**NRC INSPECTION MANUAL**	IRIB

INSPECTION MANUAL CHAPTER 0308 ATTACHMENT 2

# TECHNICAL BASIS FOR INSPECTION PROGRAM

Effective Date: 01/01/2024

Contents0308.2-01	INTRODUCTION	10308.2-02	METHODOLOGY FOR IDENTIFYING INSPECTABLE AREAS	10308.2-03	BASELINE INSPECTION PROGRAM	103.01	PI Verification	403.02	Problem Identification and Resolution	403.03	Event Followup	503.04	 Plant Status	60308.2-04	SUPPLEMENTAL INSPECTION PROGRAM	70308.2-05	EVENT RESPONSE	80308.2-06	OVERSIGHT OF PLANTS IN EXTENDED SHUTDOWN	80308.2-07	THRESHOLD FOR DOCUMENTING FINDINGS AND INSIGHTS	90308.2-08	OTHER INSPECTION PROGRAM ASPECTS CONSIDERED BUT NOT INCLUDED	10BASIS SUMMARY SHEETS	11Exhibit 1: Adverse Weather Protection (IP 71111.01)	12Exhibit 2: Reserved (IP 71111.02)	14Exhibit 3: Reserved (IP 71111.03)	15Exhibit 4: Equipment Alignment (IP 71111.04)	16Exhibit 5: Fire Protection (IP 71111.05)	17Exhibit 6: Flood Protection Measures (IP 71111.06)	19Exhibit 7: Heat Exchanger/Sink Performance (IP 71111.07)	20Exhibit 8: Inservice Inspection Activities (IP 71111.08)	22Exhibit 9: Reserved (IP 71111.09)	24Exhibit 10: Reserved (IP 71111.10)	25Exhibit 11: Licensed Operator Requalification Program and Licensed Operator Performance(IP 71111.11)	26Exhibit 12: Maintenance Effectiveness (IP 71111.12)	28Exhibit 13: Maintenance Risk Assessments and Emergent Work Control (IP 71111.13)	30Exhibit 14: Reserved (IP 71111.14)	32Exhibit 15: Operability Determinations and Functionality Assessments (IP 71111.15)	33Exhibit 16: Reserved (IP 71111.16)	35Exhibit 17: Evaluations of Changes, Tests, and Experiments (IP 71111.17T)	36Exhibit 18: Plant Modifications (IP 71111.18)	37Exhibit 19: Post-Maintenance Testing (IP 71111.19)	38Exhibit 20: Refueling and Other Outage Activities (IP 71111.20)	39Exhibit 21: Safety System Design and Performance Capability (IP 71111.21)	40Exhibit 22: Surveillance Testing (IP 71111.22)	41Exhibit 23: Reserved (IP 71111.23)	43Exhibit 24: Testing and Maintenance of Equipment Important to Risk (IP 71111.24)	44Exhibit 25: Biennial Exercise and Drill Inspection (IP 71114.01, .06, .07 and .08)	46Exhibit 26: Alert and Notification System Evaluation (IP 71114.02)	48Exhibit 27: Emergency Response Organization Staffing and Augmentation System (IP 71114.03)	49Exhibit 28: Emergency Action Level and Emergency Plan Changes (IP 71114.04)	50Exhibit 29: Correction of Emergency Preparedness Weaknesses and Deficiencies (IP 71114.05)	52Exhibit 30: Access Control to Radiologically Significant Areas (IPs 71124.01, 71124.04, and 71124.08)	54Exhibit 31: ALARA Planning and Controls  (IPs 71124.01, and 71124.03)	55Exhibit 32: Radiation Monitoring Instrumentation (IPs 71124.04 and IP 71124.05)	57Exhibit 33: Radioactive Gaseous and Liquid Effluent Treatment and Monitoring Systems (IPs 71124.05, 71124.06, and 71124.07)	58Exhibit 34: Radiological Environmental Monitoring Program (REMP) (IP 71124.07)	60Exhibit 35: Radioactive Material Processing and Radioactive Material Handling, Storage, and Transportation (IP 71124.08)	61Exhibit 36: Discrepant or Unreported Performance Indicator Data (IP 71150)	62Exhibit 37: Performance Indicator Verification (IP 71151)	63Exhibit 38: Problem Identification and Resolution (IP 71152)	64Exhibit 39: Follow up of Events and Notices of Enforcement Discretion (IP 71153)	66Exhibit 40: Light Water Reactor Inspection Program - Operations Phase (IMC 2515)	67Exhibit 41: Risk Informed Baseline Inspection Program (IMC 2515, Appendix A)	69Exhibit 42: Supplemental Inspection Program (IMC 2515, Appendix B)	70Exhibit 43: Special and Infrequently Performed Inspections (IMC 2515, Appendix C)	71Exhibit 44: Plant Status (IMC 2515, Appendix D)	72Exhibit 45: Inspection Program Modifications During Public Health Emergencies Or Other Conditions Restricting Inspector Onsite Presence (IMC 2515, Appendix E)	73Exhibit 46: Supplemental Inspection Response to Action Matrix Column 2 (Regulatory Response) Inputs (IP 95001)	74Exhibit 47: Supplemental Inspection Response to Action Matrix Column 3 (Degraded Performance) Inputs (IP 95002)	75Exhibit 48: Supplemental Inspection Response to Action Matrix Column 4 (Multiple/Repetitive Degraded Cornerstone) Inputs (IP 95003)	77Exhibit 49: Special Inspection (IP 93812)	78Exhibit 50: Oversight of Operating Reactor Facilities in a Shutdown Condition Due to Significant Performance and/or Operational Concerns (IMC 0350)	79Table 1: Inspectable Areas by Cornerstone	81Table 2: Other Inspection Program Elements Considered But Not Included	83Attachment 1: Revision History for IMC 0308 Attachment 2	Att1-1

