Issue Date: 02/09/11 1 95003

NRC INSPECTION MANUAL IRIB

INSPECTION PROCEDURE 95003

SUPPLEMENTAL INSPECTION FOR REPETITIVE DEGRADED CORNERSTONES, 

MULTIPLE DEGRADED CORNERSTONES, MULTIPLE YELLOW INPUTS OR 

ONE RED INPUT

PROGRAM APPLICABILITY: 2515, 2201

CORNERSTONES: ALL

INSPECTION BASIS: This procedure provides the supplemental response for 

repetitive degraded cornerstones, multiple degraded 

cornerstones, multiple yellow inputs, or one red input to 

the assessment Action Matrix as described in Inspection 

Manual Chapter (IMC) 0305, ―Operating Reactor 

Assessment Program,‖ and IMC 0320, ―Operating 

Reactor Security Assessment Program‖. The intent of 

this procedure is to provide the NRC with supplemental 

information regarding licensee performance, as 

necessary to determine the breadth and depth of safety, 

organizational, and programmatic issues. As such, this 

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 does require that some sample 

reviews be performed for all key attributes of the affected 

strategic performance areas. The rationale behind this is 

that additional NRC assurance is required to ensure 

public health and safety, and security beyond that 

provided by the baseline inspection program and the 

performance indicators 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.
Issue Date: 02/09/11 2 95003

This procedure was developed with consideration of the 

following boundary conditions:

1. The NRC is performing the inspection, which 

involves a graded approach to assessing the 

licensee’s safety culture. The scope of the 

inspection requirements related to safety culture will 

be based on the results of the validation of the 

licensee’s third party safety culture assessment and 

root cause evaluation;

2. The procedure is not intended to be used for event 

response;

3. New issues identified by the team will be evaluated 

using the significance determination process during 

the course of the inspection; other process issues 

will be documented in the inspection report; and, 

4. The procedure is intended to provide insight into 

the root and contributing causes of performance 

deficiencies, but is not intended to be a substitute

for a more focused root cause analysis (or self 

assessment) of specific performance issues to be 

performed by the licensee or by a third party; and

5. In most cases the licensee has completed a rootcause, extent-of-cause, and extent-of-condition 

investigation(s) of the performance deficiencies 

which prompted this inspection and an independent 

third-party assessment of their safety culture before 

the NRC begins this inspection. In some cases 

NRC inspection of these activities can be deferred 

when warranted to accommodate a longer time 

required for the licensee to complete them. 

Flexibility is afforded to perform inspections and 

safety culture evaluations in parallel with the 

conduct of the licensee’s root cause evaluation and 

third-party safety culture assessment. A third party 

assessment is conducted by individuals who are 

not employees of the plant or the utility operators of 

the plant.

95003-01 INSPECTION OBJECTIVES

01.01 To provide the NRC additional information to be used in deciding whether the 

continued operation of the facility is acceptable and whether additional regulatory 

actions are necessary to arrest declining plant performance.

01.02 To provide an independent assessment of the extent of risk significant issues to 

aid in the determination of whether an unacceptable margin of safety or security exists.

01.03 To independently assess the adequacy of the programs and processes used by 

the licensee to identify, evaluate, and correct performance issues.
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01.04 To independently evaluate the adequacy of programs and processes in the 

affected strategic performance areas.

01.05 To provide insight into the overall root and contributing causes of identified 

performance deficiencies.

01.06 To determine if the NRC oversight process provided sufficient warning to 

significant reductions in safety.

01.07 To evaluate the licensee=s third-party safety culture assessment and conduct a 

graded assessment of the licensee’s safety culture based on the results of the 

evaluation. [C1]

95003-02 INSPECTION REQUIREMENTS

The intent of this procedure is to allow the NRC to obtain a comprehensive 

understanding of the depth and breadth of safety, organizational, and performance 

issues at facilities where data indicates the potential for serious performance 

degradation. Considerable leeway has been built into the procedure to allow it to be 

customized, to better reflect the specific nature of the previously identified performance 

issues.

This procedure was written with the assumption that supplemental inspections (either 

Inspection Procedure (IP) 95001 or IP 95002) have been conducted to evaluate the 

licensee=s root cause, extent-of-cause, and extent-of-condition evaluations and 

associated corrective actions for Awhite@ or greater performance indicators or inspection 

findings. If such supplemental inspections have not been conducted, the scope of this 

inspection should include inspection of the licensee=s evaluation of those issues.

02.01 Strategic Performance Area(s) Identification.

a. Using the information contained in the Assessment Action Matrix, identify the 

strategic performance areas for which performance has significantly declined 

(e.g. Reactor Safety, Radiation Safety, or Safeguards). The scope of this 

inspection will generally include all key attributes of the degraded strategic 

performance areas. Specific inspection requirements pertaining to each 

strategic performance area are contained in Sections 02.03 - 02.06 of the 

procedure.

b. Inspection Requirements 02.02, and 02.07 - 02.12 should always be performed 

regardless of the strategic performance areas selected for review. Attachment 

95003.01, AAdditional Emergency Preparedness Cornerstone Inspection,@ to 

this procedure should be performed when Emergency Preparedness (EP) 

Cornerstone performance issues are a contributing factor to the reason this 

procedure is being implemented, e.g., the EP Cornerstone is degraded and 

Reactor Safety is the relevant strategic performance area. However, 

implementation of Attachment 95003.01 should be considered during the 

performance of all 95003 inspections. When Attachment 95003.01 is 

implemented it supplants the EP related inspection requirements contained in 

the body of this procedure. The use of an EP inspector from a different region 

(or headquarters) should be considered to assist in the performance of 

Attachment 95003.01.
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02.02 Review of Licensee Control Systems for Identifying, Assessing, and Correcting 

Performance Deficiencies. Once significant performance concerns have been identified 

in the Action Matrix, the NRC must ensure that licensee systems for identifying, 

assessing, and correcting performance deficiencies are sufficient to prevent further 

performance degradations. The following inspection requirements evaluate whether 

licensee programs are sufficient to prevent further declines in safety that could result in 

unsafe operation.

a. Determine whether licensee evaluations of, and corrective actions to, significant 

performance deficiencies have been sufficient to correct the deficiencies and 

prevent recurrence.

b. Evaluate the effectiveness of audits and assessments performed by the quality 

assurance group, line organizations, and external organizations. Focus on how 

the performance data is integrated with other data to arrest declining 

performance. This review should include the organization=s response to EP 

related corrective actions identified as a result of actual events, exercises and 

drills.

c. Determine whether the process for allocating resources provides for appropriate 

consideration of safety and compliance, and whether appropriate consideration 

is given to the management of maintenance backlogs and correction of workarounds.

d. Evaluate whether licensee performance goals are congruent with those 

corrective actions needed to address the documented performance issues.

e. By reviewing selected aspects of the employee concerns program and the 

results of surveys or other workplace environment evaluations, ensure that 

employees are not hesitant to raise safety concerns and that safety significant 

concerns entered into the employee concern program receive an appropriate 

level of attention.

f. Determine whether there is a mechanism for all members of the workforce to 

suggest improvements and explain their disagreements with technical 

resolutions of identified deficiencies. Determine whether there is a feedback 

mechanism in which the evaluation of deficiencies and follow-up corrective 

actions are reported back to the identifying workers.

g. Evaluate the effectiveness of the organization=s use of industry information for 

previously documented performance issues.

02.03 Assessment of Performance in the Reactor Safety Strategic Performance Area 

(Initiating Events, Mitigation Systems, Barrier Integrity, and Emergency Preparedness 

Cornerstones).

a. Inspection Preparation

1. Develop an information base to allow the review of the effectiveness of 

corrective actions.

(a) Compile performance information from the licensee=s corrective 

action program, audits, self-assessments, licensee event reports 

(LERs), and the inspection report record (both the inspection reports 

and the PIM) for the time period determined by the team manager. 

To the extent possible keeping in mind the needs of the inspection 
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team, maximize the use of electronic data from the licensee and 

minimize the impact of the data request on the licensee.

(b) Review the compiled information and sort the issues by the key 

attributes listed below. Licensee corrective actions for the issues 

should be assessed as part of the following key attribute reviews.

2. Select a system(s) for focus using the plant specific individual plant 

evaluation (IPE) and issues identified as part of the performance 

information developed above.

3. Review inspection reports and critique findings from EP related event 

response and drills. Review a summary of recent EP corrective actions. 

Review recent changes to the Emergency Plan (Plan) changes. Review 

licensee analyses of corrective actions related to specific findings and 

general audits where available. Develop an inspection plan to address 

concerns identified as well as the inspection requirements.

4. Perform the following inspection requirements for each key attribute 

focusing on the selected system. While the inspectors should focus on 

the selected system, other systems and components may be reviewed as 

necessary to assess licensee performance for the following key attributes.

b. Key Attribute - Design. Inadequacies in the design, the as-built configuration, or 

the post- installation testing of plant modifications can cause initiating events, 

affect the capability and reliability of mitigating systems, and the margin of safety 

in barrier design. As plants age, their design basis may be misunderstood or 

forgotten such that an important design feature may be inadvertently removed or 

disabled as changes are made to the plant.

Independently assess the extent of risk significant design issues by performing 

the following inspection requirements. The review shall cover the as-built design 

features of the selected system to verify its capability to perform its intended 

functions with a sufficient margin of safety. Focus will be on system 

modifications rather than original system design. Information from this inspection 

will be used to assess the licensee=s ability to maintain and operate the facility in 

accordance with the design basis.

1. Assess the effectiveness of corrective actions for deficiencies involving 

design.

2. Select several modifications to the system for review and determine if the 

system is capable of functioning as specified by the current design and 

licensing documents, regulatory requirements, and commitments for the 

facility.

3. Determine if the system is operated consistent with the design and 

licensing documents.

4. Evaluate the interfaces between engineering, plant operations, 

maintenance, and plant support groups.

c. Key Attribute - Human Performance. By nature of the design of nuclear power 

plants and the role of plant personnel in maintenance, testing and operation; 

human performance plays an important part in normal, off-normal and 
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emergency operations. Human performance impacts each of the cornerstones 

and therefore should be considered across this entire inspection.

The team members reviewing this key attribute should coordinate their activities 

to ensure that the following inspection requirements are addressed:

1. Assess the effectiveness of corrective actions for identifying, evaluating, 

and correcting deficiencies involving human performance.

2. Review specific problem areas and issues identified by inspections to 

determine if concerns exist in the human performance cross-cutting area 

components as detailed in IMC 0310, AComponents within the CrossCutting Areas.@.

3. Conduct EP Emergency Response Organization Performance-Drills, in 

accordance with IP 82001, ―Evaluation of Emergency Preparedness‖ with 

a sampling of shift crews and management teams to assess their ability to 

implement the Emergency Plan. 

d. Key Attribute - Procedure Quality. Inadequate procedures can cause initiating 

events by inducing plant personnel to take inappropriate actions during plant 

operations, maintenance, calibration, testing, or event response. Adequate 

procedures also assure proper functioning of mitigating systems during 

operation, maintenance, and testing. Emergency and abnormal operating 

procedures are also essential for mitigating system performance and assuring 

appropriate actions will be taken to preserve reactor coolant system (RCS) and 

containment integrity. To the extent that there are procedure deficiencies 

associated with the above noted activities, they should be identified as causes of 

problems in other key attributes.

Determine the technical adequacy of procedures by verifying that they are 

consistent with desired actions and modes of operation by completing the 

following inspection requirements.

1. Assess the effectiveness of corrective actions for deficiencies involving 

procedure quality.

2. Evaluate the quality of procedures and as applicable determine the 

adequacy of the procedure development and revision processes.

3. Review a sample of Emergency Plan Implementing Procedure (EPIPs) 

changes against the requirements of the Plan and corrective action 

assessments. Determine if the EPIP change process is adequate in 

correcting EPIP related deficiencies and maintaining Plan commitments in 

EPIP instructions.

e. Key Attribute - Equipment Performance. Equipment failure or degradation can 

cause initiating events during power operation and losses of decay heat removal 

during shutdowns. To limit challenges to safety functions due to equipment 

problems, licensees should have programs to achieve a high degree of 

availability and reliability of equipment that can cause initiating events. The 

availability and reliability of equipment is also critical to mitigating the impact of 

initiating events on plant safety. Strong preventive and corrective maintenance 

programs are an integral part of assuring equipment availability and reliability.
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Determine that the licensee is adequately maintaining and testing the functional 

capability of risk significant systems and components by completing the 

following inspection requirements.

1. Assess the effectiveness of corrective actions for deficiencies involving 

equipment performance, including equipment designated for increased 

monitoring via implementation of the Maintenance Rule. 

2. Determine if the licensee has effectively implemented programs for control 

and evaluation of surveillance testing, calibration, and post-maintenance 

testing.

3. Assess the operational performance of the selected safety system to verify 

its capability of performing the intended safety functions.

4. Review a sample of EP related equipment and facilities (including 

communications gear) against Plan commitments. Review the adequacy 

of the surveillance program to maintain equipment and facilities. Review 

the correction of deficiencies identified by the surveillance program. 

5. Assess decision-making regarding longstanding equipment issues (i.e.

whether conservative decisions were made and decisions supported long 

term equipment reliability)

6. For any unresolved long-term equipment issues, determine whether 

inadequate resources were a cause or contributed to any inappropriate 

delay in resolving those issues.

f. Key Attribute - Configuration Control. Loss of configuration control of risksignificant systems or equipment can lead to the initiation of a reactor transient 

and/or can compromise mitigation capability. Maintaining proper water 

chemistry in the RCS is essential to long term reliability of both the nuclear fuel 

and the RCS pressure boundary. Proper configuration control is necessary to 

maintain assurance that the RCS pressure boundary is maintained intact and 

monitored for degradation. Containment integrity depends on maintaining the 

configuration of penetrations and safety-related systems that need to respond 

following an accident. Also, maintaining the containment within its design limits 

ensure that it will be able to accommodate a design basis or severe accident.

Assess the licensee=s ability to maintain risk-significant systems and the 

principle fission product barriers in configurations which support their safety 

functions by completing the following inspection requirements.

1. Assess the effectiveness of corrective actions for deficiencies involving 

configuration control.

2. Perform a walkdown of the selected system. In addition, if the selected 

system does not directly have a containment over-pressure safety function 

(such as containment spray), conduct an additional review of such a 

system.
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(a) Independently verify that the selected safety system is in proper 

configuration through a system walkdown.

(b) Review temporary modifications to ensure proper installation in 

accordance with the design information.

3. Determine that the work control process uses risk appropriately during 

planning and scheduling of maintenance and surveillance testing activities 

and the control of emergent work.

4. Determine whether the primary and secondary chemistry control programs 

adequately control the quality of plant process water to ensure long-term 

integrity of the reactor coolant pressure boundary.

5. Assess the programs and controls (tracking systems) in place for 

maintaining knowledge of the configuration of the fission product barriers 

including: containment leakage monitoring and tracking, containment 

isolation device operability (valves, blank flanges), and reactor coolant 

leak-rate calculation and monitoring.

6. Review the results of the plant specific IPE relative to the system(s) 

selected. Determine if the IPE is being maintained to reflect actual system 

conditions regarding system capability and reliability.

g. Key Attribute - Emergency Response Organization Readiness. Implementation 

of the Emergency Response Plan is dependent on the readiness of the 

emergency response organization to respond to an emergency. In this usage, 

Areadiness@ means the ability of the licensee to activate timely Emergency 

Response Organization (ERO) augmentation of on shift personnel as necessary 

to implement the emergency plan. Self-assessments of readiness during drills 

and activation tests are used to identify areas for improvement. Selfassessment and corrective action resolution is critical to ERO readiness.

1. Assess the effectiveness of corrective actions for deficiencies involving 

ERO readiness.

2. Verify that adequate staffing is available on shift for emergencies.

3. Verify the capability to activate and staff the emergency response facilities 

and augment the response organization within the requirements of the 

licensee emergency response plan

4. Verify licensee ability to meet Emergency Plan goals for activation by 

implementing IP 71114.03, AEmergency Preparedness Organization 

Staffing and Augmentation System.@ If this IP has been implemented 

recently, the inspector may exercise judgment as to the need to implement 

the IP as part of the 95003 inspection effort. If Attachment 95003.01 is 

being implemented, there are additional requirements under this key 

attribute to consider.
Issue Date: 02/09/11 9 95003

02.04 Assessment of Performance in the Radiation Safety Strategic Performance Area 

- Occupational Radiation Safety.

a. Inspection Preparation

1. Develop an information base to allow review of the effectiveness of 

corrective actions.

2. Compile performance information from the licensee=s corrective action 

program, audits, self-assessments, LERs, and the inspection report record 

(both the inspection reports and the PIM) for the designated time period.

3. Review the compiled information and sort the issues by the key attributes 

listed below.

4. Perform the following inspection requirements for each key attribute. Note 

that specific areas such as external and internal dosimetry are not 

specifically delineated and treated as key attributes, but should they need 

to be closely examined, the inspector should use the procedures listed in 

IMC 2515, Appendix B, Attachment 1 for evaluating extent of condition.

b. Key Attribute - Program/Processes for Occupational Radiation Safety. The 

effective implementation of the required radiation protection (RP) program and 

implementing procedures contribute to proper control and minimization of 

occupational exposures. Programmatic deficiencies, inadequate procedures, 

and/or improper implementation have all resulted in significant, uncontrolled

occupational exposures in excess of regulatory limits (both from internal and 

external radiation sources). Worker radiation exposure controls are governed by 

both administrative and physical controls which serve as protective barriers that 

prevent excessive, unintended exposures in high and very high radiation areas, 

and significantly contaminated and airborne areas. The facility’s ALARA 

program focuses on ensuring that plant operations and maintenance activities 

are performed using planning, methods and procedures based to achieve 

occupational doses that are ALARA. 

The team members reviewing these aspects of the key attribute should 

coordinate their activities to ensure that the following inspection requirements 

are addressed:

1. Assess the RP organization to ensure it is clearly defined (assignment of 

duties, authorities, and responsibilities); scope of program and the staffing 

are adequate.

2. Select several implementing procedures (from three or more 

programmatic areas - job controls/coverage, surveys, RWP issuance, etc.) 

and evaluate their technical adequacy. Focus on problem areas identified 

in previous inspections. Review the procedure development process and 

determine its adequacy.
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3. Observe planned work activities in high radiation, high airborne, and/or 

highly contaminated areas and determine effectiveness of work planning, 

coordination, implementation and lessons learned.

4. Determine adequacy/implementation, with respect to ALARA, of the 

radiological source term controls and quality of related chemistry controls.

5. Determine adequacy/implementation, with respect to ALARA, of work 

planning and controls, focusing on outage maintenance periods.

6. Determine effectiveness and degree of management support and 

integration of ALARA into facility craft work units. 

c. Key Attribute - Plant Facilities/Equipment & Instrumentation for Occupational 

Radiation Safety. To properly conduct adequate radiation monitoring and 

surveillance activities and to protect workers, the facility is required to maintain 

fixed and portable radiation survey equipment (for airborne and external 

hazards), respiratory protection, communication, temporary ventilation and 

shielding, and anti-contamination clothing. Routine calibration and maintenance 

of this equipment ensures its continued operability. If problems are identified 

during the inspection preparation or during the conduct of this team inspection, 

the team should pursue the inspection requirements below. The team members 

reviewing this aspect of the key attribute should coordinate their activities to 

ensure that the following inspection requirements are addressed:

1. Select several implementing procedures (from three or more 

programmatic areas - survey instrument calibration, self-contained 

breathing apparatus maintenance, etc.) and evaluate their technical 

adequacy. Focus on problem areas identified in previous inspections. 

