RTNSS) are also included in the D-RAP.

The initial scope of the RAP is established during design certification, standard design approval, or manufacturing license approval.  A COL applicant extends the scope of the RAP to address safety-significant SSCs that were not included in the original scope.  For example, design of the ultimate heat sink of an active plant may not be within the scope of the certified design referenced by the COL applicant.  The DRAP should be fully described in the application so that it can be reviewed by the NRC staff in accordance with the Standard Review Plan.  If the program is not fully described in the COL application or is not acceptable to the NRC staff based on that description, the COL will contain ITAAC that will be verified by the NRC staff as follows:

a.	For the safety-related SSCs within the scope of the RAP, confirm that the initial issuance of engineering for procurement and construction has been performed under a QA Program that meets 10 CFR Part 50 Appendix B requirements.

b.	For the non-safety-related SSCs within the scope of the RAP, confirm that the initial issuance of engineering for procurement and construction has been performed under a reliability assurance program that the staff has reviewed and accepted. 

Note that the D-RAP ITAAC is meant to verify proper programmatic control of the design process, not the design itself.  Field inspections of SSCs within the scope of the RAP should be conducted pursuant to ITAACs that are associated with those SSCs and not under the D-RAP ITAAC.

07.14	Reportability Under 10 CFR Part 50.55(e) and 10 CFR Part 21.  10 CFR Part 50.55(e) and 10 CFR Part 21 contain the regulatory requirements for reporting defects and noncompliance and implement Section 206 of the Energy Reorganization Act (ERA).  72 FR 49352, dated August 28, 2007, provided a description of how Section 206 of the ERA is implemented through the regulations contained in 10 CFR 50.55(e) and 10 CFR Part 21 for plants licensed pursuant to Part 52.  For COL holders, 10 CFR 50.55(e) is applicable prior to an affirmative 10 CFR 52.103(g) finding and 10 CFR Part 21 is applicable after this finding.  10 CFR 50.55(e) is also applicable to entities that are performing construction or the functional equivalent of construction; however, when an entity is acting as agent of the licensee, the licensee retains ultimate reporting responsibility.  As noted in Section 07.18 below, the licensee is responsible to identify when an ITAAC completion determination basis is in question and a post-closure notification to the NRC is required.  The staff inspects licensee issued 10 CFR 50.55(e) and 10 CFR Part 21 reports per IP 36100.01 and IP 36100.  The requirements of 10 CFR Part 21 

Issue Date:  11/25/20	18	2506

and 10 CFR 50.55(e) are nearly identical.  However, 10 CFR 50.55(e) does not include suppliers.

The Federal Register Notice (FRN) also addressed reportability for other aspects of regulatory life other than construction and operation, such as an early site permit (ESP) and a design certification/rule (DCD).  The FRN position was that the ESP and design certification were the equivalent of a basic component that a licensee was going to use to construct a plant.  As such, 10 CFR Part 21 reporting requirements would be applicable for defects.  These evaluations would be for items specific to the NRC approved document (ESP or DCD) referenced.  For example, specific deviations in the revision of the AP1000 DCD approved by the NRC for the design certification contained in 10 CFR Part 52 Appendix D would be required to be evaluated under 10 CFR Part 21, and if determined to be a defect, reported to the NRC by Westinghouse Electric Company.

07.15	Documenting Inspection Results.  The purpose of reporting inspection results is to document the inspection scope and the findings identified while conducting the inspection.  The NRC does not have objective criteria for evaluating inspector observations.  Therefore, NRC staff will not document inspector observations in baseline inspection reports or incorporate them into the assessment process.  The scope of daily activities conducted by the resident inspectors does not require documentation in inspection reports.  Issues identified during inspections will be documented in accordance with the requirements in IMC 0613, “Power Reactor Construction Inspection Reports,” and IMC 0617, “Vendor and Quality Assurance Implementation Inspection Reports,” as appropriate.

07.16	Construction Project Resource Estimate	.  The initial direct inspection effort estimate is 35,000 hours per unit over the life of the construction project.  This estimate is based on the inspection hours that were expended during construction of the last several completed reactor units.  This number includes 15,000 hours for ITAAC-completion inspections, 10,000 hours for programmatic and operational program inspections, 5,000 hours for reactive inspections above the baseline program in response to licensee performance issues, allegations, and non-performance issues/events, and 5,000 hours for technical support for construction inspection.  See the following table for a summation of the inspection effort estimate:

Notes:

a. ITAAC direct inspections include all the necessary vendor or field inspections, engineering analyses, technical assistance requests, report reviews needed to close 

Issue Date:  11/25/20	18	2506

the ITAAC, pre and post-COL inspections, DAC follow-up, and design change reviews (15,000 inspector hours).

b. Inspection of Construction and Operational Programs include QA verifications, IMC-2504 construction programs, pre-operational inspections, and operational program readiness reviews (10,000 hours).

c. Reactive and allegation inspections include inspections required for allegation response, baseline inspection sample expansion, or the follow-up of performance problems and non-performance issues and events.

d. Engineering resources for non-ITAAC inspections, reactive inspections, and design verification may be used, in part, to verify licensee compliance with post-COL FSAR commitments and license conditions.  A panel of technical experts will provide a recommendation to management about which, if any, of these post-COL commitments warrant independent verification.  If needed, the panel will also recommend what type of verification (e.g., direct inspection, engineering inspection) is most appropriate.

e. Direct inspection hours do not include hours for preparation, documentation, and inspector travel and which are also billed to the licensee.

07.17	ITAAC Closeout Process	.  Office Instruction LIC-211, Inspections, Tests, Analyses, and Acceptance Criteria Closure Verification Process (ICVP),” contains guidance for closing ITAAC.  An issued COL contains ITAAC that must be performed by the licensee.  Once the licensee has performed an ITAAC, the licensee will close that ITAAC.  For each closed ITAAC, in accordance with 10 CFR 52.99(c)(1), the licensee is required to notify the NRC that the prescribed inspections, tests, and analyses have been performed and that the prescribed acceptance criteria have been met.  That notice must have sufficient information to support these two conclusions and is called an ITAAC closure notification (ICN).

The staff reviews all ICNs to determine if the ITAAC can be verified as completed.  This process is led by the program office and closely coordinated with Region II, OGC, NRR technical divisions, and NSIR.

During the ICN review, the staff will verify that the NRC inspections of the ITAAC are complete and that all related ITAAC inspection findings are closed.  The results of the reviews of the ICNs by the staff and the documentation of the completion of the CIP for COL ITAAC facilitate the staff recommendation regarding the 10 CFR 52.103(g) finding on whether all of the COL acceptance criteria are met.

The lead Branch Chief will identify in CIPIMS whether or not the planned inspections for the given ITAAC have been completed.  This acknowledgement will be accomplished by indicating [YES] in the “all inspections complete” block on the inspection planning/strategy page in CIPIMS.  A YES in the block will be understood to indicate:  1) the lead Branch Chief has verified all planned inspections are complete and that there are no plans to conduct future inspections of the respective ITAAC (for non-targeted ITAAC, a YES in the block will indicate that planned inspections, if any, are complete and that there are no plans to conduct future inspections of the respective ITAAC); 2) no inspection-related open items affect the ITAAC’s closure; and 3) Region II concurs with initiating the ICN review process to close the given 

Issue Date:  11/25/20	18	2506

ITAAC.  If the ITAAC should not be closed, the Region II lead Branch Chief shall notify the HQ branch chief in charge of ITAAC closure.

An ITAAC shall be re-opened if it is subsequently determined that the acceptance criteria had never been met or the associated inspections, tests, and analysis were invalid.  If an ITAAC is re-opened, then the responsible Region II branch chief should notify the HQ branch chief in charge of ITAAC closure and uncheck the “all inspections complete” block on the inspection planning/strategy page in CIPIMS.  Once the ITAAC is completed, a new ITAAC Closure Notification (ICN) is required to be submitted.  The responsible branch chief should follow the ITAAC Closeout instructions above.

In general, review of an ICN will not be delayed based on the existence of an allegation related to the respective ITAAC.  However, if an allegation has been or appears most likely to be substantiated and the NRC has concluded that the issue will likely be an ITAAC Finding, then processing of the ICN will be delayed.  In this case, VPO will coordinate the ICN response and subsequent NRC action with Region II (e.g., rejection of the ICN and documentation of the ITAAC Finding in an inspection report).

The staff is required to publish Federal Register Notices (FRNs) of successful ITAAC completion at intervals determined by the staff.  The periodic FRNs will inform the public that the inspections, tests, and analyses of one or more ITAAC have been performed and their acceptance criteria have been met.  The staff’s determination that the acceptance criteria of all of the COL ITAAC are met precedes the 10 CFR 52.103(g) finding.  In SRM-SECY-13-0033, “Allowing Interim Operation under Title 10 of the *Code of Federal Regulations* Section 52.103,” dated July 19, 2013, the Commission delegated to the staff the making of the 10 CFR 52.103(g) acceptance criteria finding for all ITAAC, irrespective of the pendency of any hearing.