Issue Date: 10/31/23	1	0308 Att 2

# 0308.2-01	INTRODUCTION

The power reactor inspection program is composed of several elements to provide indication of licensee performance. The key feature of the program is the baseline inspection program, which defines the minimum level of inspection that all plants will receive regardless of performance. The supplemental inspection program is performed to independently evaluate the root causes of performance deficiencies when indications of declining licensee performance are obtained through either the performance indicators (PIs) or other inspections (principally the baseline inspection program). Plant events are inspected to determine their significance and to determine the agency’s necessary response. Plants in extended shutdowns due to performance problems are inspected and assessed by a separate inspection process (i.e., Inspection Manual Chapter [IMC] 0350) because many of the PIs and much of the baseline inspection program would not be applicable.

The riskinformed baseline inspection program for power reactors defines the minimum level of planned inspections to evaluate licensee performance over a 12month period. The overall objective of the program is to monitor all power reactor licensees at a defined level of effort to assure licensees’ performance meets the objectives for each cornerstone of safety. These cornerstones support the agency’s performance goals in the NRC’s Strategic Plan.

# 0308.2-02	METHODOLOGY FOR IDENTIFYING INSPECTABLE AREAS

The objective in revising the inspection program was to develop a baseline program that is riskinformed and performance-based that identifies the minimum level of inspection required for a plant (regardless of performance) to give the NRC sufficient information to determine whether plant performance is acceptable. A key input to this effort was the regulatory framework and the cornerstones of safety, which are areas of reactor functions or licensee activities that must be performed to a certain set of objectives to ensure that the NRC’s mission is met.

The baseline inspection program was developed using a riskinformed approach to determine a comprehensive list of areas to inspect (inspectable areas) within each cornerstone of safety. These inspectable areas were selected based on their risk significance (i.e., they are needed to meet a cornerstone objective as derived from a combination of probabilistic risk analyses insights, operational experience, deterministic analyses insights, and regulatory requirements). The scope of inspection within each inspectable area was determined using the same riskinformed approach. The scope of inspection was also modified by the applicability of a PI. The more fully an indicator measures an area, the less extensive is the scope of inspection.

# 0308.2-03	BASELINE INSPECTION PROGRAM

The baseline inspection program contains certain concepts that are a change in the approach to conducting inspections from the previous core inspection program. The key concepts are summarized below:

The baseline program is the minimum level of inspection conducted at all power reactor facilities, regardless of their performance. Licensees performing at a level not requiring additional NRC interaction will only be inspected at the baseline inspection level of effort.

Inspections of performance issues beyond the baseline program are termed supplemental inspections. This increased inspection effort is based on criteria specified in the assessment program to address declining licensee performance and is not included in the baseline program.

The scope of the baseline program is defined by inspectable areas linked to the cornerstones of safety. The justification for inclusion of the inspectable area in the baseline program is described in this basis document.

The baseline program has four parts: (1) inspection in inspectable areas in which PIs are not identified and/or in which PIs do not fully cover the inspectable area; (2) ongoing verification of the information provided in PIs; (3) comprehensive review of licensee effectiveness in identifying and resolving problems, and (4) initial follow up to plant events and degraded conditions to determine their safety significance.