Review the procedure development process and determine its adequacy.

2. Observe several planned equipment maintenance or calibration activities. 

If possible, focus on equipment used in high risk areas (high radiation or 

airborne areas, potential oxygen-deficient, immediately dangerous to life 

or health (IDLH) areas, etc.).

3. Determine level of management support in maintaining adequate 

equipment and support facilities.

4. Review any recommendations for plant improvements to support radiation 

safety and determine whether the decision based on these 

recommendations sufficiently supported radiation protection.

d. Key Attribute - Human Performance for Occupational Radiation Safety. Worker 

performance has an obvious, important impact on work activities in radiological 

areas. Two of the major components are health physics technician (HPT) and 

general radiation worker (crafts) groups. Human performance is impacted by 

several vital factors -- qualification and training. The selection, qualification and 

training requirements for facility personnel are generally governed by a 

commitment in the plant technical specifications (to a ANS standard). For HPTs 

and others, 10 CFR 50.120 (training rule) requires HPTs (including contractors) 

to be task qualified for their assigned normal and outage duties.
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The team members reviewing this key attribute should coordinate their activities 

to ensure that the following inspection requirements are addressed:

1. Assess the effectiveness of identifying, evaluating, and correcting 

deficiencies involving human performance.

2. Review specific problem areas and issues identified by inspections to 

determine if concerns exist in the human performance cross-cutting area 

components as detailed in IMC 0310.

02.05 Assessment of Performance in the Radiation Safety Strategic Performance Area 

- Public Radiation Safety (Radiological Effluent Monitoring, Radioactive Material 

Control, and Transportation of Radioactive Material).

a. Key Attribute - Plant Facilities / Equipment and Instrumentation for Public 

Radiation Safety. The improper installation, modification, maintenance, or 

calibration of radioactive effluent monitoring equipment, and associated 

radiochemistry laboratory equipment and meteorological system equipment can 

adversely affect licensee performance in achieving and demonstrating 

compliance with regulatory limits and ALARA design objectives for radioactive 

effluents. For transportation activities, shipping packages not prepared in 

accordance with their applicable design requirements increase the potential for 

unexpected exposure or loss of radioactive material which could result in 

uncontrolled and unnecessary exposure to members of the public. To prevent 

the inadvertent release of licensed radioactive material from the licensee=s 

control requires the use of sensitive radiation survey equipment that is properly 

setup and calibrated.

The team members reviewing this area should ensure the following 

requirements are addressed:

1. Review the results of audits and appraisals performed for the designated 

time period. Review deficiency reports (also referred to as incident reports 

or condition reports) issued for the area being inspected.

2. Perform a walkdown of the selected facility or equipment to assess its 

physical condition. Review any significant changes made by the licensee 

to the facilities or equipment that were not included in the prior inspection 

period.

3. Determine the level of management support for the maintenance of 

facilities and equipment of the program.

b. Key Attribute - Program/Process for Public Radiation Safety. Procedures must 

be technically adequate and implemented appropriately to ensure the proper 

processing, control, and discharge of radioactive effluents into the environment. 

For transportation activities, procedural guidance is necessary for the proper 

evaluation of radioactive waste to determine the quantities and types of 

radioactive material present for the selection and preparation of shipping 

packages. Detailed procedures are required to conduct radiation surveys of the 

packaged radioactive waste to ensure that radiation levels are within regulatory 

limits. The performance of radiation surveys on equipment and material to be 

released from the licensee=s facility requires appropriate policy and technical 

procedural guidance for handling and processing a wide variety of potentially 

contaminated materials.
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The team members should review the licensee=s program documents and 

implementing procedures to ensure the following requirements are addressed:

1. Review the results of audits and appraisals performed for the designated 

time period. Review deficiency reports (also referred to as incident reports 

or condition reports) issued for the area being inspected.

2. Select several implementing procedures from the area being inspected 

and review for general quality (i.e., clearly written, contain specific actions, 

and contain data record sheets) and technical adequacy.

3. Review the licensee=s program documents and implementing procedures 

for any recent significant changes in the area being inspected.

4. Review the records which resulted from the implementation of the 

selected procedures. Review them to determine if the procedure was 

correctly used.

c. Key Attribute - Human Performance for Public Radiation Safety. Human 

performance can directly affect radioactive waste processing, radioactive 

effluent processing, and transportation programs. It is important to ensure that 

plant workers are adequately trained and qualified to perform their job function. 

Periodic retraining is also needed to ensure that workers maintain their 

qualifications and are updated with new information and requirements.

The team members reviewing this key attribute should coordinate their activities 

to ensure that the following inspection requirements are addressed:

1. Assess the effectiveness of identifying, evaluating, and correcting 

deficiencies involving human performance.

2. Review specific problem areas and issues identified by inspections to 

determine if concerns exist in the human performance cross-cutting area 

components as detailed in IMC 0310.

3. Observe the performance of activities described in the selected 

procedures. If the activities are not scheduled to be performed during the 

inspection time period, request that the activity be simulated so that 

worker performance and the adequacy of the procedure can generally be 

assessed.

4. Interview several personnel (i.e., technicians, engineers, health physicists, 

and supervisors) associated with the program to assess their level of 

knowledge about the program and procedures.

02.06 Assessment of Performance in the Safeguards Strategic Performance Area

(Security Cornerstone).

a. Key Attribute—Physical Protection (PP). The objective of the physical protection 

program is to provide high assurance that the security systems and programs 

effectively protect against the design basis threat (DBT) and the loss or theft of 

special nuclear material (SNM). This area is based on defense in-depth 

strategies that are intended to provide protection against both internal and 

external threats by making effective use of all security equipment and security

measures established to support the security of the facility and the 

implementation of the licensee's protective strategy. Other elements that are 
Issue Date: 02/09/11 13 95003

critical to the effective implementation of physical protection program are 

security officer training and qualifications, weapons testing and maintenance, 

equipment testing and maintenance, implementing procedures and the 

processes to maintain security-related records pertaining to these elements.

Verify that the licensee has established and maintains a physical protection 

program which includes systems, measures, and the necessary infrastructure to 

support the protection of the facility and the implementation of the protective 

strategy in accordance with the NRC-approved security plan, regulatory 

requirements, and any other applicable Commission requirements.

b. Key Attribute - Access Authorization (AA). The personnel screening process is 

the process used to verify trustworthiness and reliability of personnel prior to 

granting unescorted access to the protected area. A failure in the Access 

Authorization Program can compromise the licensee=s ability to adequately 

protect against the insider threat of radiological sabotage.

Assess the licensee=s ability to implement the behavior observation portion of 

the personnel screening, insider mitigation program, and fitness for duty 

program.

1. Verify that the licensee is identifying problems related to the access 

authorization program at an appropriate threshold and entering those 

problems into the corrective action program. 

2. Verify that the licensee has appropriately resolved the concerns and

regulatory requirements for a selected sample of problems associated with 

AA.

3. Verify whether issues identified are programmatic or specific to an 

individual.

c. Key Attribute - Access Control (Searches of personnel, materials, packages, and 

vehicles). The Access Control program=s function is to prevent the introduction 

of contraband (i.e. firearms, explosives, incendiary devices) and unauthorized 

personnel into the plant that could aid in the attempt to commit radiological 

sabotage. The failure of the Access Control program could compromise 

required security measures in place that are required to protect vital and risk 

significant plant equipment and functions.

1. Assess whether the licensee has effective access controls, and equipment 

in place designed and functioning as intended to detect and prevent the 

introduction of contraband or unauthorized personnel into the protected 

area that could be used to commit or aid in the act of sabotage.

2. Verify that the access authorization process is properly used to ensure 

that only those personnel who have been properly screened are granted 

unescorted access to the protected and vital area. Verify that the licensee 

has procedures in place to establish, implement, and maintain a list of 

individuals who are authorized to have unescorted access to those vital 

areas during nonemergency conditions. Ensure procedures are in place 

and followed for denial or withdrawal of unescorted access and 

reinstatement, as appropriate.
Issue Date: 02/09/11 14 95003

d. Key Attribute - Response to Contingency Events (protection strategy, program 

design, and support elements). The purpose of the licensee=s contingency 

response program is to provide high assurance that the licensee can protect 

identified target sets against the design basis threat and thus prevent core 

damage. The licensee should have developed a response strategy and the 

associated security response infrastructure necessary to maintain, update, and 

implement that strategy. The strategy=s objective shall be to respond with 

sufficient force that is properly armed and appropriately trained and within the 

appropriate time to protected positions to interdict and defeat the design basis 

threat in order to protect against radiological sabotage.

Verify that the licensee has established and maintains a contingency response 

program and the necessary infrastructure to support the performance of an 

adequate protective strategy. 

e. Key Attribute - Response to Contingency Events (performance-based force-onforce exercises and target set evaluation). The licensee should be able to 

respond to contingency events with sufficient force that is properly armed and

appropriately trained and within the appropriate time to defeat the design basis 

threat in order to protect against radiological sabotage.

1. Verify through review of documents and discussions with the licensee that 

the licensee has appropriately identified its target sets and has a 

protective strategy associated to those target sets.

2. Verify through the conduct of table-top drills and NRC evaluated exercises 

that for any selected target set, the licensee=s protective strategy is 

adequate and it can protect against the design basis threat.

f. Key Attribute—Material Control and Accounting (MC&A). The basic objective of 

MC&A is to protect against the loss or misuse of special nuclear material (SNM), 

i.e., enriched uranium or plutonium. MC&A works in concert with physical 

protection to complete the Security Cornerstone with MC&A providing a record 

of the quantity and location of SNM at the facility, while physical protection 

protects the facility and the SNM located there. MC&A reduces the threat of 

loss or misuse in that it enables timely detection. Failure to maintain knowledge 

of the location of SNM significantly increases the risk of loss. This inspection is 

intended to assess the effectiveness of the licensee’s program to keep records 

of and confirm the location of the SNM it is licensed to possess.

Verify that the following aspects of the licensee’s MC&A program are effective 

and accurate.

1. Records. Verify, by interviewing licensee personnel and reviewing 

records, that activities involving both fuel and non-fuel (e.g., instruments) 

SNM are documented, that SNM transfer forms and physical inventory 

records are complete, and that any fuel failures and reconstitutions 

included documenting any changes to SNM content of the affected items 

(rods, pieces, etc.).

2. Program and Procedures. By interviewing licensee personnel and 

reviewing procedures, determine if changes made to the licensee’s

procedures might affect SNM records. Verify that changes to procedures, 

if any, have not reduced the effectiveness of the MC&A program, and 
Issue Date: 02/09/11 15 95003

confirm that the changes were approved by management, SNM 

responsibilities were clearly identified, and personnel were trained on the 

new procedures.

3. Physical Inventory.

(a) Verify that the licensee conducts a physical inventory at least every 

12 months.

(b) By reviewing the latest inventory and interviewing licensee 

personnel, verify that the inventory includes both fuel and non-fuel 

SNM and that it involved physically observing the items inventoried.

(c) Confirm that the results of previous inventories were reconciled with 

the book inventory (as documented in the accounting records, 

including the spent fuel map) and that discrepancies, if any, were 

entered in the licensee’s problem identification program for 

investigation and resolution.

(d) Verify by physical observation that fuel assemblies and containers 

with pieces of fuel rods and fuel are located where indicated by the 

MC&A records.

4. Reports. By interviewing licensee personnel, confirm that the licensee has 

made the required reports to the national Nuclear Materials Management 

and Safeguards System (NMMSS).

5. Resolution of Problems. Verify that the licensee is identifying issues 

related to MC&A and entering them in the corrective action program. 

Verify that the licensee has appropriately resolved a sample of MC&A 

issues, if any

g. Key Attribute - Security Plan Changes. The licensee=s Security Plan is the 

licensee=s plan for the physical protection at their site. An inadequate security 

plan can compromise the licensee=s ability to protect against the design basis 

threat and be ineffective against acts of radiological sabotage. Changes to 

basic security measures could allow a direct reduction in the effectiveness of the 

physical protection measures that are vital to maintaining adequate physical 

protection.

1. Verify that changes made to the licensee=s Security Plan do not reduce its 

effectiveness to prevent or mitigate an attack by the design basis threat or 

increase the likelihood of acts of radiological sabotage. 

2. Verify that changes to the licensee=s security procedures have not 

decreased the effectiveness of the previous plan, as required by 

10 CFR 50.54(p).

02.07 Evaluate the Licensee=s Third-Party Safety Culture Assessment.
Issue Date: 02/09/11 16 95003

The requirements in this section and the associated guidance in section 03.07 are to be 

implemented in evaluating the licensee’s third-party safety culture assessment. At such 

time that an industry safety culture assessment methodology is developed and found to 

be acceptable by the NRC, the requirements and associated guidance in this section 

will be evaluated for potential revisions to address the use of such a methodology. 

a. Inspection Preparation

1. Depending on the timing of the conduct of the licensee’s third-party safety 

culture assessment with respect to the NRC’s 95003 inspection, there 

may be opportunity for the NRC staff to engage with the licensee and the 

licensee’s third-party safety culture assessors before initiation of their 

assessment. This is preferable as it allows the NRC lead safety culture 

assessor (SCA) and other SCAs, as designated by the lead SCA, to 

evaluate the third-party safety culture assessment methodology. The 

licensee and the third-party safety culture assessors then have the 

opportunity to react to NRC concerns and comments on the methodology 

in advance of its implementation, and to interact to be informed of the 

status of safety culture assessment activities. In these cases, engage the 

licensee and third-party safety culture assessors using the requirements in 

this section and the associated guidance in section 03.07. Monitor the 

safety culture assessment implementation and the identification of issues 

that arise to the extent possible. 

After the conduct of the third-party safety culture assessment, follow the 

requirements in section 02.08 and associated guidance in section 03.08 to 

determine the scope of NRC’s graded safety culture assessment. It is 

important to note that, depending on the circumstances, engagement 

during the third-party safety culture assessment and the subsequent 

conduct of NRC’s graded assessment activities may occur over several 

months and may need to begin before the inspection period where the 

entire inspection team is onsite.

2. The licensee may have conducted a recent (i.e., within the last six 

months) third-party safety culture assessment before the 95003 inspection 

was initiated. If the licensee chooses not to perform another third-party 

safety culture assessment, the lead SCA and the SCA subteam should 

use the inspection requirements in this section and the associated 

guidance in section 03.07 to evaluate the recent third-party safety culture 

assessment. If the licensee’s most recent safety culture assessment was 

not performed recently (i.e., more than six months ago), the licensee 

would be expected to perform another safety culture assessment to obtain 

more current information on the site’s safety culture.

3. The lead SCA should obtain documents and information needed to 

support evaluation of the licensee’s third-party safety culture assessment 

from the licensee. The lead SCA should coordinate with the licensee to 

schedule interviews with the personnel who performed the assessment 
Issue Date: 02/09/11 17 95003

and licensee staff and managers responsible for implementing actions 

taken in response to the assessment. 

4. Obtain information on any safety culture assessments conducted by the 

licensee within the past five years.

b. Evaluation

The lead SCA and the other SCAs, as assigned, should:

1. Review the documents relating to the licensee=s third-party safety culture 

assessment conducted in response to being placed in the multiple/ 

repetitive degraded cornerstone column of the ROP Action Matrix to 

obtain a general understanding of how the assessment was conducted, 

what the assessment results were, and how the licensee responded.

2. Verify that the assessment was comprehensive, as follows:

(a) The assessment addressed all functional groups within the 

licensee=s organization, including, but not limited to, the functional 

groups that have a clear nexus to safe plant operations (e.g., 

operations, engineering, maintenance, security) and individuals from 

any contract organizations performing those functions;

(b) The assessment included all levels of management responsible for 

overall safe operation of the plant(s), up to and including corporate 

senior management;

(c) Sample sizes were sufficient to ensure that assessment results were 

representative of the populations and sub-populations addressed in 

the assessment; and

(d) Information was collected relating to all of the safety culture 

components.

Specifically note any safety culture component(s) where no 

information was collected within the scope of the licensee=s 

assessment. If any safety culture components were not addressed, 

review any justifications for not assessing the specific component(s) 

of safety culture.

3. Review the methods used by the licensee’s third-party safety culture

assessment team to collect and analyze data for adequacy and 

appropriateness.

4. Verify that the licensee=s third-party safety culture assessment team 

members were not employees of the plant or utility operator(s) of the 

plant. Review their qualifications to determine whether they were 
Issue Date: 02/09/11 18 95003

appropriately qualified to implement the tasks they performed and in 

conducting safety culture assessments overall.

5. Perform a detailed review of the results of the licensee=s third-party safety 

culture assessment to determine whether:

(a) The results drawn from the assessment were consistent with the 

data collected;

(b) The overall conclusions drawn from the assessment were consistent 

with the stated results; and

(c) If any substantial differences exist between results from the 

assessment and the results of similar assessments performed within 

the previous five years, the reason(s) for those differences are 

known and explained.

02.08 Determine Scope of and Plan for NRC Graded Safety Culture Assessment.

The lead SCA should:

a. Determine the scope of NRC’s graded safety culture assessment, based on 

results of the evaluation of the licensee’s third-party safety culture assessment 

in section 02.07, in consultation with the team leader, assistant team leader, 

Regional and program office management, and others as appropriate. 

Depending on the circumstances, the scope of the graded safety culture 

assessment may range from focusing on functional groups which the licensee’s 

third-party assessment identified as having problems/weakness or insufficiently 

evaluated, or performing an assessment of specific safety culture components, 

to conducting an NRC independent safety culture assessment. 

b. Determine the methods best suited for the graded safety culture assessment. 

Prepare the selected data collection tools, such as interview and focus group 

guides and behavioral observation checklists. Coordinate with the other 

inspection team members to determine how to obtain data from their focus 

areas to support the safety culture activities.

c. Identify the resource needs for conducting the graded safety culture 

assessment. Hold meetings with SCAs/inspectors to provide training, briefings, 

assignments, guidance, and other relevant information as needed. Establish a 

plan for communication and coordination among SCAs/inspectors during the 

conduct of the inspection to share data and other information.

d. Follow the guidance in section 1.i. in Attachment 95003.02 to develop and work 

with the licensee to disseminate a communication plan to site personnel 

regarding the NRC’s specific graded assessment activities for their site.

02.09 Perform NRC’s Graded Safety Culture Assessment.

The lead SCA and the other SCAs, as assigned, should:

a. Conduct the graded safety culture assessment based on the scope determined 

and using the tools developed from section 02.08.
Issue Date: 02/09/11 19 95003

b. Coordinate with the other inspection team members to gather insights on safety 

culture components that are part of their inspection focus areas. Participate in 

discussions with the team leader/assistant team leader/other inspection area 

leads to synthesize observations and insights and develop findings and 

conclusions. Interact with team members and group leads to assess the causes 

and contributors leading to the degraded performance in the affected SPA(s).

02.10 Performance Deficiency Cause Analysis.

Review and validate the licensee’s root cause evaluation of the risk significant 

performance issue(s). Evaluate the causes of the performance deficiencies identified 

during the inspection.