In October 2009, the NRC issued Regulatory Guide 1.215, “Guidance for ITAAC Closure under 10 CFR Part 52.”  This guide describes a method the staff considers acceptable for use in satisfying the requirements for documenting the completion of ITAAC.  In particular, this guide endorses the methodologies described in NEI 08-01, “Industry Guideline for ITAAC Closure Process under 10 CFR Part 52,” which provides an approach that COL holders may use to satisfy NRC regulatory requirements under 10 CFR 52.99 related to the completion and closure of ITAAC for new nuclear power plants.  Regulatory Guide 1.215, Revision 1, was issued in September 2012, and captures ITAAC Maintenance Rule requirements.

In accordance with 10 CFR 52.99(c)(3), no later than 225 days prior to initial fuel loading, the licensee is required to notify the NRC that the inspections, tests and analyses will be performed, and the acceptance criteria will be met for all uncompleted ITAAC prior to operation.  The uncompleted ITAAC notification (UIN) must provide sufficient information to demonstrate that the prescribed inspections, tests and analyses will be performed and the prescribed acceptance criteria will be met, including, but not limited to, a description of the specific procedures and analytical methods to be used for performing the ITAAC.  The 225-day notifications are primarily for the public to review to provide prima facie evidence for a possible hearing on the completion of ITAAC that the acceptance criteria of affected ITAAC or other ITAAC are not met.

After all ITAAC have been completed, the Director of NRR, in consultation with the appropriate Regional Administrators, will inform the Commission that all ITAAC have been met.  NRC inspection results, together with the information submitted by the licensee, will be the foundation of the staff's recommendation to the Commission in support of its finding on whether the acceptance criteria in the COL have been met.

Issue Date:  11/25/20	18	2506

07.18	ITAAC Maintenance	.  Completion of COL ITAAC will be accomplished by the licensee over a prolonged period.  For some ITAAC, this will mean significant time will elapse between the initial determination that an individual ITAAC is closed and the Commission finding, in accordance with 10 CFR 52.103(g), on whether all of the acceptance criteria are met.  An important aspect of the ITAAC maintenance process is to confirm that the acceptance criteria continue to be met for all ITAAC verified as completed until the Commission makes the 10 CFR 52.103(g) finding.  The staff recognizes that normal maintenance will be needed on SSCs with associated closed ITAAC or program elements, and such SSCs may also need repairs.  The inspection program will confirm, on a sampling basis, that the surveillance and post-maintenance testing performed in this interim period are focused not only on technical specification operability and similar operational concerns, but also on maintaining the validity of ITAAC determinations.

The licensee is responsible to identify when an ITAAC completion determination basis is in question and an IPCN to the NRC is required.  The licensee will do that based on five maintenance thresholds identified in NEI 08-01.  If one of the maintenance thresholds is exceeded, the licensee will submit an IPCN following implementation of corrective actions to ensure the acceptance criteria of the affected ITAAC continue to be met or are met again.  The NRC will review the IPCNs that identify those corrective actions to determine if the ITAAC conclusion for those ITAAC is maintained or not.  If that review determines that the ITAAC acceptance criteria are no longer met, the ITAAC will be reopened.  The licensee will have to take further corrective actions in order to reclose the ITAAC and then submit a new IPCN.  The ITAAC Maintenance Process inspection program and required notifications are described in detail in NEI 08-01, Inspection Procedure 40600, “Licensee Program for Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) Management”, and in Regulatory Guide 1.215.

07.19	10 CFR 52.103(g) Finding.  The NRC will retain records of the final review for an ITAAC to determine if it can be verified as complete and will also maintain records of all ICNs and IPCNs.  These records will be retained in ADAMS for a potential hearing.  That potential hearing on ITAAC will require prima facie evidence that the acceptance criteria of specific ITAAC are not met.  Office Instruction LIC-114, “Title 10 of the *Code of Federal Regulations*(10 CFR) Section 52.103(g) Finding and Communication Process” contains guidance for the 10 CFR 52.103(g) finding process. 

07.20	ITAAC Hearings and Interim Operations.Prior to making 10 CFR 52.103(g) finding that the acceptance criteria for all ITAAC are met, Section 189-a(1)(B)(i) of the AEA provides that the NRC shall publish in the *Federal Register* a notice of intended operation.  This notice shall provide that any person whose interest may be affected by operation of the plant, may within 60 days request the Commission to hold a hearing on whether the facility as constructed complies, or on completion will comply, with the acceptance criteria of the license.

To be granted, a request for hearing must show, prima facie, that one or more of the acceptance criteria in the COL have not been, or will not be met, and the specific operational consequences of nonconformance that would be contrary to providing reasonable assurance of adequate protection of the public health and safety.  If the Commission determines that there will be reasonable assurance of adequate protection of the public health and safety during a period of interim operation while ITAAC hearings are pending, then the staff may make the 10 CFR 52.103(g) finding that the acceptance criteria in the ITAAC are met, which would allow 

Issue Date:  11/25/20	18	2506

interim operation of the reactor pending the outcome of the hearing.  Interim operation was intended for situations in which the petitioner’s prima facie showing relates to alleged safety consequences that will not arise during the interim operation allowed, or in which mitigation measures can be taken to preclude potential safety consequences during interim operation.  For example, interim operations could be allowed where, although a petitioner has raised a question about the long-term safety of the plant and the NRC has decided a hearing on the issue is warranted, the NRC is able to determine that the plant is safe to operate during an interim period.

The staff submitted SECY-13-0033, “Allowing Interim Operation under Title 10 of the *Code of Federal Regulations* Section 52.103,” to the Commission on April 4, 2013 (ADAMS Accession No. ML12289A928), and the Commission issued the associated staff requirements memorandum (SRM) on July 19, 2013 (ADAMS Accession No. ML13200A115).  This SECY paper informed the Commission of issues associated with interim operation in the event that an ITAAC hearing is pending.  In the SRM, the Commission approved the staff’s recommendation that the Commission delegate the 10 CFR 52.103(g) finding to the staff, and the Commission directed the staff, the Office of the General Counsel, and the Office of Commission Appellate Adjudication to develop options for the ITAAC hearing formats for Commission review and approval.  An inter-agency working group developed the ITAAC hearing procedures and submitted SECY-15-0010, “Final Procedures for Hearings on Conformance with the Acceptance Criteria in Combined Licenses,” dated January 20, 2015 (ADAMS Accession No. ML14343A704).  The SECY provides the draft final procedures for ITAAC hearings for Commission approval.  In accordance with SRM-SECY-15-0010, the staff finalized the ITAAC hearing procedures and then published NRO Office Instruction NRO-REG-105, “NRC Staff Support of the Inspections, Tests, Analyses, and Acceptance Criteria Hearing Process,” (ADAMS Accession No. ML17131A137). 

If inspections are required after the 10 CFR 52.103(g) finding for ITAAC that are subject to a hearing during interim operations, they will be conducted using applicable IPs specified by IMC 2503.  Findings identified during these inspections will be dispositioned using the ROP guidance in IMC 0612, “Issue Screening” and IMC 0609, “Significance Determination Process,” and will be assigned to the ROP cornerstone that is most closely related to the finding for consideration in the ROP Action Matrix.  Note that inspection findings from an inspection of this type, although related to a subject ITAAC, would not be characterized as ITAAC Findings since ITAAC Findings are only assigned as part of the cROP. 

07.21 	Construction Assessment Program. The NRC conducts assessment of a licensee’s effectiveness in assuring construction quality in accordance with IMC 2505.  In implementing the construction assessment program, the NRC staff integrates information relevant to licensee safety performance, makes objective conclusions regarding the significance of inspection findings, takes actions based on these conclusions in a predictable manner, and effectively communicates these results to the licensees and to the public.

The construction assessment program consists of the following key principles:

a.	Inspection results will be the input to the assessment program.

b.	Inspection results will have established thresholds.

c.	Crossing thresholds will result in the NRC considering a range of actions as defined in the CAM.

The NRC determines the significance of inspection results in accordance with the construction SDP described in IMC 2519.  The construction SDP is a risk informed approach to evaluating the significance of construction inspection program findings.  The significance of inspection findings, as characterized by the SDP, is represented by a color scheme (i.e., green, white, yellow, red).  The staff uses the color of construction inspection findings as the input to the construction assessment program’s CAM.  The NRC also evaluates each finding to determine if the primary cause of the finding can be associated with one of the cross-cutting aspects.  During the assessment of licensee performance, the NRC determines if a CCI exists per the guidance in IMC 2505.