The process for planning inspections will be conducted in accordance with IMC 0305, “Operating Reactor Assessment Program.”

Risk has been factored into the baseline inspection program in four ways: (1) inspectable areas are based on their risk importance in measuring a cornerstone objective, (2) the inspection frequency, how many activities to inspect, and how much time to spend inspecting activities in each inspectable area is based on risk information, (3) the selection of activities to inspect in each inspectable area is based on plantspecific risk information, and (4) inspectors are trained in the use of risk information.

A panel consisting of Inspection Program Branch and senior regional managers and their staff developed the sample size and the number of inspection hours expected to be necessary to complete each of the inspection procedures, at the inception of the reactor oversight program (ROP). Sample size and number of hours were developed based on their expert judgement and relevant risk information on how much inspection activities would be sufficient to ensure verification that the licensee was meeting the objectives of all seven cornerstones.

After the first year of implementation of the new ROP, regional management and inspectors raised concerns regarding the lack of flexibility in the ROP inspection requirements for both sample size requirements and number of hours for each inspectable area. They were concerned with their ability to apply their inspection focus into areas they felt needed more or less inspection effort based on their overall knowledge of a specific plant. As a result, in consultation with regional management, the Inspection Program Branch changed the original sample size from a single value to a range of values which were -15 percent to +15 percent of the original sample size. The original sample size is the nominal or average of the -15 percent and +15 percent values.

The idea was that any individual plant inspection program could then be adjusted within these relatively limited ranges based on the plant-specific insights of the inspectors, but that at a nationwide program level, the average (i.e., mean) level of samples and effort would continue to fall in about the middle of these ranges. As experience with the ROP was accumulated, it was felt that these program average values and ranges could then be adjusted as needed while still retaining an appropriate degree of flexibility to accommodate plant-specific inspection focus needs.

Appendix A to IMC 2515 contains a list of baseline inspection procedures and specifies the required frequency for their performance. The baseline inspection procedures must be completed at every plant at a prescribed interval. In certain cases, completion of some inspection requirements may be accomplished through other inspections. The expectation is that the regions should normally complete the nominal (average) number of inspection samples identified in the inspection procedure. The regions may vary the inspection samples within the ranges as indicated in each baseline inspection procedure, based on the licensee performance and inspector insights. For the purposes of completing the baseline inspection program, the number of samples completed must be within the range of values specified in each inspection procedure.

Similar changes were made to the inspection hours in order to maintain the relationship between the level of inspection resources necessary to complete the inspection activities and the range of inspection samples which could be accomplished with each inspection procedure constant.

The program is indicative and not diagnostic. The baseline program delineates specific inspection activities to evaluate aspects of licensee programs and processes and their implementation by identifying findings that are indicative of licensee performance problems. Inspection findings from the baseline program are evaluated for significance and used, along with PIs, to assess licensee performance within the cornerstones of safety. The baseline inspections are not diagnostic assessments of licensee performance leading to a root cause determination. Those assessments and root cause determinations are intended to be reviewed or independently made during supplemental inspections that are outside the scope of the baseline inspection program.

The safety performance of nuclear power plants is assessed based on performance in each cornerstone of safety. Verifying that a licensee meets the objectives of the cornerstones provides reasonable assurance that public health and safety are protected. The inspectable areas verify aspects of the key attributes for each of the associated cornerstones. The cornerstones to which each inspectable area is applicable and their link to the attributes they are measuring are depicted in table 1 of this attachment and exhibits 3 through 11 of IMC 0308. Therefore, the baseline inspection program requires that most inspectable areas be reviewed at each nuclear power plant each year. Several inspectable areas are reviewed at longer frequencies.

All the important aspects of a cornerstone area are inspected where a PI has not been established (e.g., design). In cornerstone areas where the PIs provide only limited indication of performance, the inspectable areas provide indication of the aspects not measured (e.g., operator performance during an event). If performance of the cornerstone objective in a cornerstone area is sufficiently measured by a PI, the inspection effort in the baseline program only verifies that the PI is providing the intended data.

Exhibits 1 through 34 describe the scope of each inspectable area and explain the basis for why each inspectable area is included in the baseline program. Reasons for inclusion in the program may be that: (1) the area is linked to the NRC’s mission, (2) the inspectable area involves a key attribute to a cornerstone of safety, and (3) risk information justifies including the area in the baseline inspection program. These inspectable area basis summary sheets discuss the basis for each inspectable area and include risk insights (from generic risk analyses and studies), analyses of significant precursor events, and the riskinformed judgment of an expert panel of inspectors and risk analysts. The summary sheet for each inspectable area also identifies whether a PI applies to the area and what inspections may be needed in addition to the information provided by the PIs in the area. The baseline inspection procedures are written to focus on the more risksignificant aspects of the inspectable areas as discussed in the summary sheets, aspects that directly support the desired results and promote the important attributes of the cornerstones of safety. The scope of any associated PIs is summarized in the inspectable area portions of the baseline inspection procedures.