02.11 NRC Assessment.

Compare the team=s findings with previous performance indicator and inspection 

program data to determine whether sufficient warning was provided to identify a 

significant reduction in safety. Evaluate whether the NRC assessment process 

appropriately characterized licensee performance based on previous information. The 

findings from this inspection requirement will not be contained in the inspection report 

associated with this inspection, but should be documented in a separate report, coaddressed to the appropriate Regional Administrator and the Director of NRR. 

02.12 Document Inspection Results.

Assess licensee performance in the affected Strategic Performance Area by considering 

the performance deficiencies, results of the inspections described above (including 

related observations and findings), and the need for any follow-up inspections. 

Document the inspection results in a single inspection report.

95003-03 INSPECTION GUIDANCE

General Guidance.

This procedure provides a framework for conducting a comprehensive assessment of 

licensee performance in affected strategic performance areas. As such, the procedure 

is broad in scope, but is designed to allow focus in certain areas where performance 

concerns have already been identified. While some inspection should be performed for 

each key attribute, certain inspection guidance is only applicable if problems are 

identified in that area. 

In order to consolidate inspection activities, the team leader may decide to include a 

continuous main control room observation as part of the inspection. The results from 

the main control room continuous observation should satisfy several inspection 

requirements for the key attributes of configuration control, equipment performance, 

human performance, and procedure quality.

Consideration should be given to having a subteam of the inspection team members 

perform an inspection effort during a plant outage window. One situation when an 

outage inspection can be considered is when the finding(s) that resulted in the 

degraded cornerstone or their underlying cause(s) are related to outage activities and 

that cannot be assessed during non-outage periods. The goal of this inspection is to be 

able to evaluate how well the licensee controls outage activities, how well they control 

the shutdown risk, how well they operate the plant during shutdown and startup 
Issue Date: 02/09/11 20 95003

including the quality of the associated plant procedures, and how well the licensee 

implements plant modifications and maintenance activities.

The team leader should ensure that all team members receive "just in time" training on 

IP 95003 processes and methods. This training should focus on unique aspects of the 

95003 inspection. Typical aspects to cover include: site performance issues, a debrief 

by the senior resident inspector including site specific terminology, interface aspects 

between the 95003 inspectors and SCAs, overview of the NRC’s independent safety 

culture assessment, and administrative details. To coordinate this training, team 

leaders should contact the Branch Chief of the Performance Assessment Branch of the 

Division of Inspection and Regional Support of the Office of Nuclear Reactor 

Regulation.

Team Staffing.

The inspection team shall be staffed with a team leader, primarily inspectors from other 

regional offices and/or headquarters and qualified SCAs. This provides sufficient 

diversity of talent and experience and knowledge, and also adds a degree of 

independence to the overall effort. The team leader selected to perform this inspection 

should have extensive experience in conducting NRC team inspections. Also, the 

inspection team should be staffed with an assistant team leader (ATL). 

Duties and responsibilities for team members are as follows:

The team leader should ensure that an appropriate balance is maintained between 

determining the depth of previously identified issues and determining the breadth of 

performance issues within the strategic performance area. Additionally, the team leader 

should plan and manage the inspection and provide oversight for the safety culture 

assessment activities; including, coordinating all interfaces between the inspection team 

and licensee personnel, NRC management, and public officials. 

The ATL duties and responsibilities should (1) mirror those of the team leader and (2) 

include the majority of the administrative tasks, and planning and managing safety 

culture assessment activities [in coordination with the lead SCA]. The 95003 inspection 

is a demanding effort, and the team leader should have flexibility to respond to 

emergent demands for briefing NRC management and public officials as well as 

maintaining overall cognizance of the inspection effort. An ATL would also aid in the 

freeing up valuable time for the team leader to effectively accomplish these duties. 

It is also desirable to staff the inspection with at least one inspector who has detailed 

knowledge of the site/plant layout. Consideration should be given to using the assigned 

resident staff or another inspector who has recently served as a resident at the site. 

The SCAs with experience and/or specialized training in safety culture assessment 

assigned to the team will solely focus on the safety culture activities. The number of 

SCAs will depend on the scope of the NRC graded safety culture assessment activities.

At least one senior reactor analyst (SRA) should be assigned full time to the team. The 

SRA assigned to this team and other risk experts as appropriate should conduct a 

detailed assessment of the individual and collective risk associated with team=s findings.

The use of contractor support should be considered for conducting aspects of the 

system design reviews, for help in reviewing the licensee=s business and strategic 

plans, and for assistance in completing the safety culture assessment activities. The 

statement of work associated with contractor efforts should specifically include 

provisions for weekend travel for contractors as well as funding for review and 

concurrence on the final report.
Issue Date: 02/09/11 21 95003

A Ateam manager@ should also be designated for the 95003 effort. Ideally, the team 

manager should be based in the sponsoring region and should be an SES-level 

manager. The role of the team manager is to coordinate important senior management 

briefings and interface with other Commission offices and external stakeholders as 

necessary during the conduct of the inspection. Additionally, the team manager is 

responsible for coordinating the acquisition of additional resources as necessary to 

support the overall effort.

Qualification Requirements for Safety Culture Assessors (SCAs)

The team leader should coordinate with the program office to identify the appropriate 

staff to function as the SCA subteam. It is important for the lead SCA to have formal 

training in the social/behavioral sciences and experience in conducting organizational 

assessment activities. Additionally, the lead SCA should have the ability to perform 

group lead functions, such as planning and directing activities, supervising the SCAs 

and other inspection team members, and communicating/coordinating with inspection 

team members/leads and internal/external stakeholders. In cases where staff meeting 

both of these criteria are not available, the use of a contractor who has the necessary 

education and experience background to perform the lead functions, with the exception 

of presenting official NRC positions, can be considered. In such cases, the contractor 

serves as the technical lead and should work with an NRC staff person who has 

leadership experience in a co-lead capacity for coordinating interfaces between 

inspection team and licensee personnel, NRC management, and public officials.

The lead SCA, in coordination with the team leader, should verify that the SCA subteam

collectively has the appropriate credentials (e.g., through education and experience)

that ensure knowledge, skills, and abilities in the following areas:

$ Knowledge of appropriate methods for gathering safety culture data and their 

strengths and weaknesses, including: (1) individual and group interviews, (2) 

structured and unstructured interviews, (3) surveys, (4) behavioral observations 

and checklists, and (5) case studies;

$ Ability to determine the applicability and likely usefulness of various datagathering methods under different circumstances;

$ Ability to implement the different methods correctly, including, but not limited to 

(1) conducting focus groups and interviews in a manner that elicits the desired 

information while reducing potential biases in the responses, (2) conducting 

reliable (i.e., repeatable) structured behavioral observations, and (3) collecting 

insights from written documentation and verbal communications;

$ Knowledge of the requirements for developing, administering, and analyzing the 

results of surveys and questionnaires, including: (1) strengths and weaknesses 

of different item types (Likert, BARS, forced-choice, etc.); (2) requirements for 

administering a survey to reduce potential biases in the responses; (3) 

behavioral statistics and the appropriate methods, and their constraints, for 

analyzing survey data; and (4) statistical requirements for the different types of 

validity and reliability, and appropriate techniques to assess/measure/establish 

them;
Issue Date: 02/09/11 22 95003

$ Knowledge of the rationale for a multiple-measures approach and ability to 

assess the limitations of a single-method safety culture assessment;

$ Knowledge of statistical and conceptual constraints on determining appropriate 

sample sizes for each method;

$ Knowledge of the alternatives for selecting samples for the assessment and the 

biases introduced by different sample selection strategies;

$ Knowledge of theories and research in organizational and human behavior;

$ Ability to integrate results from applying the different methods to arrive at 

defensible conclusions;

$ Knowledge of the ROP and applicable inspection requirements and techniques;

and

$ Knowledge of theory and research in safety culture.

The background of the selected SCAs should be evaluated promptly by the lead SCA to 

identify any training needs. The selected SCAs should complete the identified training 

before participating in inspection activities.

Inspection Planning and Logistics.

The decision to perform this inspection is based on the action matrix. Based on the 

documented performance issues and the guidance contained in this procedure, the 

team leader should develop an outline for a customized inspection plan which should 

describe the overall scope of the inspection, team member assignments, scheduling 

information, etc. The team leader should then notify the licensee of the inspection dates 

and scope, and provide the licensee a list of requested documents that the team will 

need for its initial in-office review. Once the licensee has been notified, the licensee 

should formally acknowledge the readiness for the inspection and that the root cause 

analysis and the third party safety culture assessment are typically completed. 

Prior to the start of the inspection, the team leader should also establish with the 

licensee an agreed upon method for tracking NRC information requests and potential 

issues (findings) that arise during the inspection. The NRC team should not provide 

written documentation to the licensee during the inspection, but rather, should ensure 

that both the team and the licensee have a common understanding of the developing 

issues, throughout the inspection. The joint use of a licensee developed and controlled 

issue tracking list is highly encouraged. 

Depending upon the site-specific circumstances, flexibility is provided to implement this 

procedure in a number of different ways. The timing and scope of the inspection should 

be aligned with the NRC’s understanding of the site performance issues. If a plant has 

transitioned into Multiple Repetitive/Degraded Cornerstone column in a gradual manner, 
Issue Date: 02/09/11 23 95003

the NRC will have a much clearer understanding of the plant issues and the timing of 

the inspection can await completion of the licensee root cause evaluations and safety 

culture assessments. For unique situations where a licensee has entered the Multiple 

Repetitive/Degraded Cornerstone column of the action matrix in a prompt manner 

resulting from a single Red finding, it may be prudent to schedule an early 

implementation of focused aspects of the IP95003 in order to diagnose the scope of the 

site issues in a timely manner. 

Considerations include the benefit to conduct a sequential set of focused functional area 

inspections as part of the overall 95003 effort. This could include scheduling a subgroup to perform an inspection during a plant outage (rather that having the entire team 

on-site during the outage); and scheduling NRC safety culture assessment activities to 

engage with the licensee for the planning evolutions of the third-party safety culture 

assessment and to observe the conduct of the third party safety culture assessment. 

The team manager and team leader need to be aware of the potential that a number of 

discrete functional area inspections may dilute the effectiveness of the team. If the 

entire team is on-site concurrently, they can assess the plant performance in a more 

holistic manner. If the option is elected to conduct focused functional area inspections, 

one of the SCAs should accompany each inspection group to facilitate the integrated 

assessment of the team’s observations and findings to the safety culture assessment 

activities perspective. 

The team should prepare for the inspection at a location determined by the team leader. 

During this time, the team members should provide input into the inspection plan for 

their assigned areas and should provide input to a list of any other documentation that 

will be required for review on-site. All samples selected by team members for 

inspection focus shall be coordinated with and approved by the team leader as part of 

the inspection plan. This preparation phase of the inspection should normally last one to 

two weeks. 

When the inspection is conducted with the full team, the on-site portion of the inspection 

should generally consist of two weeks on site, one or two weeks offsite, and a final 

week on-site. A final debrief should be provided to the licensee on the last day of the 

on-site inspection. A public exit meeting should be held approximately three weeks 

after completion of inspection. All team members should attend the final de-brief.

When planning for the inspection, to the extent possible, the graded safety culture 

assessment activities should be completed concurrent with the other parts of the 

inspection, for the following reasons:

a. As inspectors complete the subject inspections, they will be expected to compile 

observations that will be used in the graded safety culture assessment activities.

b. As safety culture assessment team members identify issues related to the 

subject inspections, the SCAs should inform the inspectors, so the inspectors 

may follow-up on those issues during their inspections.

c. As inspectors identify issues and make observations that have safety culture 

implications, the inspectors should inform the SCAs, so the SCAs may redirect 
Issue Date: 02/09/11 24 95003

or redeploy assessment resources to address those issues and/or incorporate 

those observations.

The team leader should therefore ensure that effective communication channels will 

exist between inspectors and SCAs responsible for completing the activities described 

above.

On a parallel path, identify documents to complete the assessment of the affected 

Strategic Performance Areas. If an evaluation of the Emergency Preparedness area 

will be performed using Attachment 95003.01, then also include the documents required 

to complete attachment 95003.01

Specific Guidance.

03.01 Strategic Performance Area(s) Identification. No additional guidance provided.

03.02 Review of Licensee Control Systems for Identifying, Assessing, and Correcting 

Performance Deficiencies.

a. The inspector should evaluate whether licensee evaluations into significant 

deficiencies are of a depth commensurate with the significance of the issue. 

Evaluations should ensure that the root and contributing causes of risk 

significant deficiencies are identified. Corrective actions should be taken to 

correct the immediate problems and to prevent recurrence. Include in the 

sample to be reviewed the licensee=s evaluations associated with Awhite@ or 

greater performance indicators and inspection findings that were not been 

previously inspected. Use the guidance contained in supplemental IP 95001 to 

help in evaluating the adequacy of the licensee=s evaluations.

 

To the extent possible, include in the sample licensee evaluations and 

assessments associated with programmatic performance issues and 

organization deficiencies, as well as those related to specific hardware issues. 

Consider the results of NRC=s evaluation of licensee root causes performed 

during IP 71152 AProblem Identification and Resolution.‖

b. Line organization, quality assurance, and external audits and assessments 

should be reviewed to determine whether the licensee has demonstrated the 

capability to identify performance issues before they result in actual events of 

undesired consequence. The findings of these audits and assessments should 

be integrated with more quantitative performance metrics and compared to 

those findings identified during this and other NRC inspections. Management 

systems should be in place to process and act upon this performance data as 

appropriate. The inspector should evaluate management=s support to the audit 

and assessment process, as evidenced by staffing of the quality organization, 

responsiveness to audit and assessment findings, and contributions of the 

quality organization to improvements in licensee activities.

With regard to EP related activities, IP 71114.05 ACorrection of Emergency 

Preparedness Weaknesses,@ contains guidance that may be useful in inspecting 

EP aspects of the PI&R program. 

c. Processes for authorizing modifications and allocating resources for completing 

work should give adequate consideration to safety (risk) and the need for 

abiding by regulatory requirements. The authorization and allocation processes 

should provide for a manageable maintenance backlog and prevent the need for 
Issue Date: 02/09/11 25 95003

multiple work-arounds that could increase the likelihood of an initiating event or 

complicate accident mitigation. 

d. The inspector should ensure that licensee performance goals are not in conflict 

with the actions needed to correct performance issues and are in alignment 

throughout the organization. To complete this requirement, a review should be 

performed of corporate, site, and organizational strategic plans, as well as other 

associated licensee documents.

e. Using the guidance contained in IP 40001, ―Resolution of Employee Concerns,‖ 

perform a limited review of the licensee=s program for the resolution of employee 

concerns. In selecting samples for review, focus on those concerns and 

programs specifically applicable to the strategic performance areas which are 

the subject of this inspection. The intent of this review is to determine: (1) 

whether weaknesses in the employee concerns program have contributed to 

previously identified performance deficiencies; (2) whether additional safety 

issues exist that have not been adequately captured by the corrective action

program; and, (3) whether weaknesses in the employee concerns program have 

resulted in issues associated with the maintenance of a safety conscious work 

environment.

f. No specific guidance provided.

g. The team=s review of licensee industry information programs should be limited to 

those problems that might have contributed to the previously identified 

performance concerns. Determine whether the licensee has adequately 

implemented actions as necessary to address the issue. For example, 

weaknesses in licensee programs to review and assess vendor information may 

have contributed to equipment problems.

03.03 Assessment of Performance in the Reactor Safety Strategic Performance Area.

a. Inspection Preparation

1. No specific guidance provided.

2. System Selection. During the planning process, the team leader should 

select a system(s) based on the plant IPE, past safety system functional 

inspections that may have already been performed on a system by the 

licensee or by other NRC teams, and through review of issues contained 

in the Assessment Action Matrix.

The team should select a number of electrical, mechanical, and 

instrumentation and control components for detailed review. The majority 

of these components should be from the principal system with the 

remainder from support systems which are necessary for successful 

operation of the principal system or from interfacing safety systems served 

by the principal system.

3. No specific guidance provided.
Issue Date: 02/09/11 26 95003

4. There are significant preparation activities associated with the review of 

the corrective actions program and the conduct of Performance-Drills. 

Guidance on those activities is given under the appropriate sections. 

b. Key Attribute - Design 

The design review portion of the inspection should be performed by inspectors 

(or contractors) with extensive nuclear plant design experience. It is also 

important that the inspectors performing the design review have a good 

understanding of integrated plant operations, maintenance, testing, and quality 

assurance so that they are able to relate their findings to the other areas being 

inspected.

The inspectors should focus their review on the system selected in paragraph 

02.03.a.2. Specific supplemental inspection procedures are available for certain 

systems (e.g. service water, electrical, I/C) and should be considered as 

additional guidance for evaluating their functional adequacy. Prior to evaluating 

the selected system, the inspectors should review the design basis documents 

such as calculations and analyses. The review should provide the inspectors an 

understanding of the functional requirements for each system and each active 

component throughout the range of required operating conditions, including 

accident and abnormal conditions. The intent is to focus on the risk significant 

aspects of design that could contribute to an increased frequency of initiating 

events, degradation of mitigation systems, or degradation of barrier integrity. 

The inspection is not intended to be a re-validation of the original system design.

In selecting a sample of modifications to the system to be reviewed, the 

inspectors should concentrate on those modifications with the potential to 

significantly alter the system design and functional capability. The sample 

should include modifications involving vendor supplied products or services 

where practicable, since the licensee=s ability to oversee vendor supplied 

services is an important aspect of design control. Inspectors should consider 

expanding the sample of modifications, if significant problems are found. This 

expansion should consider other similar modifications and should not be limited 

to the initially selected system.

The following inspection guidance covers a comprehensive number of design 

areas. The inspectors should focus their review as necessary to best reflect 

previous performance deficiencies. 

1. No specific guidance provided.

2. For the selected modifications:

(a) Verify that the design and licensing input and output information has 

been properly controlled.

(b) Check the adequacy of design calculations for the selected 

modifications and consider the following when evaluating the 

calculation design parameters of the following components:

(1) For valves: What permissive interlocks are involved? What 

differential pressures will exist when the valve strokes? Will 

the valve be repositioned during the course of the event? 
Issue Date: 02/09/11 27 95003

What is the source of control and indication power? What 

control logic is involved? What manual actions are required to 

back up and restore a degraded function? Are the valves 

subject to pressure locking? Do the valves fail to their safety 

position? Are the valves addressed in emergency or 

abnormal operating procedures?

(2) For pumps: What are the flow paths the pump will experience 

during accident scenarios? Do the flow paths change? What 

permissive interlock and control logic applies? How is the 

pump controlled during accident conditions? What manual 

actions are required to back up and restore a degraded 

function? What suction and discharge pressures can the pump 

be expected to experience during accident conditions? What 

is the motive power for the pump during all conditions? Do 

vendor data and specifications support sustained operations 

at low and high flows?

(3) For instrumentation and automatic controls: What plant 

parameters are used as inputs to the initiation and control 

system? Is operator intervention required in certain scenarios? 

Are the range and accuracy of instrumentation adequate? 

What is the extent of surveillance and calibrations of such 

instrumentation? What are the power sources during blackout 

conditions?