The NRC developed a review system that provides for continuous, quarterly, and end-of-cycle (annual) reviews of licensee performance data (inspection results).  The system is designed so that the continuous and quarterly reviews are informal reviews of performance data and are not resource intensive.  The end-of-cycle reviews are more formal and include licensee performance review meetings and an assessment report.  An agency action review is generally reserved for plants requiring consideration of agency-wide actions as determined during the Agency Action Review Meeting.

The NRC communicates assessment results through quarterly updates of assessment data and annual assessment reports.  The staff holds a public meeting with the licensee near the licensee’s facility after the conclusion of the annual assessment cycle.  The NRC makes annual assessment letters publicly available prior to the public meetings and the annual Commission meeting.

07.22 	Construction Enforcement Program.  The NRC Enforcement Policy governs the processes and procedures for the initiation and review of violations of NRC requirements.  The NRC Enforcement Manual contains implementation guidance.  Both documents are owned and issued by the Office of Enforcement (OE).  In addition, for new reactors licensed under 10 CFR Part 52, IMCs 0613, 2505, and 2519 provide guidance for assigning significance to findings and the NRC’s response to findings associated with new reactors under construction.

07.23 	NRC Allegation Program.  Management Directive (MD) 8.8, “Management of Allegations,” describes the NRC’s allegations program.  Each region’s Enforcement and Investigations Coordination Staff (EICS) coordinates the processing of allegations that it has received or has been assigned.  Each region has developed and issued office instructions and procedures to implement the requirements of MD 8.8.  The Allegations COE coordinates the processing of allegations received by or assigned to Headquarters, which provides a centralized location for administering an effective program to manage allegations.  The COE’s charter (ADAMS Accession No. ML12045A102) provides details on the purpose, organization, and responsibilities of the Allegations COE.

07.24 	Construction Experience Program (ConE).  The ConE program supplements and supports the agency’s operating experience (OpE) program described in MD 8.7, “Reactor Operating Experience Program” and IMC 2523, “NRC Application of Operating Experience in the Reactor Oversight Process.”  The ConE program is led by the ConE COE in NRR.  Office Instruction NRR-LIC-401, “NRR Reactor Operating Experience Program,” documents the ConE process.

As described in NRR-LIC-401, the ConE program collects, screens, and evaluates lessons learned from nuclear construction and operating experience for application into the NRC’s new reactor licensing and inspection programs.  The ConE program communicates design and construction lessons learned to NRC staff, and when necessary, to external stakeholders through generic communications.  Region II Regional Office Instruction (ROI) No. 0608, “Handling of Operating Experience in Region II,” provides regional guidance for using OpE in inspection planning and communicating potentially generic safety questions and construction deficiencies to cognizant headquarters personnel.

07.25	Annual cROP Self-Assessment.In the SRM in response to 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 provide the Commission with an annual self-assessment report of the implementation of the construction inspection program.  In response, the staff has developed and conducts an annual cROP self-assessment in accordance with IMC 2522, “Construction Reactor Oversight Process Self-Assessment Program.”  

The cROP self-assessment process uses program evaluations and performance metrics to determine its success in meeting the goals and intended outcomes of the cROP.  The level of effectiveness of the cROP is determined by considering whether the program meets its goals and achieves its intended outcomes.  The cROP is intended to successfully:

a.	Monitor and assess licensee performance.  

b.	Identify performance issues through NRC inspection.  

c.	Determine the significance of identified performance issues.  

d.	Adjust resources to focus on significant performance issues.  

e.	Evaluate the adequacy of corrective actions for performance issues.  

f.	Take necessary regulatory actions for significant performance issues.  

g.	Communicate inspection and assessment results to stakeholders.  

h.	Make program improvements based on stakeholder feedback and lessons learned.  

Periodically, the cROP self-assessment program collects information from various sources, including CIPMS, the inspection program, periodic independent audits, stakeholder surveys, public comments, and other stakeholder interactions.  The staff reports the results of the annual self-assessment to the Commission via a SECY paper in support of the Agency Action Review Meeting.  

07.26	Transition from cROP to ROP.  10 CFR 52.103(g) does not allow license holders to operate a new reactor facility until the NRC finds that all the acceptance criteria in the combined license are met.  The appendices to 10 CFR Part 52 further define facility operation as beginning at fuel load (Appendix D, IX.B.2 for the AP1000).  Per 10 CFR 52.103(h), ITAAC are no longer requirements after the NRC has found the acceptance criteria to be met.

Because 10 CFR 52.103(h) removes ITAAC as regulatory requirements after all acceptance criteria are met, the technical specifications and all other applicable regulatory requirements from 10 CFR Parts 50 and 52, including license conditions, govern the operation of the facility.  This becomes the basis for the transition to the ROP.  Once the Commission finds that all acceptance criteria in the license have been met, the CIP will end and inspections under the ROP will begin.  At that time, the lead inspection responsibility will switch from Region II to the host region.

Implementation of the ROP for newly constructed facilities may involve changes from that used on current plants due to the lack of historical data for most performance indicators and the lower risk profile for the new plants.  The NRC will conduct inspections under the guidance of IMC 2514, “AP1000 Reactor Inspection Program - Startup Testing Phase,” and IMC 2515.  The staff will disposition findings identified during these inspections under the provisions of IMC 0612, “Issue Screening,” and IMC 0609, “Significance Determination Process,” and will document the findings using IMC 0611, “Power Reactor Inspection Reports.”  Assessment of the facility will transition from the construction assessment program described in IMC 2505 to the operating reactor assessment program described in IMC 0305.

Operational programs will be inspected during the life of the facility under IMC 2504, IMC 2514, or IMC 2515.

2506-08	REFERENCES	

IMC 0102, “Oversight and Objectivity of Inspectors and Examiners at Reactor Facilities”

IMC 0305, “Operating Reactor Assessment Program”

IMC 0609, “Significance Determination Process”

IMC 0611, “Power Reactor Inspection Reports”

IMC 0612, “Issue Screening”

IMC 0613, “Power Reactor Construction Inspection Reports”

IMC 0617, “Vendor and Quality Assurance Implementation Inspection Reports”

IMC 0620, “Inspection Documents and Records”

IMC 2501, "Construction Inspection Program:  Early Site Permit"

IMC 2502, “Construction Inspection Program:  Pre-Combined License (Pre-COL) Phase”

IMC 2503, “Construction Inspection Program:  Inspections of Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) Related Work”

IMC 2504, “Construction Inspection Program:  Inspections of Construction and Operational Programs”

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

IMC 2507, “Vendor Inspections” 

IMC 2514, “AP1000 Reactor Inspection Program - Startup Testing Phase”

IMC 2515, “Light-Water Reactor Inspection Program – Operations Phase”

IMC 2519, “Construction Significance Determination Process”

IMC 2522, “Construction Reactor Oversight Process Self-Assessment Program”

IMC 2523, “NRC Application of Operating Experience in the Reactor Oversight Process”  

NEI 08-01, Inspection Procedure 40600, “Licensee Program for Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) Management”

Management Directive 8.7, “Reactor Operating Experience Program”

Management Directive 8.8, “Management of Allegations”

Office Instruction LIC-114 , “Title 10 of the *Code of Federal Regulations*(10 CFR) Section 52.103(g) Finding and Communication Process”

Office Instruction NRO-REG-103, “Inspections, Tests, Analyses, and Acceptance Criteria Closure Verification Process”

Office Instruction NRO-REG-105, “NRC Staff Support of the Inspections, Tests, Analyses, and Acceptance Criteria Hearing Process”

SECY-07-0047, “Staff Approach to Verifying the Closure of Inspections, Tests, Analyses, And Acceptance Criteria Through a Sample-Based Inspection Program”

SECY-13-0033, “Allowing Interim Operation under Title 10 of the Code of Federal Regulations Section 52.103”

SECY-15-0010, “Final Procedures for Hearings on Conformance with the Acceptance Criteria in Combined Licenses”

END

Exhibits:

1. Construction Reactor Oversight Process Overview

2. Construction Regulatory Oversight Framework

Appendices:

A. Construction Inspection Program Guidance

B. Construction Inspection and Assessment Program Bases

Attachments:

1. Acronyms

2. Revision History for IMC 2506

Issue Date:  11/25/20	18	2506

Exhibit 1 - Construction Reactor Oversight Process Flowchart

Issue Date:  11/25/20	E1-1	2506

Exhibit 2 - Construction Regulatory Oversight Framework

Issue Date:  11/25/20	E2-1	2506

Appendix A - Construction Inspection Program Guidance

2506A-01		PURPOSE

This appendix provides detailed guidance for the construction inspection program (CIP).

2506A-02		BACKGROUND

NRC inspectors should conduct inspections in accordance with IPs.  Procedures and guidance cannot address all the unique circumstances that an inspector might encounter.  Therefore, inspectors should exercise initiative based on their expertise, experience, and risk insights, to meet all the inspection objectives.