The exhibits related to the physical protection inspection procedures were removed because of the Commission’s decision that certain security information will no longer be publicly available.

In addition to the inspectable areas identified for many of the key attributes of each cornerstone of safety, the baseline inspection program also consists of inspection activities devoted to: (1) PI verification, (2) problem identification and resolution, (3) event followup, and (4) plant status. As discussed below, Exhibit 36 through 44 describe the scope and basis for these inspection activities and other inspection program policies and practices (e.g., IMC 2515).

## 03.01	PI Verification

The monitoring of plant performance primarily relies on information provided by PIs and inspection findings in areas not measured, or not adequately measured, by PIs. The baseline inspection program will also selectively collect and review licensee plantspecific raw data on a periodic basis to independently verify the accuracy and completeness of the PI data.

Each PI is verified annually. The annual verification compares the reported PI data to samples of raw data available (e.g., operating logs, corrective action program records, maintenance records). Some PIs can be verified in conjunction with other baseline inspections if the PI is difficult to accurately verify from plant records. The PI verification inspection also reviews corrective action program records to determine if any problems the licensee may have had in collecting PI data were adequately resolved and updates provided to the NRC.

If a PI is found to be invalid based on inaccurate or incomplete data, then the associated cornerstone may not be adequately evaluated, and additional inspections within the areas measured by the PI are scheduled. The baseline inspection program also provides guidance for NRC actions in response to incomplete or unreported PIs.

Exhibits 36 and 37 describe the scope and basis for: (1) the NRC response to discrepant or unreported PI data, and (2) PI verification inspection activities, respectively.

## 03.02	Problem Identification and Resolution

Inspection of licensee problem identification and resolution (PI&R) programs has been a key component of the ROP since the inception of the program. The PI&R Comprehensive Review Attachment 4, “Problem Identification and Resolution Basis and History,” (ML20247J599) is an in-depth history of the procedure and its basis.

One of the primary means by which licensees maintain an appropriate level of safety is through an effective PI&R program to correct deficiencies involving human performance, equipment, programs, and procedures. The NRC’s confidence in the effectiveness of these programs is the basis for the NRC’s policy of closing lowerlevel violations when they are entered into the licensee’s corrective action program without independently verifying the final corrective actions, which in turn is a basis for implementation of the ROP. Section 2.3.2 of the Enforcement Policy describes the NRC policy for dispositioning most violations associated with green findings as NCVs for licensees that have “implemented a corrective action program that is determined to be adequate by the NRC.” That policy states that “[t]he NRC will credit a formal corrective action program that has been inspected and found to meet regulatory guidance, industry standards, or both.” The inspection program verifies that our confidence in licensees’ programs is still deserved and periodically verifies the final actions on some of the lower-level violations are proper.

The process for evaluating PI&R consists of a performancebased review of the licensees’ deficiency reporting process, selfassessments, quality assurance audits, causal analyses of events, and corrective actions. The review of corrective actions includes following them up to validate their effective implementation. The NRC reviews the licensee’s activities in this area to verify that: (1) the scope of licensees’ identification and resolution programs bounds the key attributes in the cornerstones; (2) causes of problems and issues have been properly determined and corrective actions are timely and effective; and (3) the generic implication or extent of condition has been appropriately considered. Issues identified regarding the licensee’s implementation of its corrective action program are assessed for risk significance using the Significance Determination Process (SDP).

The NRC program to review activities in this area has four parts. The first part is conducted during inspection of the associated inspectable areas within each cornerstone. The second part is a semiannual trend review. The third part is a sample of issues that are selected annually for more indepth review. The fourth part is a biennial review of the licensee’s PI&R programs. The biennial review complements the reviews done throughout the year. The results of the biennial review are then integrated with the PI&R insights gained via the other inspections.

NRC inspectors use licensees’ selfassessments to help direct these baseline inspections into worthwhile areas. However, licensees’ selfassessments will not be used to reduce or replace baseline inspections. Exhibit 38 provides additional information on the scope and basis for PI&R inspections.