(c) Compare the as-built design with the current design basis and the 

licensing requirements for the selected system and consider the 

following questions:

(1) Verify that the modification does not invalidate assumptions 

made as part of the original design and the accident analyses, 

including interfaces with supporting systems. For example, 

are service water flow capacities sufficient with the minimum 

number of pumps available under accident conditions? Are the 

voltage studies accurate and will the required motor operated 

valves (MOVs) and relays operate under end-of-life battery 

conditions and degraded grid voltages? Are fuses and thermal 

overloads properly sized? Are current dc loads within the 

capacity of the station batteries? Is the instrumentation 

adequate in range and accessibility for operations to control 

the system under normal and abnormal conditions? Are 

maintenance frequencies sufficient to maintain the equipment 

within the range of acceptable operating parameters such as 

motor operated valve friction factors? Are test results for the 

system consistent with the design assumptions?

(2) Does the modification invalidate design input parameters 

provided to accident analyses vendors? 
Issue Date: 02/09/11 28 95003

(3) Have modified structures surrounding safety equipment, 

components, or structures been evaluated for seismic 

2-over-1 considerations? Have modified equipment or 

components under the scope of 10 CFR 50.49 been 

thoroughly evaluated for environmental equipment 

qualification considerations such as temperature, radiation, 

and humidity?

(d) Verify whether the selected modifications have introduced an 

unreviewed safety question.

(e) For the selected system, review recent changes to maintenance 

procedures and operating procedures to confirm that the changes 

have not introduced new design parameters or changed current 

design parameters. Confirm that any such design changes have 

been subjected to the formal design change process (e.g. 50.59 

review).

Examples of potential inadvertent design changes follow:

(1) changing maintenance/surveillance procedures to tighten the 

packing on the main steam non-return check valves such that 

they are no longer free-swinging gravity-closing valves;

(2) changing emergency operating procedures to require that 

operators immediately throttle auxiliary feedwater following a 

reactor trip to prevent pump runout/failure that could otherwise 

occur during a main steam line break.

(f) Ensure that verification and validation of computer programs used 

for design and for monitoring of important safety features has been 

adequately accomplished.

3. Consistency between system design and operation.

(a) Verify that training programs are consistent with the current design.

(b) Verify that operator actions can be performed in the required timeframe to mitigate design basis events. Verify that any changes to 

operator actions resulting from system modification(s) have been 

subjected to a safety evaluation and are consistent with the UFSAR 

including the accident analyses.

(1) Was reliance on the operator actions approved by the NRC? 

(2) Is there reasonable assurance that, under all anticipated 

circumstances (e.g. lighting, ambient temperature, radiation 

levels) operators can perform the actions within the times 

assumed in the accident analyses?
Issue Date: 02/09/11 29 95003

4. Evaluation of communications affecting design control.

(a) Assess the ability to communicate accurate information on the 

status of system modifications. Plant policies on updating design 

related material such as the UFSAR may not support timely 

documentation of changes to the system. Verify that provisions are 

in place and being followed to assure the accurate recording of the 

as-designed and as-built conditions during the interim period 

between modification implementation and incorporation into the 

plant design basis documents.

(b) Verify that operations involve engineering in determining the 

operability of degraded safety systems and components (SSC=s).

(c) Verify that operations, engineering, maintenance, and affected plant 

support groups are involved in the evaluation and concurrence 

process for approving:

(1) performance of non-routine maintenance activities

(2) temporary modifications

(3) field change requests

(d) Review the licensee=s control of vendor supplied services and 

products including the evaluation for technical adequacy and quality 

assurance. The licensee=s evaluation and control of vendor supplied 

services and products should be multi-disciplinary in its approach, 

including operations, engineering, maintenance, and the affected 

plant support groups.

(e) Verify that self-revealing deficiencies and those identified by the 

licensee=s vendor control process are properly communicated to the 

vendor.

c. Key Attribute - Human Performance.

1. Using data from the licensee=s corrective action program, LERs, and 

audits, determine if human performance issues have contributed to 

performance issues. Evaluate the overall effectiveness of human 

performance corrective action commitments. Determine if the problems 

were reviewed by the appropriate level of management and prioritized 

according to their safety significance. Evaluate whether the corrective 

actions were technically correct and implemented in a timely manner.

2. Review the following human performance components, as related to the 

previously identified human performance issues.

(a) Work Control
Issue Date: 02/09/11 30 95003

(1) For operations, assess whether: 

(aa) The turnover environment is adequate for clear

communication; 

(bb) On-coming operators are walking down panels with 

current operators or independently; 

(cc) The turnover process is proceduralized and procedures 

are being followed;

(dd) Necessary plant status information is identified, and 

equipment/ operational problems are discussed in 

enough detail for the oncoming shift to understand. After 

turnovers, verify that the operators have sufficient 

knowledge of the plant conditions and activities in 

progress.

(ee) Review the licensee=s administrative procedure for the 

shift supervisor=s conduct and duties. Verify that shift 

command and control is maintained.

Inspectors should try to observe at least two different shifts, 

including a back-shift.

(2) For on-line maintenance work windows, complex surveillance 

and tests, verify that the activities are coordinated with the 

control room, the shift supervision is maintaining effective 

control of plant operations, and the control room is 

implementing the compensatory measures required by the 

risk/safety evaluation. Observe pre-evolution briefings and 

communication between operations and other disciplines to 

verify that effect on safety and risk is being considered.

(3) Review a number of scheduled and non-scheduled 

maintenance activities. Question the control room operators 

to determine their awareness of ongoing activities that could 

affect plant operations, and the priorities in resolving plant 

issues and equipment problems. The intent here is for the 

inspector to verify that control room personnel are 

appropriately aware of ongoing activities, such as 

maintenance, surveillance and testing, plant equipment taken 

out of service, and their impact on plant operation; and are 

implementing the necessary actions. 

(4) Perform a tour of the plant and note indications of operator 

work-arounds or conditions that might require work-arounds 

including:

(aa) Unapproved job aids or marking; 

(bb) Equipment that is not performing as designed;
Issue Date: 02/09/11 31 95003

(cc) The potential for adverse environmental condition(s), 

e.g., insulation removed from high energy lines, doors left 

open that are required for area isolation during a high 

energy line break in an adjacent area, and open doors 

that may render blowout panels and back-draft dampers 

inoperable.

(5) The inspector should review a sample of written logs and shift 

status reports or updates to verify that they:

(aa) Provide sufficient detail to allow a full understanding of 

operationally significant matters, including abnormal 

occurrences or test results and any compensatory 

measures taken;

(bb) Describe changes in plant or equipment status.

(6) Human-System Interfaces including work area design and 

environmental conditions.

(aa) Using the guidance contained in IP 71841, AHuman 

Performance,@ perform a review of identified problem 

areas.

(bb) As necessary, if specific problem areas are identified the 

inspector should:

(i) walk down several control panels to evaluate the 

size, shape, location, function or content of 

displays, controls, and alarms; 

(ii) evaluate work areas for accessibility of equipment, 

equipment layout, emergency equipment location, 

including location of remote panels;

(iii) evaluate the impact of environmental conditions on 

human performance.

(7) An evaluation should be performed to assess whether 

communications between departments and licensee 

management provide information needed for continued safe 

plant operation. Included should be:

(aa) An evaluation of the responsiveness and timeliness to 

requests for assistance and problem resolution; 

(bb) An evaluation as to whether other departments are aware 

of the extent and significance of deficiencies that crosscut organizational boundaries.

(b) Decision-making - For identified areas of human performance 

problems, assess whether the following decision-making practices 
Issue Date: 02/09/11 32 95003

support human performance while observing control room and local 

operations and other work activities:

(1) The roles and authorities of personnel are clearly defined and 

understood.

(2) Operational decisions and their bases are communicated.

(3) Interdisciplinary input and reviews of safety-significant or risksignificant decisions are sought.

(4) Decision-making is systematic when personnel are faced with 

uncertain or unexpected plant conditions.

(5) Conservative assumptions are used and possible unintended 

consequences are considered.

(c) Work Practices - Assess whether personnel work practices support 

human performance. 

(1) Observe operators perform evolutions, tests, and response to 

annunciators, if possible. Evaluate whether the evolution was 

performed in accordance with approved directives and night 

orders, if applicable. Directives and night orders are often 

issued by plant management, and disciplines such as 

chemistry, reactor engineering, and systems engineering.

(2) Observe routine activities of licensed and non-licensed 

personnel. 

(aa) Verify that procedural requirements are being met and 

that procedures are implemented using the correct level 

of use (i.e. continuous, reference, etc.). 

(bb) Determine whether deficiencies are resolved using the 

corrective action program rather than implementing their 

own work-arounds. 

(cc) If possible, during evolutions, tests, and response to 

annunciators, determine whether operator actions or 

compensatory measures were required due to degraded 

equipment of plant conditions, resulting in an operator 

work-around. 

(dd) Determine that human error prevention techniques, such 

as holding pre-job briefings, self and peer checking, and 

proper documentation of activities, are used 

commensurate with the risk of the assigned task, such 

that work activities are performed safely.

(ee) Determine that supervisory and management oversight of 

work activities, including contractors, is effective.
Issue Date: 02/09/11 33 95003

(ff) Determine that personnel do not proceed in the face of 

uncertainty or unexpected circumstances.

(gg) Determine whether these individuals are knowledgeable 

about the current status of SSCs and equipment 

performance and understand the impact of ongoing work 

activities.

(3) Assess the quality of communications by observing whether:

(aa) Communications are consistent with licensee procedures 

during the conduct of operations, maintenance and 

testing activities;

(bb) Instructions or information disseminated using the plant=s 

phone and paging systems are clearly and concisely 

communicated;

(cc) Personnel inform the appropriate level of management of 

any abnormal conditions or significant changes in plant 

equipment and systems.

(4) TS and/or procedure prerequisites are satisfied before 

procedures are executed.

(5) Assess whether the operators exhibit attentiveness and are 

pro-active in assessing plant conditions that may indicate a 

safety concern; 

(d) Resources - Assess that personnel, equipment, procedures, and 

other resources are available and adequate to assure nuclear 

safety.

(1) For identified areas of human performance problems, verify 

that training and personnel qualifications are adequate and 

appropriate for the level of work being performed.

(aa) If possible, observe classroom training and work in 

progress using the checklists of NUREG-1220, Training 

Review Criteria and Procedures, Rev.1.

(bb) Using the guidance in IP 41500, ―Training and 

Qualification Effectiveness,‖ perform a limited review of 

training problem areas. If necessary, interview trainees, 

supervisors, and instructors using the IP 41500 guidance.

(2) In instances where previous performance issues were related 

to the use of excess overtime perform the following reviews.

(aa) Review the licensee=s process for controlling overtime.
Issue Date: 02/09/11 34 95003

(bb) Interview personnel identified as having worked overtime 

to determine how management ensures that personnel 

are not assigned to safety related duties while in a 

fatigued condition.

(cc) Interview personnel involved in working hours in excess 

of those listed in the plant=s technical specifications (with 

or without approval) to evaluate indications of 

recurrent/routine use of overtime.

(dd) Interview personnel involved in working hours in excess 

of those listed in the plant=s technical specifications to 

determine whether they are willing to report whether they 

or others are fatigued.

(ee) Refer to IP 93002, ―Managing Fatigue‖ for guidance on 

the requirements of 10 CFR 26, Subpart I – Managing 

Fatigue. Assess the need for performing a specific 

section of the procedure based on previous performance 

issues related to fatigue management. 

(3) If applicable, review the control room disabled annunciator 

logs. For selected safety-significant annunciators, question 

the operators as to why annunciators are in alarm conditions, 

what operator response was required by the procedure(s) and 

if taken, if continuously lit annunciator windows prevent 

annunciation of new alarm conditions, and why and how 

annunciators are removed from service. For control room and 

local annunciators that cause operator distractions, determine 

if a controlled process for their removal is in place that 

includes an assessment of operational impact, compensatory 

actions, authorization, and corrective actions for restoration. 

Also, review the alarm summary printout to determine if any 

significant alarms occurred that were not documented in the 

control room logs, and whether the operators were aware of 

and had taken appropriate action. Review of the alarm 

summary printout may lead to important operator performance 

indication during and after a transient.

(4) Review a sampling of work packages to verify that the 

documentation is complete, understandable, and accurate.

(5) If applicable, review inadequate equipment labeling.

(6) If applicable, review inadequate maintenance, surveillance, or 

operating procedures.

3. The guidance for observed Performance-Drills found in IP 82001.01, 

―ERO Performance Drills‖ and IP 82001.02, ―ERO Performance Drills 

Dose Assessment‖ may be used to construct drill scenarios and evaluate 

performance.
Issue Date: 02/09/11 35 95003

(a) Evaluate Performance-Drills with a sample of off-duty shift crews, 

including the Shift Supervisor and appropriate support personnel. 

During the drill evaluate capability to (1) classify hypothetical 

conditions notify local authorities (3) perform dose calculations (4) 

recommend appropriate protective actions. This scope allows the 

assessment of licensee performance in all the RSPS. The 

inspection report should document licensee capability accordingly. 

The distinction between low significance mis-steps and the 

capability to implement the Plan to protect public health and safety 

should be clearly delineated.

(b) A small sample of significant changes to the licensee's emergency 

operating, abnormal operating, emergency response procedures 

and equipment can be examined and discussed with personnel to 

determine whether they are aware of the changes, understand them 

and have received training appropriate for their use.

(c) It should be noted that there is no intent to inspect the licensee=s 

ability to critique the performance-drills. The inspection is performed 

to verify the licensee=s ability to implement the Emergency Plan, not 

verify the ability to critique drills as is done under the baseline IP. 

As such, poor performance should be documented as observations 

under Ascope@ in the EP section of the inspection report. Corrective 

action program identification numbers may be included in the report 

to facilitate verification of correction during future inspections. 

d. Key Attribute - Procedure Quality.

1. Evaluate to what extent procedure quality has contributed to previously 

identified performance issues. In performing this evaluation, select a 

sample of procedures which reflect instances where problems with 

procedures have been documented in LERs, NRC inspection reports, or 

licensee assessments or audits. Focus on the technical adequacy of the 

procedures using the following guidance as applicable. Evaluate the 

licensee=s actions to address the procedure inadequacies.

2. Development and review of procedures. 

(a) When reviewing procedures, the inspector should assess the 

technical adequacy of the procedures and determine if the 

procedural steps will achieve required system performance for 

normal, abnormal, remote shutdown, and emergency conditions. 

The inspectors should determine if the system is operated in 

accordance with the system design.

(b) Determine whether the procedures will accomplish the activity within 

the design characteristics and regulatory requirements. During this 

evaluation, the review may include technical specifications, limiting 

condition for operation, UFSAR descriptions, vendor manuals, 
Issue Date: 02/09/11 36 95003

design information, piping and instrumentation drawings (P&IDs), 

and instrumentation and electrical wiring and control diagrams. 

(c) Review maintenance procedures for technical adequacy. Determine 

if the procedures are sufficient to perform the maintenance task and 

provide for identification and evaluation of equipment and work 

deficiencies. Verify the use of quality verification holdpoints for 

independent verification of important attributes. Check the procedure 

content against the vendor manuals to verify that the procedure 

satisfies the vendor requirements for maintaining the equipment in 

proper working order. Verify that important vendor manuals are 

complete and up-to-date. Documents, such as vendor manuals, 

equipment operating and maintenance instructions, or approved 

drawings with acceptance criteria, may by reference be part of a 

procedure. If these documents are so used, the documents (or 

applicable portions) require the same level of review and approval 

as the procedure that references it.

(d) If the technical adequacy of procedures is a concern review the 

following.

(1) Review a sufficient number of procedures to provide 

assurance that the procedures (including checklists, and 

related forms) in the plant working files are current.

(2) Verify that personnel have the ability to reference an 

up-to-date and accurate copy of documents. This is necessary 

because the controlled drawings may not be revised, unless 

changes due to modifications are extensive. As an interim 

measure, some utilities have marked-up a controlled set of the 

control room documents to show the design changes. In such 

situations, the inspector should also verify that revisions of the 

controlled documents incorporating the marked-up changes 

are performed in a timely manner following the modification.

(3) Procedure changes should be in accordance with licensee 

processes and regulatory requirements. Verify the adequacy 

of all procedure changes which resulted from recent (within 

the last year) license change(s) or revision(s) to a technical 

specification.

(4) Verify procedure changes are in conformance to 

10 CFR 50.59. This item applies only to changes to 

procedures which are described or summarized in the 

UFSAR, normally a small portion of the procedures in use at 

the facility. General guidance and contrasting examples 

relating to the procedure changes which can be made by the 

licensee are described in NRC Inspection Manual Part 9900, 
Issue Date: 02/09/11 37 95003

"Guidance on 10 CFR 50.59 -- Changes to Facilities, 

Procedures, and Tests (or Experiments)."

(5) Through discussions with personnel and a review of approved 

procedures, determine if skilled craft, engineering, and 

technical support personnel contribute to the development, 

review, and approval of procedures. Are special or complex 

procedures Adry run@ and discussed prior to use?

(6) Incorporating accepted human factors principles about format 

and writing style into procedures increases the likelihood that 

the procedures will be easier to use and follow. Standards for 

format and writing style can usually be found in the licensee's 

writer's guide. Usability should be determined by evaluating 

the degree to which procedures follow the guidance outlined in 

the writer's guide.

(7) When a writer's guide is not available or if the writer's guide is 

in question, procedure usability can be determined by 

evaluating the elements of writing style, format, and 

organization described in IP 42700, APlant Procedures.@

(e) Verify temporary procedures were properly approved and did not 

conflict with technical specifications requirements. Review a sample 

of temporary procedures and temporary procedure changes issued 

during the past year to determine that the approval and subsequent 

review requirements of the technical specifications are being 

followed. Determine whether the licensee has procedural limitations 

on how long a temporary procedure or a temporary procedure 

change can be in effect, and compare this with observed practices. 

Verify that unapproved Aprocedures@ are not instituted by night 

orders, work orders, etc.

(f) Review the method by which the licensee incorporates temporary 

changes to emergency or significant event procedures. The method 

used should not be so complicated as to preclude proper and timely 

operator action during abnormal plant conditions.

(g) IP 42001, "Emergency Operating Procedures," and the NUREGs 

referenced in it provide additional guidance for reviewing, 

developing, implementing, changing and maintaining emergency 

operating procedures. The team leader should consider adding an 

emergency preparedness specialist inspector to the team if a 

detailed review of emergency plan implementing procedures is to be 

conducted.

(h) IP 82001.05 contains guidance for the inspection of emergency plan 

implementing procedures.
Issue Date: 02/09/11 38 95003

3. No specific guidance provided.

e. Key Attribute - Equipment Performance.

1. Corrective actions 

(a) Based on implementation of the maintenance rule, 10 CFR 50.65 

"Requirements for Monitoring the Effectiveness of Maintenance at 

Nuclear Power Plants", the inspectors should evaluate the 

maintenance area by concentrating on performance examples that 

have shown to be a product of poor maintenance programs. 

Performance issues should be identified by the inspectors during the 

review of non-conformance reports, machinery history results, plant 

tours, observation of maintenance work activities, LER reviews, and 

NRC and licensee's assessments. Risk significant SSCs identified 

with poor performance should receive the highest priority. After 

identifying the performance issue, the inspectors should attempt to 

determine its cause and use this performance example as a means 

to establish issues in any of the maintenance related programs. The 

inspectors should also see if the licensee appropriately implemented 

the maintenance rule in correcting the performance issue and 

whether the licensee is maintaining an appropriate balance between 

SSC availability and reliability.