2506A-03		DISCUSSION

2506A.03.01	Inspector Policy

2506A03.01.01	Construction Resident Inspector (CRI) Policy

The CRIs provide the major onsite NRC presence for direct observation and verification of licensees’ ongoing activities.  CRIs shall be qualified under IMC-1245, “Qualification Program for New and Operating Reactor Program.”  CRIs are responsible for being aware of major activities and the status of construction activities.  The CRIs are also the primary NRC onsite evaluators for events or incidents.  CRIs will perform the greater part of initial event-related inspection, and may be augmented by other inspectors, depending on the type and significance of the event.  Regional managers will decide when normal inspection activities and staffing will resume.

2506A03.01.02	Regional and Vendor Inspector Policy

Inspectors conduct inspections as directed by their supervisors and shall be qualified under IMC 1245.  In addition to baseline inspection program procedures, inspectors often will conduct inspections under other program elements, such as allegation follow-up, etc.  Certain aspects of their inspection activities may be conducted in the office (e.g., portions of procedure review and administrative program inspection).

2506A03.01.03	ITAAC Inspections

Most ITAAC inspections will take place onsite.  Exceptions to this policy are described in IMC 2506, Section 07.06.  If during the inspection planning process, the licensee requests that an ITAAC inspection be conducted at a licensee’s agent’s facility (e.g., at a vendor facility), then NRC inspectors should conduct the inspection at the agent’s facility, if practical, and provided the requirements in Section 07.06 are met.  In this case, the licensee will ensure that all documentation that supports process, design and development activities, testing, etc. is available at the agent’s facility, and that the appropriate personnel will support the inspection.

Issue Date:  xx/xx/13	AppA-7	2506

Issue Date:  11/25/20	App A-6	2506

2506A03.01.04	Inspection Coordination

The senior CRI and the Region II Division of Construction Oversight must be kept advised of regional and headquarters inspectors’ activities at the facility.  The associated regional branch chief must ensure coordination of regional and headquarters inspection activities using the guidance for visits to operating sites provided in IMC 0301, "Coordination of NRC Visits to Commercial Reactor Sites."

Regional and headquarters-based inspectors should contact the senior CRI or the senior project inspector before each inspection to get information concerning the availability of specific licensee personnel, the status of construction activities that may affect the planned inspection, and the status of allegations at the facility.  In addition, they should contact the senior CRI prior to coming to the site and as soon as is convenient after they arrive at the site to ensure a coordinated NRC presence at the facility.  The senior CRI should inform the regional and headquarters inspectors of any unique activities in progress and offer specific inspection suggestions.  The regional and headquarters inspectors should brief the senior CRI about the results of their inspection before the exit meeting with the licensee’s management.  The senior CRI, or CRI if the senior is unavailable, should attend all exit meetings where significant issues are expected to be discussed.

2506A03.01.05	Third Party Assistance

Refer to IMC 2515, “Light-Water Reactor Inspection Program Operations Phase,” Section 11-04 for guidance regarding third party assistance requests.

2506A03.02		General Inspection Policies

2506A03.02.01	Management Entrance and Exit Meetings

The inspection entrance and exit meetings are the primary opportunities to communicate issues with licensees.  Refer to IMC 2515, Section 12-01 for guidance regarding management entrance and exit meetings.

2506A03.02.02	Findings Outside of Inspector’s Qualifications

Refer to IMC 2515, Section 12-04 for guidance regarding findings outside of inspector’s qualifications.

2506A03.02.03	Event Response

Licensees often notify inspectors of events or conditions in anticipation of the inspectors’ interest in an issue.  Such notifications do not exempt the licensee from reporting events and conditions through the required regulatory processes (e.g., 10 CFR Parts 21, 50.55(e), 50.72, or 50.73).  Inspectors should inform the licensee that documents given to inspectors are subject to Freedom of Information Act requests and may be placed in the Public Document Room.

Refer to IMC 2504, Appendix C for guidance regarding the decision-making process for Regional and Headquarters staff to use in planning an appropriate response to potentially significant, non-performance related, issues or events at reactor construction sites.

2506A03.02.04	Communication with Local Public Officials

Refer to IMC 2515, Section 12-05 for guidance regarding communications with local public officials.

2506A03.02.05	Witnessing Unsafe Situations

Refer to IMC 2515, Section 12-06 for guidance regarding witnessing unsafe situations.

2506A03.02.06	Memoranda of Understanding with the Occupational Safety and Health Administration

In general, OSHA has jurisdiction over plant conditions that result in an occupational risk, but do not affect the safety of licensed radioactive materials.  For example, in a construction environment, there might be exposure to toxic non-radioactive materials and other industrial hazards.  IMC 1007, “Interfacing Activities between Regional Offices of NRC and OSHA,” contains specific guidance to be used to implement the Memorandum of Understanding between OSHA and the NRC.

2506A03.02.07	Inspector Functions during Period of Lapsed Appropriation.

NRC Management Directive 4.5, “Contingency Plan for Periods of Lapsed Appropriations,” has defined the resident and selected region-based inspector functions as an excepted NRC activity that will continue during the period of restricted NRC operations.  

Resident inspectors will continue with their respective functions, which includes the following activities:

· Completion of all of the activities in this IMC that are normally assigned to a resident inspector, including completion of all baseline activities that have been assigned to a resident inspector for the site.

· Completion of reactive inspection activities pursuant to NRC Management Directive 8.3, “NRC Incident Investigation Program,” and IMC 2504, Appendix C, “Response to NonPerformance-Related Issues or Events.”  The decision to initiate a reactive inspection shall be made in consultation with the “excepted function” Regional and Program Office managers.  If applicable, the requirement for Senior Risk Analyst participation in establishing the risk significance of an event that meets the deterministic criteria is waived.  Residents and “excepted function” Regional managers are granted the discretion to use available tools, including SDP screening tools and the licensee’s risk tools, as they may be available and appropriate.

· Completion of baseline, reactive, and supplemental inspection activities not covered above that are approved by regional management as being within the technical expertise of the residents at the site and that have been scheduled for completion during the period of lapsed appropriation.

· Emergency response, incident response, allegation, enforcement, public communication, and support for emergency licensing action activities that are typically performed by resident inspectors.

Region-based inspectors will continue with their respective functions, which includes the following activities:

· Initial operator licensing activity.  

· Operator requalification licensing inspection.  

· Event response which would require regional specialist expertise.  

The resident and region-based inspection functions do not include Construction Reactor Oversight Process activities or other program activities in the applicable IMCs shown below that require substantial support or approval from the Regional Office or a Program Office:

· IMC 2505, “Periodic Assessment of Construction Inspection Program Results”

· IMC 2519, “Construction Significance Determination Process”

· IMC 0613, “Power Reactor Construction Inspection Reports”

2506A03.03		Construction Resident Inspector Program

The CRI program requires the selectees to be qualified under IMC 1245, “Qualification Program for New and Operating Reactor Program.”  Region II management will make the selection of CRIs.  Staffing levels at the construction resident offices will depend on many factors but will largely be based on the amount and type of ITAAC activities occurring on-site.  

Construction sites are likely to be co-located with an existing operating reactor site that will have its own resident inspection staff.  The activities at the construction sites must not detract from the safety oversight responsibilities the NRC has toward the nearby operating facilities.  The inspection programs for construction and operating sites are significantly different from each other, and the training and qualifications for CRIs are different than for operating reactor resident inspectors.  Thus, the NRC has committed to keeping the CIP separate from the operational inspection program.

There may be a need for both the construction and operational resident inspectors to be knowledgeable about issues that can affect both areas.  Allegations, environmental issues, security and emergency response programs, etc., are examples of potentially common issues.  Generally, the construction resident inspector will not be expected to provide backup site coverage for the operations resident inspectors.  The construction resident inspectors will not normally be expected to respond to a plant event and will not be designated as a back-up responder for the operating reactor.  However, the construction residents, as well as any other qualified NRC inspectors, could be directed to provide coverage for a site event if they are on site and no operating resident inspectors are available, at least until the operating resident inspection staff can arrive on site.

Site coverage requirements and back shift inspections by resident and regional inspectors during construction will be determined by Region II management.

All CRIs will stipulate a 7-year maximum tour length.  This policy does not preclude CRIs from relocating for promotions, voluntary reassignments, or management-directed reassignments.

CRIs are expected to relocate site assignment after 7 years.  CRIs due to rotate during the winter months or early spring may be granted an extension to the summer months with Regional Administrator approval.  CRIs may be extended to no later than 1 year beyond completion of start-up testing of the last unit completed at a construction site with Regional Administrator approval.  Any extensions beyond one year after start-up testing of the last unit completed at the site must be approved by the Deputy Executive Director for Reactor and Preparedness Programs (DEDR). 