A determination that the NRC no longer has confidence in the licensee’s PI&R program implies it is incapable of participation in the ROP because section 2.3.2.a of the Enforcement Policy no longer applies. This determination should be coincident with entrance into Column 5 of the Action Matrix or the IMC 0350 Process, an Action Matrix deviation, or other appropriate oversight mechanisms. IMCs 0305 and 0350 provide governance for these processes. A determination of that level can only be reached by regional management in consultation with the NRR Officer Director and may require approval by the Executive Director for Operations.

## 03.03	Event Followup

The NRC normally follows up plant events in three ways: (1) events of low safety significance receive minimal follow up, usually by the resident inspectors, (2) events of moderate safety significance receive more follow up, often by one or two regional inspectors, and (3) events of greater safety significance are followed up by a special team. The baseline program is designed to initially screen all operational events and licensee event reports and to follow up only some of the more routine, noncomplex events. The baseline program includes a procedure for event followup to be used in conjunction with inspections in the various inspectable areas. Whether to follow up other events with regional discretionary resources would depend on the significance of the event as determined by the baseline inspection program.

Events of low safety significance, such as uncomplicated reactor trips, are reviewed by resident or regionbased inspectors to verify that the events are not complicated by conditions such as loss of mitigation equipment or operator errors. The baseline inspection program’s event followup procedure focuses the inspector’s initial evaluation of events on communicating details regarding the event to risk analysts for their use in determining risk significance. Inspectors will identify equipment malfunctions and unavailability, operator errors, and other complications.

The followup of more extensive, nonroutine events is outside of the scope of the baseline inspection program and would be performed with reactive inspection resources. The decision to follow up such events would be made on a casebycase basis by NRC regional management and as directed by senior NRC management in accordance with NRC Management Directive (MD) 8.3, "NRC Incident Investigation Program." Significant operational events (defined in MD 8.3) are followed up by a graded response consisting of inspections such as those conducted by Incident Investigation Teams (IITs) and Augmented Inspection Teams (AITs), and Special Inspections (SIs). Followup of these events is discussed in more detail later in this attachment.

Exhibit 39 provides additional information on the scope and basis of event followup activities performed as part of the baseline inspection program.

## 03.04	****Plant Status

The primary objective of the plant status activities is to ensure that the inspectors are aware of current plant conditions and equipment problems and have a level of understanding of the risk significance of proposed or ongoing operations, maintenance, and testing by the licensee. Plant status focuses on identifying and understanding emergent plant issues, current equipment problems, and ongoing activities and their overall impact on plant risk. These activities also provide an independent assessment of the licensee’s effectiveness in entering program, system, and component deficiencies into the corrective action program.

The plant status portion of the inspection program is important because it will be used by the inspectors in the riskinformed process to select inspection samples and to modify the scope and depth of inspections in other inspectable areas that support assessment of all cornerstone areas. This awareness of plant conditions, emerging problems or work, and activities planned by the licensee is used by the inspectors in determining which inspection procedures to use and the specific samples for inspections within the inspectable areas of the baseline inspection program. Therefore, since plant status is conducted in part to prepare for other baseline inspection activities, this effort is not considered part of the direct inspection effort under the baseline program.

Exhibit 44 provides additional detail regarding the scope and basis of the plant status activities performed under the baseline inspection program.

# 0308.2-04	SUPPLEMENTAL INSPECTION PROGRAM

The supplemental element of the inspection program was designed to apply NRC inspection resources in a graded manner when risk significant performance issues are identified, either by inspection findings evaluated using the SDP or when PI thresholds are exceeded. Depending on the risk significance and breadth of the identified performance issues, the supplemental inspections provide a range of activities including: oversight of the licensee’s root cause evaluation of the issues; expansion of the baseline inspection sample or a focused team inspection (as necessary to evaluate extent of condition); or a broad scope multidisciplined team inspection which would include inspection of multiple cornerstone areas and inspection of crosscutting issues. Any new performance issues identified during the supplemental inspections are evaluated by the SDP, and new findings issued. The need for additional NRC actions, including additional supplemental inspections, are governed by the assessment program Action Matrix.

At the lowest level, the intent of supplemental IP 95001, “Supplemental Inspection Response to Action Matrix Column 2 (Regulatory Response) Inputs,” is to review and selectively challenge aspects of the licensee’s root cause evaluation, but not to perform an independent assessment of the performance issue. However, the identification by the NRC of significant issues pertaining to the adequacy of the licensee’s root cause evaluation may result in the expansion of the procedure as necessary to independently complete the inspection requirements. Also, the original performance issue will not be removed from consideration of actions in the Action Matrix until satisfaction of all supplemental inspection objectives.