(b) Examples of maintenance program issues include a relatively large 

maintenance work request backlog, related maintenance work not 

being accomplished in accordance with written administrative and 

procedural controls, and not identifying procedures for needed 

changes.

2. Programs and processes for testing 

(a) Determine that effective methods have been implemented for review 

and evaluation of surveillance test/calibration data, including 

procedures for reporting deficiencies, failures, malfunctions, etc., 

identified during the tests/calibrations or inspections with required 

verification of operability.

(b) Review a sample of post-maintenance tests to ensure that the tests 

are adequate to ensure that the equipment has been returned to an 

operable configuration.

(c) Verify that the surveillance test procedure acceptance criteria are 

adequate to demonstrate continued operability.

(d) Verify that the licensee is effectively calibrating instruments that are 

important to safety. The Technical Specifications do not specify 

calibration requirements for some of these instruments, for example: 

boric acid tank temperature; discharge pressures for various 
Issue Date: 02/09/11 39 95003

engineered safety feature pumps; safety injection accumulator level 

and nitrogen cover gas pressure; cooling water flow to containment 

coolers; main steam isolation valve limit switches used to verify 

valve closure time and provide input to reactor protection system.

3. Operational performance of systems and components. Observe any 

maintenance or testing performed on the selected system while the 

inspection team is onsite.

(a) Walk through the system operating procedures and the system 

P&IDs. If any special equipment is required to perform these 

procedures, determine if the equipment is available and in good 

working order. Verify that the knowledge level of operators is 

adequate concerning equipment location and operation.

(b) Conduct interviews with licensee personnel to determine how the 

system is operated. Determine if system operation is consistent with 

the intended safety function.

(c) Determine if the environmental conditions assumed under accident 

conditions are adequate for remote operation of equipment, such as 

expected room temperature, emergency lighting, steam, radiation 

levels, etc.

(d) Review the maintenance program for the selected system to 

determine if the preventive maintenance (PM) requirements are 

adequate and comprehensive. 

(e) Review applicable design documents, vendor manuals, generic 

communications (i.e., Bulletins, Information Notices, Generic Letters, 

and special studies) and verify that the licensee has integrated and 

implemented the applicable items into the maintenance program.

(f) Conduct interviews with personnel to determine what maintenance 

and modifications have been performed. Determine if the 

maintenance and modifications are consistent with the licensing 

basis. 

(g) Determine if engineering input into maintenance activities is at an 

appropriate level to ensure safe and reliable plant operations.

(h) Verify that methods and responsibilities have been designated for 

performing functional testing of structures, systems, or components 

following maintenance work and/or prior to their being returned to 

service.

4. IP 82001.04, ―Facilities and Equipment,‖ contains guidance for the 

inspection of EP related equipment and facilities that may be useful.
Issue Date: 02/09/11 40 95003

5. Review records of decisions regarding actions to address long standing 

issues to determine whether the decisions appropriately and 

conservatively considered safety.

6. Review records of decisions regarding actions to address long standing 

issues to determine whether resource implications were appropriately 

considered and whether inadequate personnel, equipment, or procedures 

contributed to a delay in resolving the issue.

f. Key Attribute - Configuration Control.

1. Select a sample of the corrective action process/PIM issues related to 

configuration control and review the adequacy of the corrective actions 

implemented. Review all operability determinations that have been 

completed on the selected system. 

2. System Walkdown 

(a) For the selected system, obtain current drawings and review the 

associated operating procedures and UFSAR sections. Review the 

licensee's system lineup procedure, system design basis 

documents, and determine whether the documents are consistent 

with the as-built configuration.

Compare system line-up procedures with drawings to ensure that 

they are consistent (e.g., valve positions, installation of blank flanges 

and caps).

(b) Review jumper, lifted lead, and other temporary modification logs. 

Determine (1) if an adequate technical review was performed before 

the plant modification was performed to ensure the absence of 

unreviewed safety questions, and (2) if plant drawings were 

updated, as needed, to reflect the change. The licensee's controls 

for limiting the duration of temporary modifications should be 

reviewed. Assess the role of the plant, system, and design 

engineering groups in the temporary modification process.

(c) Determine if accessible valves in the system flow path are in the 

correct positions by either visual observation of the valve; by flow 

indication; or by stem, local or remote position indication and that 

they are locked or sealed, if appropriate. 

(d) Verify that valves do not exhibit excessive packing or boron leakage, 

missing hand-wheels or bent stems. Ensure that local and remote 

position indications are functional and indicate the same values. 

Remote manual operating devices should be functional.

(e) Verify that pump seals do not show signs of excessive leakage.

(f) Verify that cooling water is aligned to bearings and seals and that 

oil bubblers and bearings do not show signs of excessive leakage.
Issue Date: 02/09/11 41 95003

(g) Verify that power is available and correctly aligned, functional, and 

available for components that must activate on receipt of an initiation 

signal.

(h) Verify that major and support system components are correctly 

labeled, lubricated, cooled, and ventilated to ensure fulfillment of 

their functional requirements.

(i) Review system mechanical joints (packing, flanges, body to bonnet 

joint) leakage requirements and verify that known leakage is 

properly addressed and that observed leaks are accounted for.

(j) Determine if selected instrumentation, essential to system actuation, 

isolation, and performance, is correctly installed and functioning, 

correctly calibrated, and displaying indication consistent with 

expected values. Instrument elevations are consistent with design 

documents.

(k) Identify whether actual or potential adverse environmental 

condition(s) exist, and the adequacy of any compensatory 

measures.

(l) Identify whether components inspected for the system are 

consistent with the UFSAR description. Determine whether a 

10 CFR 50.59 safety evaluation was performed for any items that 

differ from the UFSAR description.

(m) Identify additional equipment conditions and items that might 

degrade plant performance by verifying whether:

(1) Freeze protection, such as insulation, heaters, heat tracing, 

temperature monitoring, and other equipment, is installed and 

operational.

(2) Hangers and supports are in their proper positions, aligned 

correctly, and intact.

(3) No unauthorized ignition sources or flammable materials are 

present in the vicinity of the system being inspected.

(4) Cleanliness is being maintained.

(5) Temporary storage of material and equipment is in 

accordance with the licensee=s seismic control procedures and 

does not interfere with equipment operations or operator 

actions.

3. Maintenance Work Control
Issue Date: 02/09/11 42 95003

(a) Determine the nature and extent of the licensee's backlog of 

corrective and preventive maintenance, especially concerning 

equipment of high safety significance. Assess the licensee's efforts 

to integrate preventive and corrective maintenance to minimize 

equipment unavailability.

(b) Assess the licensee=s process for planning work, including the 

assessment of risk and the inclusion of new emergent work into the 

schedule. Review the licensee=s policies with respect to schedule 

generation and the use of risk insight. Select several work 

packages on safety related equipment and determine how risk was 

factored into their scheduling. Assess whether the licensee 

evaluates possible interactions between components in service and 

those to be tested. Assess whether the need for planned 

contingencies, compensatory actions, and abort criteria were 

considered.

(1) What risk assessment tools are provided to the operators?

(2) What risk training has been given to the planning staff?

(3) Who has the absolute say in allowing work to progress?

(4) How is emergent work factored into previous risk evaluations?

(c) For the selected systems review the operating performance history 

and compare it with the assumed out-of-service times in the IPE. 

Ensure that the assumptions are conservative with respect to actual 

equipment performance.

(d) If warranted as a result of past performance deficiencies, select one 

or more safety system tag-outs for inspection. Determine if the 

tagout is adequate for the work to be accomplished. Verify in the 

plant that operators are thorough in tagging and isolation of plant 

equipment. Verify by observation that tags are properly hung and 

equipment has been placed in the designated position. Determine if 

equipment status changes and corresponding entry into or exit from 

technical specification action statements are appropriately 

documented.

(e) If warranted as a result of past performance deficiencies, determine 

if the licensee has adequate controls to ensure the independent 

verification of equipment status, particularly when equipment is 

returned to service.

(f) Verify that maintenance activities are coordinated with control room 

operations and that appropriate briefings and turnovers are held with 

control room operators.
Issue Date: 02/09/11 43 95003

(g) Equipment that is environmentally qualified should be identified as 

such prior to maintenance and sufficient controls should exist to 

ensure it is returned to that status upon reassembly.

(h) The inspectors should review the following: long-term (typically 

greater than six months) tagouts (caution and danger tags), disabled 

control room annunciators and instruments, control room 

deficiencies, operator work arounds and other equipment deficiency 

tracking systems to assess the significance of these conditions.

(i) If warranted as a result of past performance deficiencies, review the 

licensee=s process for using rapid response maintenance teams.

(j) Verify that work control procedures have been established to require 

special authorization for activities involving welding, open flame, or 

other ignition sources and take cognizance of nearby flammable 

material, cable trays, or critical process equipment. Ensure that 

work control procedures have been established to require a 

firewatch, with capability for communication with the control room, if 

an activity identified above is to be performed in the proximity of 

flammable material, cable trays, or vital process equipment. 

Procedures should address scaffold controls around safety, critical 

or operating equipment.

4. Chemistry Controls - limit reviews to primary and secondary chemistry 

which could degrade the RCS pressure boundary.

(a) Review records of completed chemical analyses to determine if 

required analyses have been performed.

(b) Review trends of recorded water quality data.

(c) Assess corrective actions taken when chemical variables have 

exceeded the established levels or limits, including consideration of 

the timeliness of these actions.

(d) Assess the effectiveness of measures taken to prevent the 

introduction of chemical contaminants into primary and secondary 

coolant water and to detect the presence of these contaminants.

(e) Review licensee evaluations of parameter trends associated with 

steam generator leakage.

5. Fission Product Barrier Assessment

(a) Observe a selected portion of the containment isolation lineup and 

independently verify whether valves, dampers and airlock doors are 
Issue Date: 02/09/11 44 95003

being properly controlled in accordance with the Technical 

Specifications.

Select several components and independently verify that they are in 

their required positions. Where possible, confirm valve position 

indication by direct observation of valve mechanism. For valves that 

isolate on a containment isolation signal verify proper breaker 

position and availability of power supply. Also, for motor and airoperated valves, verify they are not mechanically blocked and power 

is available, unless it is required to be otherwise. Inspect piping and 

the associated test, vent and drain valves, if any, for possible 

leakage paths.

(b) Assess the licensee=s method of calculating the RCS leakrate.

(c) Containment temperature and pressure monitoring - review the 

licensee=s procedures for ensuring that the containment atmosphere 

and/or water space meets the design basis assumptions for average 

temperature and pressure.

g. Key Attribute - Emergency Response Organization Readiness.

The guidance of IP 71114.03 is applicable and may be useful. 

03.04 Assessment of Performance in the Radiation Safety Strategic Performance Area 

- Occupational Radiation Safety

a. Inspection Preparation. 

1. No specific guidance provided.

2. Look particularly for those audits and self-assessments that probe for 

programmatic weaknesses and assess the quality of the program. Look 

for trends indicative of programmatic weaknesses. Requirements for 

reviews and audits normally are contained in the technical specifications. 

Audit teams should include someone with experience or training 

commensurate with the scope, complexity, or special nature of the 

activities audited.

3. No specific guidance provided.

4. No specific guidance provided.

b. Key Attribute - Program/Processes for Occupational Radiation Safety

1. Each position within the RP organization should have its own position 

description with authorities and responsibilities clearly defined. For 

example, each health physics technician (HPT) should know what 

authority should be exercised to ensure the RP program can be effectively 

implemented (e.g., enforce the stoppage of work, adherence to 
Issue Date: 02/09/11 45 95003

procedures). The HPT and the crafts workers should all understand these 

responsibilities and authorities. The inspector should be sensitive to the 

designated radiation protection manager=s position in the facilities 

reporting chain and level in the organization and how this affects the 

RPM=s direct recourse to onsite station manager on problems with the 

conduct of the radiation protection program. The impact of any 

organizational change in the RPM position relative to its level should be 

examined and discussed with appropriate level of management. 

Determine if adequate HPT coverage is being provided during outages 

and normal backshift operations. Determine the extent of first-line 

supervision (foremen) presence in the field -- past lack of foremen having 

direct involvement at the onset of infrequent work activities in high 

radiation areas has contributed to serious mishaps and over exposures.

2. Evaluate to what extent procedure quality has contributed to previously 

identified performance issues. Select a sample of procedures where 

problems with procedures has been documented in LERs, NRC inspection 

reports, or licensee assessments or audits. Focus on the technical 

adequacy and completeness of the procedures using the following 

guidance as applicable.

(a) When reviewing the procedures, the inspector should determine if 

the procedural steps will achieve the required goal. The inspector 

should determine if the procedure is understood and used by the 

HPTs.

(b) Verify that the licensee has a workable system to ensure that the 

plant working files contain current procedures (including checklists 

and related forms).

(c) Procedure changes should be in accordance with licensee 

processes and regulatory requirements. Verify the adequacy of all 

changes (within the last year) in a selected area of concern (e.g., 

RWP issuance).

(d) Through discussions with personnel, determine if HPT and first line 

supervision contribute to the development, review and approval of 

procedures.

(e) Verify temporary procedures were properly approved and did not 

conflict with requirements, by review of a sample of recent 

temporary procedures and revisions to them.

(f) NRC Inspection Manual 9900 provides the NRC position on control 

of procedural adherence.

3. Effective radiation work practices include considerations of high and very 

high radiation areas and awareness of potential hazards (e.g., in diving 
Issue Date: 02/09/11 46 95003

operations, removing neutron-activated items from the reactor, and other 

non-routine and infrequent operations). See Regulatory Guide 8.38, A

Control of Access to High and Very High Radiation Areas in Nuclear 

Power Plants@ for further guidance on these problem areas. Based on 

current licensee work planning, select at least three jobs being performed 

in radiologically challenging areas. Whenever possible, select jobs in 

locked high radiation areas (with >1 rem planned job collective personrem). Additionally, focus on work in Airborne radioactivity areas, with a 

special emphasis on areas were transuranic radionuclides may be 

present. 

(a) Review all pertinent job requirements (RWP, work control 

procedures, etc.), attend job briefing, and observe infield work and 

judge compliance to above requirements.

(b) Determine if the job conditions were adequately communicated to 

the worker, by pre-work briefing and work site postings. 

(c) Verify accuracy of required surveys, HPT job coverage is consistent 

with RWP requirements. Verify that worker dose monitoring is 

consistent with licensee and regulatory requirements. This should 

include the need for extremity and multi-badging for DDE. Improper 

uses of digital alarming dosimeters have resulted from (1) lack of 

training in their proper use, (2) use in high noise areas or under 

protective clothing, which made the alarm inaudible, and (3) poor (or 

no) procedures for their use.

(d) Attend any post-job debriefing to capture any lessons learned 

discussion. Determine how (if) licensee incorporates applicable 

lessons learned into procedures, RWP process, etc.

(e) Review the diving procedure and determine if it meets of intent of 

Regulatory Guide 8.38, Appendices A and B. See Information Notice 

97-68, ALoss of Control of Diver in a Spent Fuel Pool@ for further 

guidance.

(f) Transuranics can be a potential airborne problem at plants with 

previous fuel performance problems (fuel leakers). See Information 

Notice 97-36, AUnplanned Intakes By Worker of Transuranic 

Airborne Radioactive Materials and External Exposure Due to 

Inadequate Control of Work@. Note that while a plant (with a history 

of fuel leakers) may not have seen significant evidence of 

transuranics for years (on loose-contamination smears or routine air 

samples), alpha contamination may be incorporated into a corrosion 

layer on the interior surfaces of system components that carry 

primary system fluids. When these interior surfaces have be 

perturbed (by mechanical actions like scabbling), high levels of 
Issue Date: 02/09/11 47 95003

transuranic airborne activities have resulted in significant, unplanned 

worker intakes. 

(g) Review each planned special exposure to determine whether it 

meets the requirements of 10 CFR 20.1206. See Regulatory Guide 

8.35, "Planned Special Exposures."

(h) Review a selected sample of the records of exposures of declared 

pregnant women to determine whether, in each case, the dose to 

the embryo/fetus meets of the requirements of 10 CFR 20.1208. 

See Regulatory Guide 8.36, "Radiation Dose to the Embryo Fetus."

4. Review the extent to which the licensee has implemented or assessed 

methods offering significant potential for reducing occupational radiation 

exposure by reducing out-of-core radiation sources/fields. The following 

techniques are reported to be available for reducing exposure [See the 

Electric Power Research Institute (EPRI) report TR-107991, "Radiation 

Field Control Manual - 1997 Revision," October 1997.] 

(a) PWRs: Methods available now that can provide an immediate 

impact are (a) chemical decontamination together with a modified 

pH primary chemistry control program (2.2 ppm Li, pH 7.2- 7.4) and 

use of Zircalloy fuel grids, and (b) valve maintenance procedures to 

remove Co debris. Methods available now that will have a slower 

impact are (a) Zircalloy fuel grids without decontamination, (b) 

electropolishing of replacement steam generators, c) cobalt 

replacement guidelines and NOREM valves, (d) use of low-cobalt 

Inconel 690 tubing for replacement steam generators.

(b) BWRs: Methods available now that can provide an immediate 

impact are (a) chemical decontamination together with (1) 

replacement of control blade pins and rollers and (2) zinc injection, 

(b) installation of cobalt-free feedwater control valves, and c) valve

maintenance procedures to remove Co debris. [Note: The use of 

natural zinc injection has resulted in problems at some BWRs. The 

zinc-65 produced by neutron activation of zinc has caused higher 

radiation fields, higher volumes of radioactive waste, and in at least 

one case, surface contamination problems. As of October 1997, 

these problems are minimized by the use of depleted zinc. The 

industry is currently developing the most cost effective approach to 

zinc injection. Methods available now that will have a slower impact 

are (a) pins and rollers replacements and zinc injection without 

decontamination, (b) electropolishing/pre-conditioning replacement 

components, and (c) cobalt replacement guidelines. Methods 

showing promise are (a) replacement of in situ pins and rollers, and 

(b) NOREM cobalt-free hardfacings for valves. 
Issue Date: 02/09/11 48 95003

The techniques above involve cobalt source reduction, 

preconditioning of out-of-core surfaces, control of crud transport 

(water chemistry control), and chemical decontamination.

Licensees should not be expected to implement a method for 

reducing out-of-core radiation sources/fields until the method has 

been fully tested and proven by a full-scale field demonstration in 

one or more nuclear power plants. The term "fully tested and 

proven" means that the technique has been fully scoped and reliable 

generic technical basis documentation is available for the licensees 

to evaluate the potential for their particular plant application.

5. The licensee should have an appropriate basis for establishing dose goals 

and objectives. Goals should be frequently monitored and actions taken 

as necessary when goals are exceeded. Goals should be set for the 

facility as a whole, for different divisions or groups within the facility, and 

for specific work activities. 

Review work tasks to verify that pre- and post-job ALARA reviews were 

conducted. Determine whether the pre-job reviews adequately addressed 

the work to be performed, and whether lessons learned from post-job 

reviews are factored into future work/training. Ensure that the radiological 

significance of work performed under the direction of licensee 

vendors/contractors is adequately reviewed before the work is started. 