As CRIs approach the completion of construction, the agency will consider inspector requests for a lateral transfer or reassignment to an Operational Resident Inspector (ORI).  Earlier transfers can be made when consistent with agency needs.  In either case, CRIs are encouraged to make their desires and career goals known to their management as far in advance as possible.

CRIs should not normally be reassigned to the same facility, after having been an ORI or CRI, even after an intervening assignment.  Reassignments may be made to co-located facilities that would cause CRIs to interact with different licensee management (e.g., reassignment from an operating unit to a unit under construction at the same facility).  In this case, the seven-year site time would be reset.  This policy applies to total site tour length and it is not affected by a promotion from resident inspector to senior resident inspector at an operating or construction site.

CRIs should not be assigned to a different location within the first 4 years after relocating unless specifically approved by the DEDR or based on identified agency needs.  

This policy applies to the Resident and Senior Resident Inspectors assigned at any of the reactor sites (i.e., construction or operating), fuel facilities, and gaseous diffusion plants. 

2506A03.04	Inspection Program Modifications in the Event of A Pandemic

In the event of a pandemic, the NRC’s Pandemic Response Plan (PRP) requires that the NRC maintain aspects of the inspection program, identified as priority functions.  Additionally, the NRC’s PRP allows modifications to less critical aspects of the inspection program in order to address limited inspection resources.

Therefore, the regional administrator may authorize postponing “supplemental” and “generic safety” inspections.  The NRC may reduce baseline inspection activities, commensurate with available inspection and licensee resources.  Event response inspections will continue.  If necessary, the NRC will reduce the baseline inspection program such that inspectors will monitor key construction activities, if available, or by remote means.  The NRC will resume normal inspection activities once the pandemic has passed.  The NRC will make reasonable efforts to complete missed baseline inspection activities in a reasonable timeframe.

2506A03.05		Licensee Strike Contingency Plans

In the event that a strike occurs at a reactor construction site, inspectors should implement applicable portions of IP 92709, “Contingency Plans for Licensee Strikes or Lockouts,” IP 92711 “Implementation of Licensee Contingency Plans During a Strike/Lockout,” and IP 92712, “Resumption of Normal Operations after a Strike.”

Appendix B - Construction Inspection and Assessment Program Bases

2506B-01	PURPOSE

This appendix provides bases used in the development of the construction inspection and assessment programs.

2506B-02	BACKGROUND

The staff has interacted with stakeholders and the Commission in developing the construction licensing, inspection and assessment programs.  This appendix captures the bases for the development of the oversight and assessment programs of reactors under construction.

2506B-03	DISCUSSION

B03.01 	Organizational Structure

The current fleet of operating reactors was constructed pursuant to regulations contained in 10 CFR Part 50.  The Office of Nuclear Reactor Regulation (NRR) is responsible for the oversight of reactor construction activities under 10 CFR Part 50.  Similarly, NRR had responsibility for oversight of construction activities under 10 CFR Part 52, which was first issued in 1989.  The industry expressed renewed interest in reactor construction in the late 1990’s and early 2000’s.  As workload increased and to prepare for and manage future reactor and site licensing applications, the Nuclear Regulatory Commission (NRC) established the Future Licensing Organization as a temporary organization in NRR in March 2001.  In July 2001, the NRC permanently established the organization as the New Reactor Licensing Project Office.

On August 12, 2005, in SECY-05-0146, the staff proposed a reorganization of NRR to be in the best organizational (programmatic and technical) position to review new reactor license applications.  In this proposal, which was approved by the Commission on August 25, 2005, the NRC created the Division of New Reactor Licensing to place greater organizational emphasis in this area.

On February 26, 2006, in SECY-06-0041, the staff proposed strategies to support implementation of the new reactor construction inspection program.  On April 21, 2006, the Commission approved the formation of a dedicated organization for new reactor construction inspection in the Region II Office in Atlanta, Georgia.  The Commission stated that this organization would have total responsibility for all construction inspection activities across the country, including both the day-to-day onsite inspections and the specialized inspection resources needed to support NRC oversight of the construction of any new nuclear power plants.  This approach was intended to ensure consistency in implementing the new inspection program and quickly incorporate ongoing lessons learned into the entire program.

On July 21, 2006, the Commission approved the staff’s recommendation, as described in SECY-06-0144, to reorganize the Office of Nuclear Reactor Regulation into two offices:  the Office of New Reactors (NRO) with responsibility and authority for new reactor licensing and the Office of Nuclear Reactor Regulation (NRR) with responsibility for operating reactor licensing.  The Commission also approved the staff’s recommendation to create a Deputy Regional Administrator for Construction in Region II.  On April 16, 2012, the NRC implemented centers of expertise (COE) within NRO and NRR in the areas of allegations, operating experience and construction experience (OpE/ConE), electrical engineering, and vendor inspection.

The Allegations COE is led by the Office of Enforcement (OE) with a dedicated liaison supporting the CIP to ensure allegations associated with Nuclear Regulatory Commission (NRC) regulated activities are processed in accordance with agency established policies and procedures.  The Office Allegation Coordinator (OAC) resides in OE, and coordinates allegation activities for NRR, and the Office of Nuclear Security and Incident Response (NSIR).  Executive responsibility for management and oversight of the Allegations COE is held by the OE Director.  The Allegations COE has the responsibility to coordinate with the NRO technical staff to ensure that they are appropriately involved in initial screening, follow-up, ARBs, and closure of new reactor construction allegations. 

The ConE COE is led by, and resides in, NRR.  The ConE COE focuses on knowledge sharing and coordination to systematically collect, screen, evaluate, and communicate domestic and international reactor operating and construction experience, and to apply lessons learned.

The Vendor Inspection COE supports the Allegations COE and conducts inspections to verify the effective implementation of vendor quality assurance programs in order to assure the quality of materials, equipment, and services supplied to the commercial nuclear industry.  The Vendor Inspection COE also leads efforts to address and deter the potential use of counterfeit, fraudulent, and suspect items in safety-related applications.  

2506B03.02	Construction Licensing and Inspection Programs

In the aftermath of the accident at Three Mile Island (TMI) in March 1979, the NRC suspended the granting of operating licenses for plants that were in the pipeline.  The licensing pause for fuel loading and low-power testing ended in February 1980.  In August 1980 the NRC issued the first full-power operating license since TMI, to North Anna, Unit 2, in Virginia.  In the following 9 years it granted full-power licenses to over forty other reactors, most of which had received construction permits in the mid-1970s.

The lengthy and laborious licensing procedures that applicants had to undergo in the cases of Shoreham, Seabrook, and other reactors stirred new interest in simplifying and streamlining the regulatory process.  Specifically, obtaining an operating license after construction was complete (two-step process) increased the risk and complexity of the licensing process.  This risk and complexity were a major deterrent to utilities who considered building nuclear plants.  

The NRC proposed to simplify the traditional two-step licensing process with a one-step process.  After much deliberation, the Commissioners, staff, and nuclear vendors, converged on the one-step licensing process (10 CFR Part 52) that was authorized in 1989.

The NRC issued NUREG-1055, “Improving Quality and the Assurance of Quality in the Design and Construction of Nuclear Power Plants:  A Report to Congress,” in May 1984.  The NUREG 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 TMI, 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 NRC intended the SALP assessment 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).  This project had two purposes:  1) to address programmatic weaknesses that had been identified during the inspection and licensing of plants in the 1980s, and; 2) to develop an inspection program for evolutionary and advanced reactors.  The NRC suspended the project in late 1994, because of the lack of nuclear power plant construction activities.  In October 1996, the NRC issued “Draft report on the Revised Construction Inspection Program,” which presented a framework from which the CIP could be reactivated to support NRC inspections at future nuclear power plants.  This framework included recommendations for continuous NRC onsite inspection presence that matches inspector expertise to inspection needs, an inspection procedure format that clearly defines the attributes and associated acceptance criteria that must be inspected, and a dedicated CIPIMS to be used to implement the CIP in concert with the inspection manual.

Late in 2000, the industry renewed interest in constructing new nuclear power plants.  On February 13, 2001, the Commission issued a staff requirements memorandum (SRM) for COMJSM-00-0003, directing the staff to assess its technical, licensing, and inspection capabilities and identify enhancements, if any, that would be necessary to ensure that the agency can effectively carry out its responsibilities associated with an early site permit (ESP) application, a license application, and the construction of a new nuclear power plant.

The staff first responded to this SRM in a memorandum dated May 1, 2001.  This memo outlined several organizational changes, including the temporary establishment of the Future Licensing Organization in NRR, which was responsible for coordinating the preparations for the review of new applications (i.e., ESPs, design certifications, and combined licenses).  This memo also informed the Commission that NRR would reactivate the CIP revision effort suspended in 1994, and that this effort would include review and revisions of applicable IMCs and development of the associated inspection guidance and training for inspection of critical attributes of construction processes and activities.