The objective of supplemental IP 95002, “Supplemental Inspection Response to Action Matrix Column 3 (Degraded Performance) Inputs,” is not only to review and selectively challenge aspects of the licensee’s root cause evaluation, but to also independently assess the extent of condition for the individual and collective risk significant performance issues that warranted this supplemental inspection.

In general, all inspection requirements contained in these two procedures are intended to be addressed for each issue; however, the extent that they are reviewed and their specific applicability to the given issue will necessarily vary. The staff determined that this level of flexibility was necessary given the various issues that potentially could lead to supplemental inspections.

Also, these two supplemental inspections are intended to provide the information the NRC needs in order to assess safety. The NRC can acquire this information by performing independent inspections or can acquire the information by reviewing the licensee’s efforts to assess the root cause of the issue. If the licensee chooses not to provide some of the information needed to satisfy the inspection requirements, the NRC always has the option of acquiring this information by independent inspection.

At the highest level, supplemental IP 95003, “Supplemental Inspection Response to Action Matrix Column 4 (Multiple/Repetitive Degraded Cornerstone) Inputs,” is intended to determine the breadth and depth of safety, organizational, and programmatic issues. This supplemental procedure is more diagnostic than indicative and includes reviews of programs and processes not inspected as part of the baseline inspection program. While the procedure does allow for focus to be applied to areas where performance issues have been previously identified, the procedure requires that some sample reviews be performed for all key attributes of the effected strategic performance areas. The rationale behind this is that additional NRC assurance is required to ensure public health and safety, beyond that provided by the baseline inspection program and the PIs at those facilities where significant performance issues have been identified. The results of this inspection will aid the NRC in deciding whether additional regulatory actions are necessary to assure public health and safety. These additional regulatory actions could include orders, confirmatory action letters, or additional supplemental inspections, as necessary to confirm that corrective actions to the identified performance concerns have been effective.

Exhibits 46 through 48 provide additional information on the scope and basis for each of the supplemental inspection procedures.

# 0308.2-05	EVENT RESPONSE

Management Directive 8.3 provides the criteria for NRC investigatory response to significant operational events involving reactor and nonreactor facilities licensed by the NRC. The criteria define several levels of response, including an IIT and AIT. IITs inspect events having greater health and safety significance than events inspected by AITs.

As part of the development of the new reactor oversight process (ROP), MD 8.3 was revised to riskinform the deterministic criteria for event response at reactor facilities. The previous deterministic criteria for IITs and AITs is now evaluated in conjunction with risk in order to identify a graded response, based in part on the risk metric. The graded response will consist of an IIT, AIT, and Special Inspection for the lowest level of response. The risk metric of conditional core damage probability (CCDP) is used to best reflect the full extent of any loss of defenseindepth due to the event, regardless of whether the cause is due to licensee performance or otherwise. Numerical risk estimation by itself is not meaningful unless accompanied by an understanding of the most influential related assumptions and uncertainties.

One comment received during ROP development was that the risk evaluation to support a prompt NRC determination of the level of response will be hampered by lack of information in the early stages of the event. In addition, information from the NRC event response inspection may significantly revise the risk value and thereby require a different level of NRC response. This problem stresses the need to use deterministic criteria in conjunction with risk insights. In addition, the program has the flexibility to revise the level of response based on new information and changing risk levels.

Exhibit 49 provides additional information regarding the scope and basis for the Special Inspection level of event response.

# 0308.2-06	OVERSIGHT OF PLANTS IN EXTENDED SHUTDOWN

During the development of the new ROP, the staff also significantly revised its process for overseeing plants in an extended shutdown for performance problems. This process was riskinformed through new criteria that better focuses agency attention on those safety significant issues that contributed to the shutdown. These changes also made the process more objective by using the Action Matrix and SDP to establish criteria and thresholds for actions. The new guidance for plants in extended shutdowns for performance problems was incorporated into a revision to IMC 0350, “Staff Guidelines for the Assessment and Review of Plants that Are Not Under the Routine Reactor Oversight Process.” The title for IMC 0350 was later changed to “Oversight of Operating Reactor Facilities in a Shutdown Condition Due to Significant Performance and/or Operational Concerns.”

The three major aspects of the IMC 0350 process are: (1) the criteria for placing a plant into the process, (2) the scope of issues for the IMC 0350 required restart panel, and (3) the criteria for removing a plant from this process and placing it back into the routine ROP. The thresholds for placing a plant into the IMC 0350 process have been riskinformed and made more objective by using the assessment program Action Matrix. Consideration is given for placing a plant in the IMC 0350 process when a licensee’s performance is determined to be in the Multiple/Repetitive Degraded Cornerstone column of the Action Matrix, and the plant is expected to be shut down for at least 3 months.