Review the method used to perform ALARA reviews of on-going work 

activities. These reviews should identify anomalies in the expected rate at 

which personnel exposure is being accumulated.

Compare, as a minimum, the licensee's total annual collective dose 

(person-rem) against their goals. Determine whether the licensee's 

collective doses are increasing or decreasing. Discuss with the licensee 

reasons for any trends and actions they are taking or have taken that 

impacted the trend. Determine whether the licensee is effective in 

identifying causes of higher than necessary doses and in effecting 

corrective actions. Determine whether the licensee reviews dose 

experience for specific jobs against available industry norms for similar 

jobs.

For plants planning their first outage, or for experienced plants performing 

significant tasks (e.g., 10-year in-service inspection) for the first time, 

determine the extent to which the outage experience of other similar 

plants is being used in the planning process. For plants that have 

experienced outages, determine the extent to which experience from, and 

lessons learned during, previous outages are being incorporated to 

improve performance. Approval of needed visits by radiation protection 

personnel to other sites to observe outage activities is a good indication of 

proper management support for ALARA.
Issue Date: 02/09/11 49 95003

6. Review the licensee's organizational structure for ALARA responsibilities. 

There should be a clear delineation of authority and responsibility, 

including dedicated ALARA staff adequate to implement the program on a 

daily basis as well as during outages. ALARA training that extends beyond 

the scope of General Employee Training for personnel such as radiation 

workers, is desirable for radiation protection technicians, and special 

maintenance teams. Professional development training should be 

available for the ALARA coordinator and related staff. To be most 

effective, mockup training should be reasonably realistic (e.g., including 

realistic temperature, humidity, and lighting) and address ALARA 

considerations.

Discuss the ALARA program with several workers to determine whether 

they understand the program, understand their role in the program, and 

are actively involved in the program.

c. Key Attribute - Plant Facilities/Equipment & Instrumentation – Occupational 

Radiation Safety 

1. No specific guidance

2. Select a variety of equipment and on-going maintenance to observe full 

calibration of beta/gamma survey instruments, as well as the daily 

source/response checks (or prior to use functional checks) for these 

instruments.

Verify that the HPT or maintenance technician is familiar with the 

procedure governing the selected activity. Determine that the HPT is 

following the procedure, and discuss any deviations (and the reasons) 

from the procedure. Be aware of the facilities Aart of the craft@ position -- a 

level of skill and proficiency that is assumed (by the level of qualification). 

This position has a impact on the level of detail of the procedures, and 

allows the HPT, etc. to perform certain tasks or actions without a 

procedure.

3. Ensure that the facility has an adequate supply of materials necessary to 

support current operations and emergent work/special outages. This 

includes anti-c=s, respiratory protection, temporary shielding, temporary 

portable ventilation equipment, personal cooling devices (heat stress) and 

other needed equipment.

Determine if the facility has adequate areas for personal and equipment 

decontamination, equipment maintenance and calibration (including spare 

parts).

Discuss the budgetary process with the RPM and first line supervisors and 

examine and determine reasons for budget item disapprovals for selected 

(rejected items). Focus on those budget items that had been approved by 

the RPM, but not supported by upper management. Determine reasons for 
Issue Date: 02/09/11 50 95003

budget denials for major proposed items, and judge the impact on any 

identified program deficiencies.

. Identify plant areas that have become unusable as a result of an 

operational occurrence and licensee actions to control and recover such 

areas. (See SECY-89-326 dated 10/20/89 located at microfiche address 

70038-056.)

4. Determine whether recommendations for plant improvement appropriately 

considered radiation safety. These considerations include whether: all 

potential impacts of the improvement on radiation safety were considered 

and incidents which negatively impacted radiation safety occurred after a 

decision not to incorporate the recommendation.

d. Key Attribute - Human Performance for Occupational Radiation Safety

The inspector should be aware that worker performance has an obvious, 

important impact on work activities in radiological areas. Two of the major 

components are health physics technicians (HPTs) and general radiation worker 

(crafts) groups. Human performance is impacted by several vital factors B

qualification and training. Selection, qualification and training requirements for 

facility personnel are generally governed by a commitment in the plant technical 

specifications (to an ANS standard). For HPTs and others, 10 CFR 50.120 

(training rule) requires HPTs (including contractors) to be task qualified for their 

assigned normal and outage duties.

1. Using data from the licensee=s corrective action program, LERs, and 

audits, determine if human performance issues have contributed to 

performance deficiencies. Evaluate the effectiveness of corrective actions

by reviewing the corresponding commitments. Determine if the problems 

were reviewed by the appropriate level of management and prioritized 

according to their safety significance. Evaluate whether the corrective 

actions were technically correct, and developed and implemented in a 

timely manner.

2. Review the following components of human performance, as related to the 

previously identified human performance issues.

(a) Work Control - If problem areas and issues were identified by 

inspections with respect to work control, including coordination and 

communication among activities; practices such as pre-job briefings, 

effective communications, and shift turnover; human-system 

interfaces, work area design, and environmental conditions, or 

minimization of work-around, then:

(1) Determine if shift turnover time is sufficient, and that 

appropriate plant/work status/conditions are discussed. 

Determine if the radiation protection log (or night-orders) is 

governed by training or procedure, and whether it is a reliable 
Issue Date: 02/09/11 51 95003

and consistent tool for the HPTs. At the end of an inspection 

shift, attend a radiation protection HPT shift turnover and 

identify any weaknesses or deficiencies in the communication 

exchange. Discuss these with the on-shift management.

(2) As necessary, if specific problems are identified in this area, 

the inspector should:

(aa) Evaluate work areas for accessibility, equipment layout, 

emergency equipment location, power supplies for infield 

sampling, etc.

(bb) Evaluate the impact (and means to compensate) for 

temperature extremes (heat stress), and other industrial 

hygiene hazards that might hamper radiation safety 

performance.

(cc) Observe HPT interaction with crafts during development 

of RWP (on High Radiation area work, if possible). 

(dd) Determine how and when HPTs inform the appropriate 

level of management on any abnormal condition or 

significant changes in work environments.

(ee) Evaluate the use of Engineering and Radiation Protection 

staff support during high risk (dose) work.

(ff) Evaluate the impact (and means to compensate) for 

temperature extremes (heat stress), and other industrial 

hygiene hazards that might hamper radiation safety 

performance.

(3) Assess the quality of communications by observing HPT 

interaction with crafts during development of RWP (on High 

Radiation area work, if possible). 

(b) Decision-making - If problem areas and issues were identified by 

inspections with respect to decision-making, then conduct 

observations of planning activities to determine whether decisionmaking involves contingency planning and use of conservative 

assumptions, and decisions are communicated to affected 

personnel. For identified areas of human performance problems, 

verify that the following decision-making practices support human 

performance while observing control room and local operations and 

other work activities:

(1) The roles and authorities of personnel are clearly defined and 

understood.
Issue Date: 02/09/11 52 95003

(2) Operational decisions and their bases are communicated.

(3) Interdisciplinary input and reviews of safety-significant or risksignificant decisions are sought.

(4) Decision-making is systematic when personnel are faced with 

uncertain or unexpected plant conditions.

(5) Conservative assumptions are used and possible unintended 

consequences are considered.

(c) Work Practices - If work practices, such as peer- and self-checking, 

procedural use and adherence, human error prevention techniques, 

or management and supervisory oversight, were identified as 

problem areas, then conduct in-field observations of work in 

radiological areas, and focus on HPT and worker performance 

relative to required RP work practices. 

(1) Assess the quality of communications by determining how and 

when HPTs inform the appropriate level of management on 

any abnormal condition or significant changes in work 

environments.

(2) Evaluate the use of Engineering and Radiation Protection staff 

support during high risk (dose) work.

(3) Verify that TS and/or procedure prerequisites are satisfied 

before procedures are executed.

(4) Assess whether radiation protection technicians stop work due 

to radiological considerations when appropriate.

(d) Resources - If problem areas or issues were identified by 

inspections with respect to available resources such as sufficient 

trained and qualified personnel to maintain work hours within limits, 

or tools and equipment, then:

(1) Review the licensee=s overtime program and process to 

determine how management ensures that workers are not 

assigned safety related duties while in a fatigued condition. 

Interview workers to determine if they worked hours greater 

than specified in the technical specifications (with or without 

approval) to evaluate any repetitive nature that can lead to a 

degraded performance (See HPPOS #=s 024,173, and 253). 

(2) Interview health physics technicians and other Radiation 

Protection staff that worked hours greater than the plant=s 

technical specifications (with or without approval) to evaluate 
Issue Date: 02/09/11 53 95003

whether these personnel feel free to report whether they or 

others are fatigued. 

(3) Refer to IP 93002, ―Managing Fatigue,‖ for guidance on the 

requirements of 10 CFR 26, Subpart I – Managing Fatigue.

Assess the need for performing a specific section of the 

procedure based on previous performance issues related to 

fatigue management.

3. Experience, qualification, and training of radiation protection staff –

Review the applicable experience, qualification and training of selected 

members of the licensee=s (and its contractor=s) RP organization.

(a) Review the licensee=s program to provide training and periodic 

retraining to plant and contractor personnel on assigned duties and 

on safety significant changes to programs and procedures. 

Determine whether this training includes lessons learned from 

recent industry events and NRC communications (i.e., Information 

Notices, Generic Letters, Administrative Letters, etc.) and the proper 

use of human performance tools. By discussion with selected 

personnel, review of training lesson plans, and completed training 

records, determine if the requirements of 10 CFR 19.12 are met. If 

possible, observe portions of the general employee training (focus 

on the practical aspects of the training).

(b) Review applicable radiation protection worker qualification and 

training of selected members of other facility work units (including 

contractor employees).

(c) Interview several RP personnel, including first-line supervisors, 

professional RP staff, and the designated RP manager. Assess their 

level of knowledge about the program and applicable implementing 

procedures. 

(d) Select individuals on the radiation protection staff and contractor 

personnel. By a review of applicable documentation, direct 

observation and discussion with the technicians, determine if they 

meet the training and qualification requirements of their assigned 

duties/position. Licensee administration technical specifications 

normally contain a specific commitment to an industry standard on 

personnel selection, training and qualification. 

(e) For a selected sample of contractor health physics technicians 

(HPTs), review the actions taken by the licensee, in accordance with 

the training rule (10 CFR 50.120) to ensure that these individuals 

are task qualified to perform their assigned outage activities. The 

following general guidance exists concerning the10 CFR 50.120:
Issue Date: 02/09/11 54 95003

(1) The only radiation protection personnel covered by the new 

rule are "radiological protection technicians" (HPTs) who are 

employees of the power plant. No supervisory, managerial or 

technical staff are covered. Contractor HPTs are not covered 

unless they occupy regular positions performing independently 

within the licensee's organization. If short-term contractor 

HPTs (e.g., outage workers) are assigned to work 

independently, they must be qualified to perform their 

assigned tasks. 

(2) The training rule covers qualification only in the sense of job 

task qualification, not qualification based on pre-selection 

criteria. Furthermore, successful completion of a training 

program required by the rule does not obviate the need to 

comply with other training or qualification requirements 

imposed by other regulations and/or license conditions.

By direct observation and discussion with HPTs providing job 

coverage, determine if they have knowledge of the job 

activities and radiological conditions to provide adequate 

coverage. In discussions with HPTs, focus on ensuring 

adequate knowledge of radiological hazards associated with 

plant systems [especially neutron-activated components such 

as traversing incore probes (TIPs), incore neutron detectors, 

and cabling, as discussed in Information Notice No. 88-63 and 

its Supplements 1 and 2, "High Radiation Hazards from 

Irradiated Incore Detectors and Cables"]. Interviews using 

event scenarios and role-playing (preferably in small groups) 

may be useful for assessing HPT knowledge and capabilities.

Review the licensee's method to provide training of permanent

and contractor personnel on safety- significant changes in 

procedures and recent events. Emphasis should be on 

training provided to the increased work force required for the 

outage. Discuss with plant management and the RPM.

(f) Review training records and lesson plans, for a sampling of station 

and contractor crafts workers, to determine if the requirements of 

10 CFR 19.12 are satisfied. If possible, observe portions of the 

general employee training (focus on the Ahands-on portions@ of the 

training).

4. Interviews should focus on identified program deficiencies, root causes, 

and action plans for improving performance. Discuss how improvements 

will be implemented -- what programmatic changes are needed, how 

these will be accomplished, etc. 
Issue Date: 02/09/11 55 95003

03.05 Assessment of Performance in the Radiation Safety Strategic Performance Area 

- Public Radiation Safety (Radiological Effluent Monitoring, Radioactive Material 

Control, and Transportation of Radioactive Material).

a. Key Attribute - Plant Facilities / Equipment and Instrumentation for Pubic 

Radiation Safety

Perform an extensive tour of the facility which includes interviews with plant and 

contractor personnel to evaluate the adequacy of the plant facilities, equipment 

and instrumentation.

1. No specific guidance provided.

2. Evaluate the physical condition of the facilities, equipment and 

instrumentation. Determine if the facility is appropriate for its intended use 

and adverse conditions (i.e., radiation levels, temperature, lighting, 

industrial hygiene hazards, etc.) that may hamper the performance of the 

workers are minimized.

(a) Verify that equipment and instrumentation are operable, calibrated, 

source checked, and maintained as specified in the licensee=s 

procedures. Where appropriate, verify that the alarm/trip setpoints 

are correctly set to meet the requirements of the technical 

specifications or regulatory requirements.

(b) Review the licensee=s use of computers and software used to 

perform selected tasks. Review the licensee=s technical evaluation 

to ensure the software is appropriate for its intended use. Verify that 

the computer software has been verified and validated.

(c) Perform direct observation of the calibration of selected equipment 

and instruments. Verify that the proper materials, as specified in the 

procedure, are being used to perform the calibration. If radioactive 

sources are being used, are they properly transported, handled, 

used, controlled, and stored in accordance with approved plant 

procedures. Is the process in accordance with the plant=s ALARA 

program?

3. Verify that there is an adequate supply of spare parts and materials 

needed to maintain the equipment and instruments.

b. Key Attribute - Program / Process for Public Radiation Safety

For the assessment of procedures which implement the program being 

inspected, evaluate to what extent procedure quality has contributed to 

previously identified performance issues. In performing this evaluation, select a 

sample of procedures which reflect instances where problems with procedures 

have been documented in LERs, NRC inspection reports, or licensee 

assessments, audits or corrective action programs. The inspectors should focus 

on the technical adequacy of the procedures using the following guidance as 

applicable.
Issue Date: 02/09/11 56 95003

1. No specific guidance provided.

2. Review of procedures

(a) When reviewing the licensee=s procedures, the inspector should 

assess the technical adequacy of the procedures and determine if 

the procedural steps will achieve the required result.

(b) Determine whether the procedures are consistent with the technical 

specifications, program documents, and regulatory requirements. 

During this evaluation, the review may include technical 

specifications, program documents, UFSAR descriptions, vendor 

manuals, design information, and instrumentation diagrams.

(c) If applicable, review maintenance procedures associated with the 

instrumentation and equipment being inspected for technical 

adequacy. Determine if the procedures are adequate to perform the 

maintenance task and provide for identification and evaluation of 

instruments and equipment and work deficiencies. If applicable, 

verify the use of quality verification holdpoints for independent 

verification of important attributes. Check the procedure content 

against the vendor manual to verify that the procedure satisfies the 

vendor requirements. Verify that the vendor manuals are complete 

and up-to-date. Documents, such as vendor manuals, equipment 

operating and maintenance instructions, or approved drawings with 

acceptance criteria, may by reference be part of a procedure. If 

these documents are so used, the documents (or applicable 

portions) require the same level of review and approval as the 

procedure that references it.

(d) Through discussions with personnel and a review of approved 

procedures, determine if radiation protection and technical support 

personnel contribute to the development, review, and approval of 

procedures. Determine if unique and/or complex high radiation work 

procedures are reviewed and approved by personnel responsible for 

work performance.

(e) Incorporating accepted human factors principles on format and 

writing style into procedures increases the likelihood that the 

procedures will be easier to use and follow. Standards for format 

and writing style can usually be found in the licensee=s writer=s 

guide. Usability should be determined by evaluating the degree to 

which procedures follow the guidance outlined in the writer=s guide.

(f) When a writer=s guide is not available or if the writer=s guide is in 

question, procedure usability can be determined by evaluating the 

elements of writing style, format, and organization described in 

IP 42700, APlant Procedures.@
Issue Date: 02/09/11 57 95003

(g) Verify temporary procedures were properly approved and did not 

conflict with technical specification requirements. Review a sample 

of temporary procedure changes issued during the past year to 

determine that the approval and subsequent review requirements of 

the technical specifications are being followed. Determine whether 

the licensee has procedural limitations on how long a temporary 

procedure or a temporary procedure change can be in effect, and 

compare this with observed practices.

3. If the technical adequacy of procedures is a concern, review the following.

(a) Review a sufficient number of procedures to provide assurance that 

the procedures (including checklists and related forms) in the plant 

working files are current.

(b) Verify that personnel have the ability to reference an up-to-date and 

accurate copy of documents. This is necessary because the 

controlled drawings may not be revised, unless changes due to 

modifications are extensive. As an interim measure, some utilities 

have marked-up a controlled set of documents to show the design 

changes. In such situations, the inspector should also verify that 

revisions of the controlled documents incorporating the marked-up 

changes are performed in a timely manner following the 

modification.

(c) Procedure changes should be in accordance with licensee 

processes and regulatory requirements. Verify the adequacy of all 

procedure changes which resulted from recent (within the last year) 

license change(s) or revision(s) to a technical specification or the top 

tier program document. Verify procedure changes are in 

conformance to 10 CFR 50.59. This item applies only to changes to 

procedures which are described or summarized in the UFSAR, 

normally a small portion of the procedures in use at the facility. 

General guidance and contrasting examples relating to the 

procedure changes which can be made by the licensee are 

described in NRC Inspection Manual Part 9900, AGuidance on 

10 CFR 50.59 -- Changes to Facilities, Procedures, and Test (or 

Experiments).@

4. Review a selection of records produced from implementation of the 

procedures. Review the record file system to determine if the records are 

adequately filed and controlled in accordance with the procedure. Verify 

that the records are legible and have the appropriate sign-offs as required 

by the procedure.

c. Key Attribute - Human Performance for Public Radiation Safety
Issue Date: 02/09/11 58 95003

1. Using data from the licensee=s corrective action program, LERs, and 

audits, determine if human performance issues have contributed to 

performance issues. Evaluate the overall effectiveness of human 

performance corrective actions by reviewing the licensee=s corrective 

action commitments. Determine if the problems were reviewed by the 

appropriate level of management and prioritized according to their safety 

significance. Evaluate whether the corrective actions were technically 

correct and implemented in a timely manner.

2. Review the following components of human performance, as related to the 

previously identified human performance issues.

(a) Work Control - Are assignments and technical information from 

management being effectively being communicated to the workers?

(b) Decision-making - For identified areas of human performance 

problems and while observing control room and local operations and 

other work activities, verify that the following decision-making 

practices support human performance:

(1) The roles and authorities of personnel are clearly defined and 

understood.

(2) TS and/or procedure prerequisites are satisfied before 

procedures are executed.

(3) Operational decisions and their bases are communicated.