On October 12, 2001, the staff further responded to COMJSM-00-0003 by submitting SECY-01-0188, “Future Licensing and Inspection Readiness Assessment.”  This SECY paper included the “Future Licensing and Inspection Readiness Assessment Report,” summarizing the efforts of an interoffice working group.  This report included resource estimates for revising IMCs 2511, 2512, 2513, and 2514; indicated that the NRR Inspection Program Branch (IPB) would lead CIP revisions; and discussed the formation of the New Reactor Licensing Project Office in NRR.  IPB formed the CIP team, composed of representatives from each region, new reactor licensing staff, and inspection program management, and tasked it with updating the inspection and assessment program for use in inspecting reactors to be licensed and 

Issue Date:  11/25/20	App B-1	2506

Issue Date:  11/25/20	App B-6	2506

constructed under 10 CFR Part 52.  NUREG 1789, “10 CFR Part 52 Construction Inspection Program Framework Documents,” which was issued in April 2004, describes the work of this team.

The CIP developed by this team has four phases.  The first and second phases support a licensing decision for an ESP and the COL application.  Inspections will initially be performed to confirm the accuracy of data submitted to the NRC in support of safety evaluations for an ESP and COL.  The third and fourth phases support construction activities and the preparations for operation.  Prior to and during plant construction, the NRC will conduct off-site inspections to review vendor activities and licensee oversight of these activities.  During plant construction, on-site inspections will focus on verifying satisfactory completion of ITAAC, as specified in the FSAR, and also on inspecting programs for operational readiness and transition to power operations.

2506B03.03		ITAAC and Operational Programs History

The history of ITAAC is coupled with the history of nuclear power plant standardization, particularly with the standardization of the processes for issuing combined construction permits and conditional operating licenses.  Early in the commercial nuclear power industry, there were many first-time nuclear plant applicants, designers, and consultants, and many novel design concepts.  Accordingly, the process was structured to allow the NRC to make licensing decisions while design work was still in progress and to focus reviews on individual plant-specific and site-specific considerations.  The NRC issued construction permits with the understanding that open safety issues would be addressed and resolved during construction and that issuance of a construction permit did not constitute Commission approval of any design feature.  Consequently, the operating license review was very broad in scope.

The premise of 10 CFR Part 52 Subpart C is that a mature nuclear industry can describe and evaluate plant designs on a generic basis, and to have designs essentially complete in scope and level of detail prior to construction.  This makes it possible to combine the construction permit with much of the operating license.  This concept was incorporated into 10 CFR 52.97(b)(1), which states that the Commission shall identify within the combined license the inspections, tests, and analyses, including those applicable to emergency planning, that the licensee shall perform, and the acceptance criteria that, if met, are necessary and sufficient to provide reasonable assurance that the facility has been constructed and will be operated in conformity with the license, the provisions of the Atomic Energy Act, and the Commission's rules and regulations.  The combined license can then authorize full-power operation following an opportunity for a hearing on a more limited set of issues related to whether acceptance criteria for an ITAAC have not or will not be met.

10 CFR Part 52 did not specifically address whether COLs should contain programmatic ITAAC.  In the early 1990s, several SECY papers addressed programmatic ITAAC issues, as did several letters from industry.  The staff addressed this issue with the Commission in SECY-00-0092, “Combined License Review Process,” dated April 20, 2000.  The SECY paper discussed requiring programmatic ITAAC in COLs.  In the SRM for this SECY, the Commission directed the staff to interact with stakeholders on the need for and scope of programmatic ITAAC and formally provide the Commission with a recommendation as to how to proceed on programmatic ITAAC.

Issue Date:  11/25/20	App B-1	2506

Subsequently, the staff submitted SECY-02-0067, “Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) for Operational Programs (Programmatic ITAAC).”  The staff requested the Commission’s approval to require that COL applications contain programmatic ITAAC for a wide range of operational programs such as training, quality assurance, fitness for duty, and others.  In the SRM for this paper, the Commission disapproved the staff’s proposal.

On February 26, 2004, the staff submitted SECY-04-0032, “Programmatic Information Needed for Approval of a Combined License without ITAAC,” which requested the Commission’s approval of a staff proposal regarding the level of programmatic information needed for approval of a COL without ITAAC for any particular program.  Specifically, the staff recommended that the Commission approve the categorization of operational programs into five different categories (A-E) and, that procedure-level information be provided or available to the NRC to support review of a COL application.  The staff further stated that if such information cannot be provided or made available during the COL application review, ITAAC would be necessary for that program.

In the SRM associated with SECY-04-0032, the Commission approved the categorization of operational programs into five categories but disapproved the staff’s recommendation concerning the need for procedure-level information to support review of a COL application.  The Commission further stated that the staff should continue the practice of inspecting relevant licensee procedures and programs in a similar manner as was done in the past and consistent with applicable inspection programs.  The Commission also stated that the staff should continue to ensure, consistent with the inspection and enforcement processes, that licensees address pertinent issues prior to fuel loading.

The Commission directed the staff describe what information would be necessary for the COL application for each of the programs for which the staff had previously assumed ITAACs would be required (i.e., fire protection, training, quality assurance during operation, fitness for duty, access authorization, radiation protection, physical security, licensed operator, and reportability programs) by December 31, 2005.

On October 28, 2005, the staff submitted SECY-05-0197, “Review of Operational Programs in a Combined License Application and Generic Emergency Planning Inspections, Tests, Analyses, and Acceptance Criteria,” which requested Commission approval of a staff proposal to include license conditions for operational programs in a COL.  The staff concluded that a COL applicant could fully describe all operational programs and their implementation in the COL application, with the exception of emergency preparedness (EP).  The staff noted that if these programs and their implementation are fully described, they would not require ITAAC.  The staff would inspect operational programs and their implementation as they are developed and put into place.  These inspections would verify that the program being implemented is consistent with the FSAR.  In addition, these inspections would verify that any changes made to the programs as described have not adversely impacted the bases for the Commission's findings of reasonable assurance.  Any adverse impacts discovered during inspection would be subject to enforcement action.  In the SRM associated with SECY-05-0197, the Commission approved the use of license conditions for operational program development and implementation.

2506B03.04 	ITAAC Inspection Philosophy

Complete coverage and direct inspection of the activities associated with the entire population of ITAACs contained in an approved design would be an inefficient and unnecessary use of 

Issue Date:  11/25/20	App B-1	2506

dedicated NRC inspection resources.  The ITAAC inspection philosophy contained in IMC 2503 recognizes that several ITAAC are expected to be closely related, thereby providing the NRC with the opportunity to evaluate a group of ITAAC (an ITAAC family) based upon an inspection of some representative ITAAC within the family.  To facilitate the inspection of representative ITAAC within a family to confirm adequate licensee control and completion of the ITAAC, the NRC developed the ITAAC Matrix, a high-level inspection planning tool.  Such an inspection approach allows for the efficient use of NRC inspection resources for ITAAC inspections and for the routine evaluation of the construction processes that result in the ITAAC products and completion.

2506B03.05		ITAAC Matrix Structure

The ITAAC Matrix identifies 25 core categories.  These categories cover the licensee’s construction programs and processes for the quality construction of a nuclear power plant.  The matrix contains six columns (i.e., the programmatic activities) and 19 rows (i.e., the process activities).  It covers the technical disciplines and programmatic controls for the fabrication and installation of structures, systems, and components (SSCs).  The inspections developed from the matrix confirm that the completed facility will perform as designed, and that the licensee has implemented the required program elements.

The matrix facilitates identification of common ITAAC families and provides a foundation for an efficient inspection sampling approach.  The matrix bins ITAAC that are connected by their common characteristics within an ITAAC family.  For example, all ITAAC for I&C would be “binned” at the intersection of row (10) and column (A).  The matrix structure facilitates the inspection of the sample of ITAAC and ensures adequate coverage of all construction disciplines, whether directed to a specific category (e.g., [03] Piping), or to a process (e.g., [B] Welding).

2506B03.06		ITAAC Matrix Contents

A panel of NRC experts populated the matrix for each certified plant design by grouping the ITAAC into families.  They selected the row (programmatic functions) and column (process attributes) that best covered the construction activities involved with each ITAAC for the relevant plant design.  The staff can populate a matrix once ITAAC are codified for a certified design.  The staff would review plant specific ITAAC for placement within the proper matrix families.  The NRC plans to update a matrix following certified revisions to designs.  The process is summarized as follows:

(a)	An NRC expert panel reviews all the ITAAC for each certified design and for each custom design.  An expert panel generally consists of three NRC personnel with expertise in plant construction, reactor risk, and project licensing, including relevant plant design and ITAAC experience or knowledge.

(b)	The expert panel convenes to populate the matrix, reviews each of the ITAAC, and places it in one of the blocks of the ITAAC Matrix.

(c)	Once the expert panel determines where in the matrix each of the ITAAC for a particular design should be placed, all facilities constructed with that particular design will use that specific, populated ITAAC Matrix.