The second area of the IMC 0350 process is the criteria used to determine what issues need to be resolved before restarting a plant. The scope of issues to be considered prior to restart has been risk informed by using the SDP. The issue(s) that have to be resolved before a plant restarts, and would be within the restart panel’s scope, should have risk significance (i.e., White, Yellow, or Red), but the issue(s) would not be limited to any specific performance area.

The third area of the IMC 0350 process is the criteria for returning a plant to the routine ROP. These criteria also have been riskinformed by using the Action Matrix. The approval for exiting the process and returning a plant to the routine ROP is based on the licensee satisfactorily resolving all performance issues with low to moderate or greater risk significance (i.e., White, Yellow, or Red), and has operated the plant for approximately four calendar quarters while accumulating PI data needed for the routine assessment process.

After a year of implementation under the ROP, and from the experience gained with one licensee that was under the IMC 0350 process, the staff revised the process to clarify the conditions for entering the process, as well as clarifying the responsibilities of designated positions, while encouraging the continued collection of PIs, if they remain valid. In December 2003, IMC 0350 was revised to provide a comprehensive correlation between aspects of the ROP and the IMC 0350 process, to provide an enhanced structure to the inspection approach for IMC 0350 plants, and to incorporate other lessons learned and clarifications. Additional detail on the scope and basis for the IMC 0350 process can be found in exhibit 50.

# 0308.2-07	THRESHOLD FOR DOCUMENTING FINDINGS AND INSIGHTS

The ROP uses PIs and inspection findings evaluated for risk in determining a plant’s performance in meeting the objectives of the seven cornerstones, and in determining agency actions. Therefore, the format for inspection reports for the baseline inspection program was developed to document only those issues that meet a minimum threshold for safety importance. This change removed from the reports much of the discussion regarding inspector observations of licensee activities (both positive and negative), minor findings and minor violations identified by the inspector, and licensee identified findings of very low significance that would not be used in objectively assessing performance.

Stakeholder feedback during the pilot program indicated that many inspectors and regional managers were uncomfortable with removing from inspection reports these “insights” into licensees’ performance. The inspectors and regional managers feel they need these observations (i.e., issues that may have very little or no risk significance individually) to better assess crosscutting areas, such as problem identification and resolution and human performance. Some licensees also expressed their concern with no longer having these insights and observations from NRC inspectors. Therefore, the guidance for inspection reports was changed to allow inspectors to document observations in conjunction with and to support a finding.

The assessment program was developed to use objective and repeatable indications of problems to assist the NRC in assessing licensee performance and to determine the appropriate level of NRC’s response. Positive findings, which generally are subjective and usually have no measurable basis in regulation or safety, were not included in the assessment process. Therefore, it was decided that they should not be documented in inspection reports. Stakeholder comment was solicited during the development and pilot program for the ROP to determine whether positive inspection findings should be captured and incorporated into the oversight process.

The consensus was that the NRC does not have objective criteria for evaluating positive findings. Therefore, because the assessment process does not explicitly incorporate positive findings, they should not be documented in inspection reports. However, positive aspects of licensee operations will be reflected in those items for which the SDP credits mitigation capabilities, and those positive aspects will be recorded in inspection reports as assumptions used in characterizing inspection findings.

Although not documented and used in the assessment process, inspection observations (both positive and negative) and minor violations should be verbally communicated to the licensee in routine interactions by the inspectors during the inspection. It was also considered to be more appropriate for licensees to communicate positive aspects of their operations to the public.

# 0308.2-08	OTHER INSPECTION PROGRAM ASPECTS CONSIDERED BUT NOT INCLUDED 

While developing the baseline inspection program, several additional inspectable areas were considered, but not included in the program. These other inspectable areas were not included for several reasons, such as adequate coverage of the area by an existing PI, adequate coverage by another inspectable area, or having low safety significance relative to the other inspectable areas in the program. Some of these additional inspectable areas that were considered, and the basis for not including them in the baseline inspection program, are listed in table 2 of this attachment. In addition to inspectable areas, table 2 also includes other inspection program attributes that were considered but not included.