(4) Interdisciplinary input and reviews of safety-significant or risksignificant decisions are sought.

(5) Decision-making is systematic when personnel are faced with 

uncertain or unexpected plant conditions.

(6) Conservative assumptions are used and possible unintended 

consequences are considered.

(c) Work Practices - Assess this area while observing the performance 

of procedures.

(d) Resources 

(1) Review the licensee=s program to provide training and periodic 

retraining to plant and contractor personnel on assigned duties 

and on safety significant changes to programs and 

procedures. Determine whether this training includes lessons 

learned from recent industry events and NRC communications 

(i.e., Information Notices, Generic Letters, Administrative 

Letters, etc.) and the proper use of human performance tools. 
Issue Date: 02/09/11 59 95003

By discussion with selected personnel, review of training 

lesson plans, and completed training records, determine if the 

requirements of 10 CFR 19.12 are met. If possible, observe 

portions of the general employee training (focus on the 

practical aspects of the training).

(2) Review the licensee=s overtime program and process to 

determine how management ensures that workers are not 

assigned safety related duties while in a fatigued condition. 

Interview workers to determine if they worked hours greater 

than specified in the technical specifications (with or without 

approval) to evaluate any repetitive nature that can lead to a 

degradation of performance.

(3) Refer to IP 93002, ―Managing Fatigue,‖ for guidance on the 

requirements of 10 CFR 26, Subpart I – Managing Fatigue. 

Assess the need for performing a specific section of the 

procedure based on previous performance issues related to 

fatigue management.

3. No specific guidance provided.

4. Interview several plant and contractor personnel (i.e., technicians, 

engineers, health physicists, and supervisors) associated with the 

program to assess their level of knowledge about the program and 

procedures and to determine their qualifications for the assigned position 

and duties. Evaluate training, experience, and qualifications by reviewing 

job documentation (usually specified in a licensee document), direct 

observation, and discussion with the individual. 

03.06 Assessment of Performance in the Safeguards Strategic Performance Area.

a. Key Attribute – Physical Protection (PP)

1. Using IP 71130.04 ―Equipment Performance, Testing, and Maintenance,‖

verify and assess whether the licensee’s testing and maintenance 

program has been appropriately developed and is being effectively 

implemented in accordance with the NRC-approved security plan, 

regulatory requirements, and any other applicable Commission 

requirements to assure the functionality and reliability of all security 

equipment necessary for effective protection of the facility and for a rapid, 

capable response to a design basis threat (DBT) and other contingencies.

2. Using IP 71130.05 ―Protective Strategy Evaluation‖ verify the licensee’s 

implementation of its nuclear security training program in accordance with 

the NRC-approved training and qualification (T&Q) plan, regulatory 

requirements, and any other applicable Commission requirement.
Issue Date: 02/09/11 60 95003

3. Using IP 71130.09 ―Owner-Controlled Area Controls‖ verify and assess 

the licensee’s OCA controls to ensure that they provide high assurance 

against the Design Basis Threat (DBT) in accordance with the NRCapproved security plan, regulatory requirements, and any other applicable 

Commission requirement.

b. Key Attribute - Access Authorization (AA)

1. Using data from the licensee=s corrective action system, and Inspection 

Procedures IP 71130.01, AAccess Authorization@ and IP 71152 verify that 

the licensee is identifying problems, entering those problems into their 

correction action system at an appropriate threshold, and effecting 

corrective action to prevent recurrence.

2. Review current regulatory requirements on behavior observation and 

identify license procedures that were changed under 10 CFR 50.54(p). 

See IP 71130.01 for further guidance. 

c. Key Attribute - Access Control

1. Using IP 71130.02, AAccess Control@ review any open LER, safeguards 

log and any self-assessments associated with access control for follow up, 

if necessary.

2. Using IP 71130.02, perform 02.02 (b) through (j), or as warranted. Pay 

particular attention to (h) if the licensee has a process for granting access 

to plant equipment, including vital equipment to authorized personnel who 

have an identified need for such access. Verify that access authorization 

criteria established by the security plan and procedures is being 

adequately implemented. See IP 71130.02-02, Section 02.02(h) for 

further guidance.

d. Key Attribute - Response to Contingency Events (protection strategy, program 

design, and support elements)

1. Using IP 71130.03, AContingency Response - Force-on-Force Testing,@

review the recorded description of the current response strategy and 

response plans to evaluate the effectiveness of the protective measures 

contained in a sample of those written response plans.

(a) Inspectors may request that the licensee perform table-top drills to 

assess and understand how the written strategy is effective and 

ascertain if it is implemented as described. See IP 71130.03, for 

further inspection guidance.

(b) Conduct interviews and meet with the appropriate licensee 

managers to review the protective strategy and discuss any 

identified weakness in that strategy.
Issue Date: 02/09/11 61 95003

e. Key Attribute - Response to Contingency Events (performance-based force-onforce Exercises and Target Set Evaluation)

Using IP 71130.05, ―Protective Strategy Evaluation‖, the inspectors should 

request that the licensee explain their target set analysis and conduct force-onforce evaluated exercises. See IP 71130.05 for guidance.

f. Key Attribute – Materials Controls and Accountability 

Using IP 71130.11 ― Material Control and Accounting (MC&A)‖ verify that the 

licensee has implemented and is maintaining an adequate and effective 

program to control and account for the SNM in its possession to ensure that the 

licensee can detect loss, theft, or diversion of SNM in a timely manner

g. Key Attribute - Security Plan Changes

Over a 2-year period, review and assess changes to the security plan and 

associated security procedures which appear to affect the ability of the 

safeguards program to prevent core damage. Determine if any changes 

decrease the effectiveness of the security plan or program, and if so, if those 

changes were reported to the NRC for approval prior to making the change. 

See IP 71130.04 for further inspection guidance.

03.07 Evaluate the Licensee=s Third-Party Safety Culture Assessment.

This step focuses on evaluating the quality of the third-party safety culture assessment 

(e.g., the methods used, sampling strategies, team qualifications, and the use of safety 

culture assessment protocols that are acceptable to the NRC). At such time that an 

industry safety culture assessment methodology is developed and found to be 

acceptable by the NRC, the guidance in this section will be evaluated for potential 

revisions to address the use of such a methodology. 

The assessment method(s) used by the licensee’s third-party contractor should follow 

the professional standards and methodologies established for conducting organizational 

assessments which are similar to the licensee safety culture assessment. For example, 

if surveys are used, general survey techniques for ensuring the reliability and validity of 

the methodology and results should be followed (guidance for NRC evaluation of 

surveys can be found in Enclosure 95003.02-F). Using such methods provides NRC 

with some assurance of the validity and reliability of the results. In contrast, if the 

assessment does not follow such methods or meet such criteria that will be factored into 

the NRC’s decision regarding the scope of the graded safety culture assessment. 

a. Inspection Preparation

1. The lead SCA should begin interactions with the licensee as early as 

possible during the planning and conduct of the third-party safety culture 

assessment to gain an understanding of the assessment approach. 

Monitoring and observations should continue throughout the assessment 

to the extent possible. Care must be taken to minimize any potential 
Issue Date: 02/09/11 62 95003

effects of NRC’s presence during assessment activities on participants’ 

behavior and consequently the results. Generally it would not be 

appropriate to observe the conduct of the third-party assessment 

interviews or focus groups. However, it would be appropriate to review 

the planned third-party assessment focus group protocol in advance and 

interview/focus group notes and summary documents afterwards.

Communicate frequently with the licensee to stay informed of the status of 

implementation activities (e.g., conduct of survey, analysis of results) and 

emerging issues. Be aware of how the licensee and/or the third-party 

personnel resolve these issues.

2. No specific guidance.

3. From the licensee, obtain the following:

(a) Tools and instruments used to conduct the licensee’s third-party 

safety culture assessment(s). These could include (but are not 

limited to) questionnaires, interview guides, or checklists, and the 

charter for the assessment(s).

(b) Documents produced by the assessment team that conducted the 

licensee=s most recent safety culture assessment. These could 

include (but are not limited to) an assessment plan, surveys, 

interview plans and reports, status memos, briefing notes, and 

interim and final reports.

(c) Documents that characterize the licensee=s response to the most 

recent safety culture assessment. These could include (but are not 

limited to) memoranda, meeting notes, corrective action program 

records, project plans, or other initiatives that were associated with 

or were initiated as a result of the assessment.

(d) Names, qualifications, and contact information for the personnel who

conducted the assessment.

Note: If the tools, instruments, or related licensee documents are 

proprietary, handle them in accordance with standard NRC 

procedures for handling proprietary information.

4. Obtain any safety culture assessments conducted at the site within the 

past five years to look for trends, licensee actions to address issues raised 

by the assessments, and information regarding effectiveness of the 

actions taken to resolve the issue.

b. Evaluation

1. The licensee=s terminology may differ from NRC terminology for the same 

application, e.g., the licensee may call safety culture components by other 
Issue Date: 02/09/11 63 95003

terms such as safety culture attributes or principles, but the concepts 

addressed should be similar.

2. Verification of comprehensiveness of licensee assessment

(a) It is important to verify that adequate samples of functional groups 

and organizational levels were assessed. That is, a safety culture 

assessment that focuses only on the functional groups who perform 

work that has a clear nexus to safe plant operations (e.g., 

operations, maintenance, engineering, security) but excludes 

individuals from other support groups or contract organizations will 

be incomplete. Functional groups, such as human resources, 

financial services, and some technical support organizations, and 

contractor groups often fulfill roles in the organization that are 

important in shaping the site=s safety culture. 

(b) Similar to the discussion in section 2.(a). above, a safety culture 

assessment that focuses only on some of the organizational levels 

may bias the results. 

(c) A key question to answer about the licensee=s third-party safety 

culture assessment is whether the sample sizes used were 

adequate to ensure that the findings and conclusions from the 

assessment were representative of the populations and 

subpopulations of interest.

(1) In general, if the licensee=s assessment team administered a 

survey in-person to groups of licensee employees and 

contractors and their sampling plan was to obtain responses 

from all site personnel, the number of survey respondents 

should be about 80% of the site population.

(2) If the licensee=s assessment plan was to administer the site 

survey by mail or electronically, the number of survey 

participants should typically fall between 60% and 70% of 

those who were asked to participate.

(3) If the survey results were based on lower percentages of the 

population than was identified in the licensee=s sampling plan, 

then the licensee=s assessment team should have collected 

and analyzed information to demonstrate that those who did 

participate and those who did not were not systematically 

different in a way that could bias the results of the survey. 

For example, if the survey systematically excluded everyone 

on the back shift, it is unlikely that the results would be valid. If 

there are inconsistencies in response rates among functional 

groups, i.e., certain group(s) exhibited lower participation 
Issue Date: 02/09/11 64 95003

rates, the licensee’s assessment team should have taken 

actions to understand the reasons for the differences and the 

effect on the accuracy of the data. 

Additional guidance related to appropriate sample sizes for 

individual and group interviews, structured behavioral 

observations, and event follow-up studies can be found in 

Enclosures 95003.02-C, 95003.02-D, and 95003.02-E, 

respectively.

(d) The safety culture components are detailed in IMC 0310.

3. In determining whether the methods used by the third-party assessment 

team to collect and analyze the data were adequate and appropriate:

(a) Determine whether the licensee=s third-party safety culture 

assessment contractor ensured, to the extent possible, that 

information obtained during the assessment was not attributable to 

individual participants in any reports of assessment results or in 

discussions with others who were not members of the assessment 

team. 

(b) If the third-party safety culture assessment included interviews, then 

evaluate the interview questions, the plan by which interviewees 

were selected, and the interview techniques used by the 

assessment team. (For related guidance, see Enclosures 95003.02-

B and 95003.02-C.)

(c) If the assessment included focus groups, then evaluate the 

questions used in the focus group meetings, the plan by which 

participants were selected, and techniques used to facilitate 

participation in the meetings. (For related guidance, see Enclosures 

95003.02-B and 95003.02-C.)

(d) If the assessment included document reviews, then evaluate the 

assessment team=s selection of documents and their review 

methodology.

(e) If the assessment included direct observations of meetings and/or 

work activities, then evaluate the assessment team=s selection of 

meetings and activities to observe, the observers, and the 

observation methodology. If possible, observe similar meetings 

and/or work activities, to place the assessment team’s observations 

in proper context. (For related guidance, see Enclosures 95003.02-

D.)

(f) If the assessment included a structured survey, then determine if 

acceptable survey practices were used. Evaluate the survey 

instrument used, a sampling of raw survey data including write-in 
Issue Date: 02/09/11 65 95003

comments (if available), survey results, and documentation that 

describes how the survey was developed and the methods used to 

administer it, and the statistical analyses applied to the survey data 

to determine if acceptable survey practices were followed. (For 

related guidance, see Enclosure 95003.02-F.)

(g) For each method used, determine whether the sample sizes were 

adequate to ensure that results from the method were 

representative.

(h) For each of the methods used, determine whether:

(1) any method was likely to introduce any systematic bias into 

the results;

(2) the methods were applied consistently; and

(3) if multiple methods were used, the third-party assessment 

team verified the consistency of the results obtained from the 

different tools and instruments. 

(i) Do not consider normative data about other sites or other industries 

provided by the licensee=s assessment of safety culture when 

developing insights about the third-party assessment, except if the 

licensee also provides detailed information to permit verification of 

the applicability of the normative data (e.g., nature of the norms, 

sample size and representativeness, procedures followed in 

obtaining the samples).

4. In determining whether the licensee=s assessment team members were 

independent and qualified:

(a) Verify that the third-party assessment team did not include any 

members of the licensee=s organization or utility operators of the 

plant (licensee team liaison and support activities are not team 

membership).

(b) Determine whether the assessment team members who designed 

the safety culture assessment and analyzed the results were 

qualified through education and/or experience. There should be 

members on the team who have knowledge in conducting safety 

culture /organizational assessment types of activities, particularly at 

nuclear facilities. If the assessment includes a survey, verify that the 

team included members with survey design, administration, and 

analysis expertise.

(c) Determine whether the assessment team included members with 

knowledge in the technical areas and organizational issues being 

assessed.
Issue Date: 02/09/11 66 95003

5. Review the following items related to the licensee’s third-party safety 

culture assessment results:

A sample of the assessment team’s interview or observation notes;

Responses to survey items both at an overall level and by functional 

groups;

Statistical analyses performed; and

Responses from previous assessment activities, if similar 

techniques, such as the same or similar survey questions, were 

used, for comparison to current results.

Evaluate these items to determine the quality and accuracy of:

The assessment team’s interpretation of the data collected;

Rollup or summaries in capturing issues and themes from the data; 

and

The messages communicated to the licensee about the results.

If the third-party assessment team’s follow-up investigation for any weaknesses 

in the safety culture components involved sensitive information about the 

behavior of an individual, and an NRC SCA/inspector must review that 

information or receives such information, the SCA/inspector shall protect 

the individual=s identity and privacy to the extent possible. The NRC shall 

not disclose to licensee personnel any detailed information about the

individual or the related events, but shall disclose only general conclusions 

about the thoroughness of the third-party assessment.

03.08 Determine Scope of and Plan for NRC’s Graded Safety Culture Assessment.

a. The scope of NRC’s graded safety culture assessment will be based on the 

results of the evaluation of the licensee’s third-party safety culture assessment. 

The lead SCA will need to make this determination, in consultation with the 

appropriate team and Regional management. The scope will depend on factors 

such as the quality of the third-party safety culture assessment scope, methods, 

sampling, and analysis, and the qualifications of the third-party safety culture 

assessment team.

In some cases the timing of the third-party safety culture assessment and the 

initiation of the 95003 inspection may allow the staff to evaluate the adequacy of 

the third-party assessment methodology before its implementation. The team 

will communicate concerns to the licensee for their action as they determine to 

be appropriate. Based on the validity of the effort, by the licensee and/or thirdparty assessment team, to address NRC concerns, the NRC graded safety 

culture assessment can be adjusted accordingly.

1. The licensee’s activities to communicate results of the assessment to 

various levels of management and staff should be evaluated to 

understand the messages being provided. Obtain documentation 

regarding the licensee’s dissemination of the third-party safety culture 

assessment results (e.g., emails, newsletters, and briefing materials). 
Issue Date: 02/09/11 67 95003

Request any department/group specific information, including talking 

points if applicable, provided to managers and/or supervisors for their 

areas.

2. If the review conducted under 02.07 does not identify any weaknesses in 

the assessment methods, conclusions, or team qualifications, then the 

graded safety culture assessment should be focused on the licensee’s 

response to the assessment results. For example, if the assessment 

identified problems in any safety culture component(s) or weaknesses in 

certain groups, then the licensee’s response to those problems, to the 

extent they are available, should be evaluated. Depending on the timing 

of the inspection period, the licensee may not have made significant 

progress in developing or implementing corrective actions. In these 

cases, effectiveness of corrective actions may need to be evaluated 

during inspection follow-up activities. The lead SCA should discuss this 

with the team leader/assistant team leader and determine how best to 

conduct the evaluation.

3. If weaknesses are noted in portions of the assessment, the graded safety 

culture assessment should be adjusted to gather additional information in 

those areas. For example:

(a) If there were functional groups that were not adequately covered in 

the assessment (e.g., either not included in the scope, or had low 

response rates), conduct appropriate activities (e.g., focus groups, 

interviews, observations) to evaluate if those groups have any 

weaknesses in safety culture components. For groups with low

survey response rates, verify the third-party assessment team’s 

conclusions about the reasons for the low participation and evaluate 

the licensee’s response, such as the licensee conducting additional 

assessment activities.

(b) If the assessment did not include certain organizational levels (e.g., 

of senior/corporate management), conduct appropriate activities 

(e.g., interviews and observations) to gain information on those 

level’s effect on the site’s safety culture, including any attitudes and 

behaviors that may be inconsistent with those described in the 

safety culture components.

(c) If issues are identified with the sample sizes, conduct appropriate 

assessment activities (e.g., focus groups and interviews) with 

groups that were inadequately sampled to determine if there are 

issues the licensee’s assessment did not identify.

(d) If any of the safety culture components are determined to be 

inadequately assessed, conduct assessment activities to evaluate 

those components using guidance from Enclosures 95003.02-A 

through F. Coordinate with the other inspection team members who 

may be focusing on related areas, particularly for components 
Issue Date: 02/09/11 68 95003

related to the SPA of identifying, assessing, and correcting 

performance deficiencies and the human performance key attribute.

4. If specific weaknesses or concerns are identified with the third-party safety 

culture assessment team’s methods, conclusions, or qualifications, the 

graded safety culture assessment should be adjusted to gather additional 

information in those areas. Conduct limited assessment activities to 

evaluate whether the licensee’s third-party assessment results are 

consistent with those gathered by NRC.

(a) If there were issues noted regarding the use of certain methods

except for surveys, see note below, NRC should independently 

conduct those activities. For example, if problems were identified 

with the conduct of focus groups or with interview techniques, NRC 

should conduct its own focus groups and interviews. 

Note: NRC does not conduct surveys. Therefore, for weaknesses 

identified in survey methodology, NRC will use other techniques 

(i.e., those described in Enclosures 95003.02-C through F) to 

evaluate the validity of the survey results.

(b) The limited assessment should start with functional groups that have 

a clear nexus to safe plant operations (e.g., operations, engineering, 

maintenance) and/or those with known problems (e.g., through the 

third-party assessment or other means) and be expanded as 

needed.