Issue Date:  11/25/20	App B-1	2506

This use of a single ITAAC Matrix format provides a consistent framework for developing the inspection programs for each of the different advanced reactor designs that are licensed and built under 10 CFR Part 52.  Additionally, this also ensures a degree of consistency in the inspection program within any specific design.

2506B03.07		What the Matrix Provides

The Matrix is a mechanism for using the guidance and knowledge base learned from the existing NRC inspection program successfully used for 10 CFR Part 50 operating plant inspections.  The Matrix incorporates this knowledge base into a related 10 CFR Part 52 ITAAC inspection framework which provides:

(a) A high-level NRC inspection planning tool for identifying related groups (i.e., “families”) of ITAAC, based upon common characteristics.

(b) A logical, convenient basis to facilitate ITAAC sampling.

(c) A consistent model for the selection of targeted ITAAC at plants of similar design.

(d) A methodology that establishes a documented process for the NRC completion of ITAAC inspections.

(e) A framework for the determination of how many ITAAC require direct inspection through the ITAAC targeting process.

(f) Use of related program and process inspections to assess the quality of plant construction, with necessary focus on the ITAAC.  This matrix set of core IPs, supplemented by some complementary supporting procedures, is a significantly smaller number of IPs than were used as part of IMC 2512 for the NRC inspection of the existing operating plants licensed under 10 CFR Part 50.

2506B03.08		Matrix Implementation

The matrix row procedures focus on inspection of quality processes for specific construction disciplines for the installation of SSCs, as well as program elements that licensees are required to implement.  The matrix column procedures address inspection criteria that cover multiple disciplines and represent ITAAC characteristics that may be common to several of the row processes.  Under this inspection philosophy, an NRC inspector conducts an inspection of an installation process and the resulting SSC, as defined by one of the matrix row IPs.  Portions of other row and column procedures may be used as appropriate for an individual ITAAC.

The inspector can focus on the available ITAAC in that matrix row.  Similarly, when an inspector reviews the program attributes from the matrix columns, the inspector can focus on the available ITAAC in that matrix column per that column IPs.  As the inspector evaluates the quality process and programmatic criteria identified by an intersection of a matrix row and column, that inspection is focused on the ITAAC within a specific family.  This allows inspection findings and conclusions to be applied to the other ITAAC in that same family, which may not have been inspected.

Issue Date:  11/25/20	App B-1	2506

2506B03.09		ITAAC Matrix Summary

The NRC developed the ITAAC Matrix to group ITAAC into families that cover all the construction programs and processes for quality construction.  The grouping of ITAAC families provides the logical connectivity between the programmatic and process inspections necessary for efficient inspection of the entire range of ITAAC.  The matrix row and column IPs guide NRC inspections that verify the quality of the construction programs and processes, and the resultant SSC quality.

2506B03.10		ITAAC Ranking and Targeting Process

The CIP cannot inspect all licensee construction activities associated with each ITAAC.  Therefore, the NRC contracted with Information Systems Laboratories, Inc. (ISL) to develop an ITAAC prioritization methodology (ADAMS Accession No. ML060740006).  The concept was to develop a selection process that could work with the ITAAC Matrix to rank the ITAAC of any particular design.  This rank would be based upon the value that NRC inspection provides to the assurance that the completed ITAAC could be accepted without need for additional confirmation.

The ITAAC prioritization methodology objective is to optimize NRC inspection resources, while providing reasonable assurance that a significant flaw in the completion of the ITAAC by the licensee will not go undetected.  The staff described the prioritization process in SECY-07-0047, “Staff Approach to Verifying the Closure of Inspections, Tests, Analyses, and Acceptance Criteria Through a Sample-Based Inspection Program.”  On May 16, 2007, in the SRM for SECY-07-0047, the Commission approved the staff’s approach for selecting ITAAC to be given priority for inspection.  The ITAAC prioritization process is documented in NRO Office Instruction OIREG-102, “Prioritization of Inspections, Tests, Analysis, and Acceptance Criteria (ITAAC) for Inspection.”

The prioritization process is based on the grouping into ITAAC families.  Once grouped into a family, the ITAAC may then be prioritized within the family.  The overall approach is that observing licensee performance of the activity on one ITAAC provides insights on licensee performance regarding other ITAACs.

The first step in prioritization involves rank-ordering the ITAAC based upon certain defined attributes that make one ITAAC more or less important to inspect than the others.  Attributes are considered to be some of the representative characteristics of any particular ITAAC.  The following five attributes were selected for ranking consideration:

a. Complexity or Difficulty of Activity.  The likelihood of errors during fabrication, installation, or testing.  As an example, a bimetallic weld on the reactor vessel safe end might be more difficult than a weld on structural steel for a seismic pipe support.  The degree of training or certification required of the “doer” such as a Level III NDE technician is an indicator of the complexity.  This typically is also related to the concept of a special process which has requirements associated with it per Appendix B to 10 CFR Part 50.

b.	Construction and Testing/Training Experience.  Whether the testing or construction activity is a “first of a kind” for construction or a new test conducted by a group with little experience.  Experience may mean limited work in the nuclear field, in a field with 

Issue Date:  11/25/20	App B-1	2506

	quality assurance requirements, or in strict adherence to procedural controls.  Additionally, this includes whether there is a history of quality or other performance deficiencies associated with the company or the activity. 

c.	Difficulty of Verifying by Other Means.  The degree that the activity can be verified by observing other functional, pre-operational tests, or performance tests.  This would also include the degree to which the sequence is a factor.  For example, the lack of access associated with buried piping or cables, coatings inside tanks, or physical interferences.  This would result in a preference to inspect while the opportunity exists, or to defer the inspection until when it may be useful to witness the pre-operational test instead.

d.	Safety Significance.  The safety significance assigned to the system, component, or structure included in the ITAAC.  This attribute will be defined by a risk weighting factor which will be assigned separate to expert panel evaluation of the other attributes.

e.	Licensee or Applicant Oversight Attention.  The amount and effectiveness of the applicant’s or licensee’s oversight attention and quality assurance efforts, including those of its contractors and suppliers.  This also includes those self-assessment reviews or independent audits in addition to the specific QA effort.  Note this may not be known early in the sequence of construction activities or until NRC has experience inspecting the licensee’s QA efforts and other self-assessment activities and generated an opinion of their performance.  

The attributes are weighted according to their impact on the overall objective.  Then, expert panels rate each ITAAC for each attribute.

In November 2005, an expert panel of NRC managers with extensive nuclear construction and NRC inspection experience convened to weigh each of the five ITAAC attributes.  The expert panel then chose utility values for the level of inspection related to each attribute.  This attribute weighting and utility selection process is part of the Analytic Hierarchy Process, which was chosen by ISL as part of the ITAAC prioritization process.  The expert panel provided the results as input to the ISL algorithm ISL to establish the basis for the subsequent evaluation of the ITAAC against each of the five attributes.  This panel performed the weighting and utility process for the five pre-selected attributes that apply to any reactor design.  Therefore, the NRC will not have to repeat these expert panel deliberations.

The rating given to each ITAAC correlates to the assurance that inspection of an ITAAC provides.  As such, the value of inspecting an ITAAC is prioritized, versus the ITAAC itself.  Therefore, the objective aims to optimize resources to ensure that no significant construction flaw is undetected.

The NRC used this process to target ITAAC for inspection for the AP1000 and ABWR.  These ITAAC are referred to as targeted ITAAC.  The numerical data for each reactor design will be different and therefore the numerical cut off value will also be different.  The selected value will be selected to provide reasonable coverage of all ITAAC for the planned NRC inspection activities for direct NRC inspection.  The second step used in the methodology required that at least one ITAAC from every family be inspected.  This requirement has been deleted.  The need to target ITAAC below the established threshold, to provide adequate inspection coverage of the construction processes and programs, is evaluated as part of the targeting process.

Issue Date:  11/25/20	App B-1	2506

2506B03.11		Site-Specific ITAAC

In addition to the ITAAC listed in the design certification rules for approved reactor designs, each COL application contains site-specific ITAAC that consist of systems that are outside the scope of the standard design.  In SECY-08-0117, “Staff Approach to Verify Closure of ITAAC and to Implement Title 10 CFR 52.99, “Inspection During Construction,” and Related Portion of 10 CFR 52.103(g) on the Commission Finding,” the staff indicated that it would review and inspect work related to the site-specific ITAAC using a method similar to the prioritization methodology described in SECY-07-0047.  Headquarters staff leads the prioritization of site-specific ITAAC in the COLs and the DCDs for future reactor designs.  Expert panels select the site-specific ITAAC samples based on safety significance and the ability to inspect.