Issue Date: 10/31/23	1	0308 Att 2

# BASIS SUMMARY SHEETS

Exhibit 1: Adverse Weather Protection (IP 71111.01)

Exhibit 2: Reserved (IP 71111.02)

Exhibit 3: Reserved (IP 71111.03)

Exhibit 4: Equipment Alignment (IP 71111.04)

Exhibit 5: Fire Protection (IP 71111.05)

Exhibit 6: Flood Protection Measures (IP 71111.06)

Exhibit 7: Heat Exchanger/Sink Performance (IP 71111.07)

Exhibit 8: Inservice Inspection Activities (IP 71111.08)

Exhibit 9: Reserved (IP 71111.09)

Exhibit 10: Reserved (IP 71111.10)

Exhibit 11: Licensed Operator Requalification Program and Licensed Operator Performance (IP 71111.11)

Exhibit 12: Maintenance Effectiveness (IP 71111.12)

Exhibit 13: Maintenance Risk Assessments and EmergentWork Control (IP 71111.13)

Exhibit 14: Reserved (IP 71111.14)

Exhibit 15: Operability Determinations and Functionality Assessments (IP 71111.15)

Exhibit 16: Reserved (IP 71111.16)

Exhibit 17: Evaluations of Changes, Tests, and Experiments (IP 71111.17T)

Exhibit 18: Plant Modifications (IP 71111.18)

Exhibit 19: Post-Maintenance Testing (IP 71111.19)

Exhibit 20: Refueling and Other Outage Activities (IP 71111.20)

Exhibit 21: Safety System Design and Performance Capability (IP 71111.21)

Exhibit 22: Surveillance Testing (IP 71111.22)

Exhibit 23: Reserved (IP 71111.23)

Exhibit 24: Testing and Maintenance of Equipment Important to Risk (IP 71111.24)

Exhibit 25: Biennial Exercise and Drill Inspection (IP 71114.01, .06, .07 and .08)

Exhibit 26: Alert and Notification System Evaluation (IP 71114.02)

Exhibit 27: Emergency Response Organization Staffing and Augmentation System (IP 71114.03)

Exhibit 28: Emergency Action Level and Emergency Plan Changes (IP 71114.04)

Exhibit 29: Correction of Emergency Preparedness Weaknesses and Deficiencies (IP 71114.05)

Exhibit 30: Access Control to Radiologically Significant Areas (IPs 71124.01, 71124.04, and 71124.08)

Exhibit 31: ALARA Planning and Controls (IPs 71124.01, and 71124.03)

Exhibit 32: Radiation Monitoring Instrumentation(IPs 71124.04 and IP 71124.05)

Exhibit 33: Radioactive Gaseous and Liquid Effluent Treatment and Monitoring Systems (IPs 71124.05, 71124.06, and 71124.07)

Exhibit 34: Radiological Environmental Monitoring Program (REMP) (IP 71124.07)

Exhibit 35: Radioactive Material Processing and Radioactive Material Handling, Storage, and Transportation (IP 71124.08)

Exhibit 36: Discrepant or Unreported Performance Indicator Data (IP 71150)

Exhibit 37: Performance Indicator Verification (IP 71151)

Exhibit 38: Problem Identification and Resolution (PI&R) (IP 71152)

Exhibit 39: Follow up of Events and Notices of Enforcement Discretion (IP 71153)

Exhibit 40: Light Water Reactor Inspection Program – Operations Phase (IMC 2515)

Exhibit 41: Risk Informed Baseline Inspection Program (IMC 2515, Appendix A)

Exhibit 42: Supplemental Inspection Program (IMC 2515, Appendix B)

Exhibit 43: Special and Infrequently Performed Inspections(IMC 2515, Appendix C)

Exhibit 44: Plant Status (IMC 2515, Appendix D)

Exhibit 45: Inspection Program Modifications During Public Health Emergencies Or Other Conditions Restricting Inspector Onsite Presence (IMC 2515, Appendix E)

Exhibit 46: Supplemental Inspection Response to Action Matrix Column 2 (Regulatory Response) Inputs (IP 95001)

Exhibit 47: Supplemental Inspection Response to Action Matrix Column 3 (Degraded Performance) Inputs (IP 95002)

Exhibit 48: Supplemental Inspection Response to Action Matrix Column 4 (Multiple/Repetitive Degraded Cornerstone) Inputs (IP 95003)

Exhibit 49: Special Inspection (IP 93812)

Exhibit 50: Oversight of Reactor Facilities in a Shutdown Condition Due to Significant Performance and/or Operational Concerns (IMC 0350)

Table 1: Inspectable Areas by Cornerstone

Table 2: Other Inspection Program Elements Considered But Not Included

Attachment 1: Revision History for IMC 0308 Attachment 2

Issue Date: 10/31/23	Att1-1	0308 Att 2