COL applications also contain ITAAC for EP and physical security.  The staff inspects selected work activities related to physical security and EP ITAAC.  In addition to the ITAAC-related work inspections, the staff performs comprehensive inspections of the operational programs for security and EP.  This includes force-on-force security inspections and NRC observations of EP exercises.

2506B03.12		Design Acceptance Criteria (DAC) ITAAC

DAC, which are a subset of the ITAAC for a given design, are another ITAAC inspection area.  DAC are design details that were not provided at the time of DCD submittal.  These design details are made available during construction and verified as part of the ITAAC to demonstrate that the system design and as-built configuration conformed to the licensing basis.  DAC are designated in three specific disciplines as outlined in SECY 92-053:  1) Digital I&C design; 2) Piping design, and; 3) Human Factors engineering.  Additionally, the ABWR design includes limited Radiation Protection DAC.

DAC associated with an ITAAC are inspected as the design detail is made available by a COL applicant or licensee.  DAC inspections will normally be led by Region II with support from NRR technical staff, because of their complexity.  Since DAC inspection are required to satisfy the associated ITAAC, all DAC inspections must be completed prior to the 10 CFR 52.103(g) finding.

B2506.03.13	Process for the Modification of the ITAAC Target Set

The prioritization process was meant to optimize resources while providing reasonable assurance that a significant flaw by the licensee would not go undetected.  The process should be adaptive and use lessons learned or inspection history to modify the prioritization.  The instructions to modify the population of “targeted” ITAAC is provided in NRO-REG-102, Prioritization of Inspections, Tests, Analysis, and Acceptance Criteria (ITAAC) for Inspection.”

2506B03.14		Construction and Operational Program Inspections

While the ITAAC will be the focus when selecting which activities to inspect, the NRC staff will inspect more than just ITAAC-related work.  Licensees are required to develop and implement construction programs that are listed in IMC 2504.  Construction program inspections that are conducted pursuant to IMC 2504 include inspection of the licensee’s ITAAC management program and QA activities affecting SSCs that are installed in the plant.  Therefore, ITAAC will be indirectly evaluated by programmatic inspections because such programs affect the quality 

Issue Date:  11/25/20	App B-1	2506

of the SSCs that are the subject of the ITAAC.  Additional construction program inspections address reporting of defects and noncompliances, fitness for duty, and the preoperational testing program.  The staff's verification that the licensee has properly developed and implemented construction programs is consistent with the NRC’s use of sampling during inspections.  Construction activities inspected by NRC are assumed to be representative of similar activities that the NRC did not directly inspect.

As the project progresses, the NRC will inspect operational programs.  The scope and content of the operational programs are reviewed during the COL review process and approved when the COL is issued.  10 CFR 52.79(a) requires that each COL application include an FSAR that provides information concerning facility design, construction, and operation.  The NRC will plan inspections of operational programs based on the following license condition per SECY-05-0197: 

“A license condition that specifies that the licensee shall make available to the NRC staff a schedule 12 months after issuance of a COL that supports planning for and conduct of NRC inspections of operational programs listed in the operational program FSAR table.  The schedule shall be updated every 6 months until 12 months before scheduled fuel load, and every month thereafter until either the operational programs listed in the FSAR table have been fully implemented or the plant has been placed in commercial service, whichever comes first.”

The licensee must implement approved operational programs prior to the milestones listed in the COL.  The staff intends to inform the Commission of the status of operational programs at the time of the 10 CFR 52.103(g) decision.

2506B.03.15	Construction Assessment Program

NRO developed a construction assessment program through interactions with its stakeholders.  IMC 2505, “Periodic Assessment of Construction Inspection Program Results,” initially issued on October 20, 2008, contains details regarding implementation of the construction assessment.  The initial version of IMC 2505 included a CAM, which provided guidance for NRC response to degraded licensee performance.  The significance of findings was determined using a traditional enforcement approach.  The staff provided a description of the construction assessment 

program to the Commission in 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.

On December 5, 2008, the Commission issued SRM M081022, which directed the staff to reconsider the construction assessment process as presented in IMC 2505 and propose policy options to the Commission.  The Commission directed that the staff 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.

The staff issued IMC 2505, Revision 1, on December 24, 2009, to provide assessment program guidance to be implemented for construction activities ongoing while the Commission made a final determination of how the assessment program should be implemented.  This revision retained much of the guidance from the initial issuance of IMC 2505, and added a safety culture approach which is similar to the approach taken in the ROP.

Issue Date:  11/25/20	App B-1	2506

In response to SRM M081022, NRO, other program offices, and the regional offices formed an interoffice working group to develop construction assessment program options for Commission consideration.  The working group interacted extensively with external stakeholders in the development of construction assessment program options for Commission consideration.  On October 26, 2010, the staff submitted SECY 2010-0140, “Options for Revising the Construction Reactor Oversight Process Assessment Program.”  In SECY 10-0140, the staff recommended that the Commission approve the development of a construction assessment program that includes a regulatory framework, the use of a construction SDP to determine the significance of findings identified during the CIP, and the use of a CAM to determine the appropriate NRC response to findings.  The Commission approved the staff’s recommendation in the SRM in response to SECY-10-0140, dated March 21, 2011.

The staff developed a new cROP that consists of many of the same objective elements as those used in the ROP, starting with a construction regulatory framework and including a construction SDP, a construction action matrix, and a similar enforcement approach to that which is in use in the ROP.  Beginning on January 1, 2012, the staff conducted a 12-month pilot program for the new cROP in accordance with the guidance in memorandum, “Pilot Program for the Construction Reactor Oversight Process Assessment and Enforcement Programs,” dated January 5, 2012 (ADAMS Accession No. ML113120210).  The pilot was conducted at Southern Nuclear Operating Company’s Vogtle, Units 3 and 4, and South Carolina Electric and Gas Company’s Virgil C. Summer, Units 2 and 3.  The staff reported the results of the pilot and planned program changes to the Commission in SECY Paper 13-0042, “Construction Reactor Oversight Process Self-Assessment for Calendar Year 2012,” and fully implemented the new programs on July 1, 2013.

Issue Date:  11/25/20	App B-1	2506

Attachment 1 - Acronyms

ABWR - Advanced Boiling Water Reactor

ADAMS - Agency wide Documents Access and Management System

AP1000 - Advanced Passive 1000

ARB - Allegation Review Board

AV - Apparent Violation

CAM - Construction Action Matrix 

CCI - Cross-Cutting Issue

CGI - Commercial Grade Items

CIP - Construction Inspection Program

CIPB - Construction Inspection Program Branch

CIPIMS - Construction Inspection Program Information Management System 

COL - Combined License 

ConE - Construction Experience

CRIs - Construction Resident Inspectors

cROP - Construction Reactor Oversight Process

DAC - Design Acceptance Criteria

DC - Design Certification

DCD - Design Control Document

DCO - Division of Construction Oversight

DCRA - Design-Centered Review Approach

DEDR - Deputy Executive Director for Reactor and Preparedness Programs

DNRL - Division of New Reactor Licensing

EDV - Engineering Design Verification

EPR - Evolutionary Power Reactor

ESBWR - Economic Simplified Boiling Water Reactor 

ESP - Early Site Permit

FEMA - Federal Emergency Management Agency

FIN - Finding

FRN - *Federal Register Notice*

FSAR - Final Safety Analysis Report

HQ - Headquarters

ICN - ITAAC Closure Notification

IMC - Inspection Manual Chapter

IP - Inspection Procedure

ITAAC - Inspections, Tests, Analyses and Acceptance Criteria

ITP - Initial Test Program

LAR - License Amendment Request

LWA - Limited Work Authorization

MOU - Memorandum of Understanding

NCV - Non-Cited Violation

NOV - Notice of Violation

NRC - Nuclear Regulatory Commission

NRO - Office of New Reactors

NRR - Office of Nuclear Reactor Regulation

OE - Office of Enforcement

OpE - Operating Experience

OSHA - Occupational Safety and Health Administration

PAR - Preliminary Amendment Request

Issue Date:  11/25/20	Att1-1		2506

Issue Date:  xx/xx/xx	Att1-2	2506

PM - Project Manager

PRP - Pandemic Response Plan

QA - Quality Assurance

QC - Quality Control

R-COL - Reference Combined License

ROP - Reactor Oversight Process

S-COL - Subsequent Combined License

SCWE - Safety Conscious Work Environment

SDP - Significance Determination Process

SSCs - Structures, Systems, and Components

URI - Unresolved Item

VIO - Violation

VOICES - Verification of ITAAC Closure, Evaluation and Status

VPO - Vogtle Project Office

Issue Date:  11/25/20	Att1-1		2506

Attachment 2 – Revision History for IMC 2506

Issue Date:  	Att2-1	2506

Issue Date:  11/25/20	Att2-2	2506

# image1.emf

# Microsoft_PowerPoint_Slide.sldx

# image2.emf

# Microsoft_PowerPoint_Slide1.sldx