(c) Based on the results of the limited assessment, adjust the scope as 

appropriate. For example, if NRC’s data validate results from the 

third-party safety culture assessment, then the focus of the graded 

assessment can be shifted to the licensee’s response to the results, 

to the extent actions have been conducted or planned. However, if 

there are inconsistencies, the scope of the graded safety culture 

assessment should be broadened, such as including additional 

assessment methods and increasing the range of functional groups 

and/or safety culture components being targeted.

(d) In planning the assessment activities, such as developing the tools 

and designating assignments, follow the guidance in section 1.b. 

from Attachment 95003.02 to ensure use of multiple methods/team 

members so that information is collected independently.

5. If substantial weaknesses are identified with the licensee’s third-party 

safety culture assessment or NRC has low confidence in the validity of the 

licensee’s results, the determination should be made whether the NRC 

should conduct an independent safety culture assessment in order to gain 

accurate insights on the contribution of weaknesses in safety culture 

components to licensee performance. If an independent NRC safety 
Issue Date: 02/09/11 69 95003

culture assessment is determined to be needed, follow the guidance in 

Attachment 95003.02 to conduct the assessment. 

b. Review Attachment 95003.02 regarding the conduct of NRC’s independent 

safety culture assessments and Enclosures 95003.02-C through F regarding 

specific data collection methods. Apply the guidance as appropriate (based on 

the specifics of the case) in planning the graded safety culture assessment and 

developing the methods and tools. Be aware of overlaps between other 

inspection focus areas and the graded safety culture assessment activities (e.g., 

in certain safety culture components or functional groups), and use the data and 

insights from the other areas to the extent possible.

c. The lead SCA will provide resource needs to the team, Regional, and program 

office management. Depending on the focus of assessment activities, specific 

expertise, such as those possessed by Headquarters staff and/or contractors, 

may be necessary to conduct the graded safety culture assessment effectively. 

The level of resources will depend on the scope and can be affected by the size 

of the site. After resources are identified, the lead SCA will determine the 

assignment of activities based on the expertise and experience of the SCAs and 

other inspection team members and hold meetings/briefings as needed to 

communicate relevant information and assignments.

d. No specific guidance.

03.09 Perform NRC’s Graded Safety Culture Assessment.

a. Follow the scope and implement the plan developed under section 02.08.

1. Evaluate the communications provided to various levels (e.g., 

management and staff) regarding the third-party safety culture 

assessment for accuracy to the assessment results. Consider asking 

participants in focus groups and interviews (if held) about information 

received in this area, and evaluate the effectiveness of the licensee’s

accuracy in conveying the intended information.

2. Evaluate the licensee’s response to weaknesses identified in any safety 

culture components, to the extent they are available during the time of the 

inspection.

(a) Determine whether the licensee appropriately identified those 

weaknesses within their corrective action program. 

In some cases, corrective actions may involve sensitive areas such 

as personnel actions or other matters that warrant confidentiality. 

These types of information may not be documented in any corrective 

action programs and must be solicited or inferred from discussions 

with licensee officials, such as Human Resource personnel or senior 

management. The lead SCA should evaluate these circumstances 
Issue Date: 02/09/11 70 95003

and conduct activities to gather this information as needed. The 

lead SCA should determine the extent of involvement of and 

knowledge by other team members in these activities on a need-toknow basis.

(b) Determine whether the licensee=s evaluations of those weaknesses 

were appropriate and the resulting planned corrective actions 

appear adequate for resolving those weaknesses. 

The breadth and depth of corrective actions should be appropriate to 

produce the targeted changes in the organization’s characteristics, 

attitudes, and behaviors that define the organization’s safety culture. 

For a discussion on what these concepts involve, review the 

introduction section of Attachment 95003.02. Although short-term or 

limited scope actions such as training or personnel changes can 

have positive impacts, effective corrective actions for producing 

lasting changes in aspects of culture require a long-term focus. 

Discrete activities such as communications (e.g., stand-downs, 

publication of policies) and training sessions should be reinforced 

and evaluated for effectiveness. The licensee should have plans to 

monitor long-term progress and the capability and flexibility to make 

adjustments to corrective action plans as needed.

(c) Determine whether the licensee has made reasonable progress in 

implementing those actions.

In making this determination, consider the types of actions and the 

timeframe of the desired results. The licensee may implement some 

actions aimed at creating immediate changes or near term 

improvements and others focused on long term changes. It is 

important to note that some cultural changes may require 

timeframes of several years or longer to develop, depending on the 

circumstances. However, short term progress can be made and 

should be monitored. Depending on the timing of the inspection, 

evaluate the progress made based on the types of corrective actions 

and their intended effects.

(d) Depending on the circumstances, the licensee may not have made 

significant progress in developing or implementing corrective actions 

by the on-site inspection period, or the corrective actions in place 

may need additional time to facilitate the intended improvements. In 

these cases, the effectiveness of corrective actions will need to be 

evaluated during inspection follow-up activities at a later time. The 

lead SCA should discuss this with the team leader/assistant team 

leader and determine how to conduct the follow-up.

3. If a limited scope NRC safety culture assessment is conducted, determine

whether the results of the licensee=s overall assessment of safety culture, 

including the third-party and any other relevant activity, are consistent with 
Issue Date: 02/09/11 71 95003

results obtained by the NRC assessment by answering the following 

questions:

$ Are the results of NRC’s data collection methods generally 

consistent with results of the licensee=s methods?

$ Do similar functional groups show differing results?

$ Did either assessment identify weaknesses in particular safety 

culture components?

$ Did the NRC SCAs reach the same general conclusions relative to 

the safety culture components?

If significant inconsistencies exist between the NRC=s completed results 

and the licensee=s overall results, then ask the licensee to determine the 

reason(s) for each inconsistency. This may require the licensee to 

perform additional assessment activities. In addition, consider increasing

the scope of the NRC’s assessment, including broadening the functional 

areas and/or increasing the depth to which applicable safety culture 

components are evaluated. 

4. The lead SCA has the flexibility to propose adjustments to the scope of 

the graded safety culture assessment to the team leader/assistant team 

leader, based on factors such as insights from the data, similarities and 

discrepancies between NRC and licensee results, licensee response and 

actions, and other emerging issues. The lead SCA should keep the team 

leads fully apprised of potential changes and coordinate increases or 

decreases in the scope and the resources needed.

5. If an independent NRC safety culture assessment is determined to be 

needed, follow the detailed guidance in Attachment 95003.02 to conduct 

the assessment. 

6. It is important to note that disclosure of any sensitive information received, 

reviewed, or collected by the NRC inspection team shall be limited to only 

those members who have a specific need-to-know for completing their 

inspection requirements. For example, although it may be necessary for 

an SCA/inspector to review case files from the licensee’s employee 

concerns program, the SCA/inspector should report only the overall 

conclusions from the review to the remainder of the team.

b. Based on results from the licensee’s third-party safety culture assessment and 

the NRC’s graded safety culture assessment, follow the guidance in section 2.e. 

of Attachment 95003.02 in compiling the data. Determine whether any trends or 

themes in a particular safety culture component exist and work with the entire 

team to determine the contribution of weaknesses in safety culture 

component(s) to the findings being identified in the inspection and to the 

affected SPA(s).
Issue Date: 02/09/11 72 95003

03.10 Performance Deficiency Cause Analysis.

The purpose of the performance deficiency cause analysis is to provide a diagnosis of 

the principle causes for the decline in performance as well as a prognosis for future 

improvement. Using the results from this inspection, in conjunction with information 

obtained from the NRC’s review of previous root cause analyses (validated by either 

IP 95001 or IP 95002) that may have been performed by the licensee or others, the 

team should group related apparent, root and contributing causes of the risk significant 

performance deficiencies using a structured approach. This analysis should also 

include or consider the existing ROP substantive crosscutting issues as well as new 

findings with cross-cutting aspects that are identified from this inspection. The team 

should integrate significant insights from the safety culture observations for this 

analysis. The outcome of this analysis should be the primary cause(s) of the decline in 

performance and a discussion of how the improvement / recovery plans will address 

these causes. The team (or at the minimum a representative from each functional area 

of the team) should participate in this analysis. It should be noted that this effort is not 

intended to be a substitute for a more focused root cause study or self-assessment by 

the licensee.

The senior reactor analyst should perform an assessment of the individual risk 

associated with the team’s findings. The SRA may perform a collective risk assessment 

by summing or qualitatively assessing the risk impacts of multiple separate or 

independent findings that overlap in time to gain an understanding of the aggregated or 

collective risk profile. When performing the collective risk assessment, it is important to

clearly ascertain the time history (appropriate identification of start and end dates) of 

each overlapping inspection finding to reach a proper result. Assessing the collective 

risk from the "roll-up" of multiple related, non-overlapping independent findings or of 

combining all of the findings identified during the inspection would produce an artificially 

high risk estimate leading to incorrect conclusions.

This information will be useful in evaluating the adequacy of licensee proposed 

corrective actions to the performance issues, and to aid in deciding if additional 

regulatory actions are warranted. 

03.11 NRC Assessment.

Perform a limited review of the NRC=s assessment and inspection process at the 

subject facility.

a. Should the results of this inspection indicate that a significant reduction in safety 

has occurred, compare the team=s findings with current assessment data (both 

PIs and inspection findings) to determine if sufficient warning was provided. If 

the results of this inspection indicate that a significant reduction in safety has not 

occurred, compare the team=s findings with the current assessment data to 

identify inconsistencies in the plant performance data.

b. Evaluate whether the NRC assessment process appropriately characterized 

licensee performance based upon the data that was provided. Evaluate for 

example, whether inspection findings were appropriately screened using the 
Issue Date: 02/09/11 73 95003

Significance Determination Process (SDP) for risk significance, and was this 

data appropriately entered into the NRC action matrix.

03.12 Document Inspection Results.

Due to the diagnostic nature of this inspection, a thorough documentation of the team=s 

observations, findings, and conclusions is required. Unlike the content of baseline 

inspection reports, this inspection report should contain sufficient observations and 

issue details to allow the development and support of the team’s diagnostic conclusions. 

The observations and findings should support the assignment of the cross-cutting 

aspects to the team’s findings, and the cross-cutting aspects or themes should support 

the diagnostic conclusions of the team. It is neither necessary nor desirable to report 

separately on every key inspection attribute. The report should focus primarily on the 

diagnostic conclusions and should logically and coherently support those conclusions. 

IMC 0612, APower Reactor Inspection Report,@ guidance regarding the threshold to only 

document greater than minor findings is not applicable to this procedure. Although 

certain issues should be evaluated using the SDP, this may not be possible for many of 

the team=s more programmatic conclusions.

Based upon insights derived from the performance deficiency causal analysis results 

(section 3.10) collectively performed by all of the team functional area groups, a crosscutting aspect is evaluated in accordance with IMC 0612 for findings identified by the 

team. The inspection report should document the information and analysis used to

assign the cross-cutting aspect and should clearly explain how the selected crosscutting aspect is applicable (i.e., was the most significant contributor) to the specific

circumstances of the inspection issue.

In the inspection report, include the following information in the major sections:

a. Strategic Performance Area Assessment

1. Inspection Scope

If only one SPA is degraded, then subdivide this section to address the

appropriate key attribute(s) of the SPA. However, if more than one SPA is 

degraded, then first subdivide this section into SPAs, and then subdivide 

each SPA subsection further to address the appropriate key attribute(s) of 

the SPA.

For the appropriate key attribute(s), describe the documents and records 

reviewed, personnel interviewed, walkdowns conducted, activities 

observed, etc., to satisfy the inspection requirements associated with the 

attribute.

2. Observations & Findings

List important observations which are not findings but which support the 

assessment result. Also list and document in accordance with IMC 0612 

any findings which were identified during this assessment.
Issue Date: 02/09/11 74 95003

3. Assessment Result

Document a summary assessment of licensee performance in each 

degraded SPA, with reference to the observations and findings which 

support the assessment.

b. PI&R Assessment

1. Inspection Scope

Describe the documents and records reviewed, personnel interviewed, 

walkdowns conducted, activities observed, etc., to complete this 

assessment.

2. Observations & Findings

List important observations regarding PI&R which are not findings but 

which support the assessment result. Also list and document in 

accordance with IMC 0612 any findings which were identified during this 

assessment.

3. Assessment Result

Describe the overall assessment of licensee performance in PI&R that is 

supported by the observations and findings revealed during this 

assessment. Ensure that the basis for this assessment is fully contained in 

the Inspection Scope and Observations & Findings sections.

c. Safety Culture Assessment Activities [C1]

1. Scope

Describe the third-party assessment evaluation conducted, such as the 

documents and records reviewed, personnel interviewed, activities 

observed, and the NRC team’s engagement, if any, with the licensee and 

the third-party assessors during the conduct of the third-party safety 

culture assessment. In addition, describe the graded safety culture 

assessment activities conducted, such as focus groups, interviews, 

document reviews, and observations. Be sensitive about documenting 

only non-proprietary information related to the third-party safety culture 

assessment.

2. Observations & Findings

Document the aggregated results derived from the evaluation of the thirdparty safety culture assessment and the graded safety culture 

assessment. Include the results of the performance deficiency causal 

analysis, evaluation of the associated cross-cutting aspects assigned to 
Issue Date: 02/09/11 75 95003

the team’s observations and findings, and consideration of accompanying 

insights from the SCAs/inspectors about the licensee’s safety culture they 

obtained during the inspection process.

3. Assessment Result and Diagnostic Conclusions

Document a summary assessment from the safety culture assessment 

activities, highlighting significant weaknesses that are found to exist in any 

safety culture components or functional/organizational area. The 

weaknesses should be supported by the observations/ findings revealed 

during the inspection process and results from the licensee’s third-party 

safety culture assessment, as applicable. Provide an evaluation of the 

licensee’s response to the identified weaknesses of any safety culture 

components. If the team’s assessment of a safety culture component has 

been documented in another section of the report (for example as part of 

the PI&R assessment documentation), that discussion can be referenced.

95003-04 RESOURCE ESTIMATE

The resource estimates provided are for direct inspection only, based on a three week 

on-site inspection. Not all areas will be performed during each inspection and the hours 

required to compete each area may be less for plants where previously identified 

performance issues were isolated. The hours required to complete each area could 

also be greater based on site-specific circumstances. For planning purposes, the ROP 

budgets 3000 hours (distributed among the four regions) to conduct one IP 95003 

inspection per year. The resource estimates are not requirements and inspection 

staffing needs are based upon site-specific circumstances.

Position/Inspected Area Manhours

Team Leader..........................................................................120

Assistant Team Leader ..........................................................120

Licensee Control Systems......................................................240

Licensee=s Safety Culture Assessment ...........................120-160

Safety Culture Assessment Activities................................80-360

Design....................................................................................360

Human Performance ..............................................................120

Procedures.............................................................................120

Equipment Performance.........................................................120

Configuration Control .............................................................240

EP without Attachment 95003.01 .............................................80

EP with Attachment 95003.01*...............................................160

Occupational Radiation Safety ...............................................200

Public Radiation Safety ............................................................60

Safeguards............................................................................TBD

Senior Reactor Analyst........................................................0-120

Review of Assessment Process ...............................................40 (not direct inspection)
Issue Date: 02/09/11 76 95003

* Including a remedial exercise in the scope of Attachment 95003.01 will require an 

additional 40 hours, resulting in a total of 200 hours.

95003-5 PROCEDURE COMPLETION

Meeting the inspection objectives defined in Section 95003-01 of this IP will constitute 

competition. Refer to IMC 0305 for additional regulatory actions and considerations.

95003-6 REFERENCES

IMC 0305, ―Operating Reactor Assessment Program‖

IMC 0310, AComponents within the Cross-Cutting Areas@

IMC 0320, ―Operating Reactor Security Assessment Program‖

IMC 0609, ―Significance Determination Process‖

IMC 0612, ―Power Reactor Inspection Reports‖

IMC 0612, Appendix C, ―Guidance for Supplemental Inspection Reports‖

IMC 2515, ―Light-Water Reactor Inspection Program - Operations Phase‖

IMC 2515, Appendix B, ―Supplemental Inspection Program‖

IP 40001, ―Resolution of Employee Concerns‖

IP 41500, ―Training and Qualification Effectiveness‖

IP 42001, "Emergency Operating Procedures"

IP 42700, APlant Procedures@

IP 71114.03, AEmergency Preparedness Organization Staffing and Augmentation

System@

IP 71114.05, ACorrection of Emergency Preparedness Weaknesses@

IP 71130.01, AAccess Authorization@

IP 71130.02, AAccess Control@

IP 71130.03, AContingency Response - Force-on-Force Testing@

IP 71130.04 ―Equipment Performance, Testing, and Maintenance‖
Issue Date: 02/09/11 77 95003

IP 71130.05, ―Protective Strategy Evaluation‖

IP 71130.09, ―Owner-Controlled Area Controls‖

IP 71130.11, ―Material Control and Accounting (MC&A)‖

IP 71152, ―Problem Identification and Resolution‖

IP 71841, AHuman Performance@

IP 82001, ―Evaluation of Emergency Preparedness‖

IP 82001.01, ―ERO Performance Drills‖

IP 82001.02, ―ERO Performance Drills Dose Assessment‖

IP 82001.04, ―Facilities and Equipment‖

IP 82001.05, ―Procedure Quality‖

IP 93002, ―Managing Fatigue‖

IP 95001, ―Supplemental Inspection for One or Two White Inputs in a Strategic 

Performance Area‖

IP 95002, ―Supplemental Inspection for One Degraded Cornerstone or Any Three White 

Inputs in a Strategic Performance Area‖

NUREG 1220, "Training Review Criteria and Procedures"

RG 8.38, "Control of Access to High and Very High Radiation Areas in Nuclear Power 

Plants"

Inspection Manual Part 9900, AGuidance on 10 CFR 50.59 -- Changes to Facilities, 

Procedures, and Test (or Experiments)‖

10 CFR 19.12, ―Instruction to workers‖

10 CFR 20.1206, ―Planned special exposures‖

10 CFR 20.1208,‖Dose equivalent to an embryo/fetus‖

10 CFR 26 Subpart I, ―Managing fatigue‖

10 CFR 50.49, ―Environmental qualification of electric equipment important to safety for 

nuclear power plants‖

10 CFR 50.54, ―Conditions of licenses‖

10 CFR 50.59, ―Changes, tests and experiments‖

10 CFR 50.65, ―Requirements for monitoring the effectiveness of maintenance at 

nuclear power plants‖
Issue Date: 02/09/11 78 95003

10 CFR 50.120, ―Training and qualification of nuclear power plant personnel‖

END

Attachments:

95003.01 Emergency Preparedness

95003.02 Guidance for Conducting a Full NRC Safety Culture Assessment

Attachment 1 Revision History for IP 95003
Issue Date: 02/09/11 Att1-1 95003

Attachment 1 – Revision History for IP 95003

Commitment 

Tracking 

Number

Issue Date & 

Accession 

Number

Description of Change Training 

Needed

Training 

Completion 

Date

Comment 

Resolution 

Accession Number

N/A 04/03/00

CN 00-003

Initial Issue. Yes

N/A 01/17/02

CN 02-001

Revised to incorporate lessons learned 

from Indian Point 2 inspections, and adds 

a section on security, which was not 

included in the initial version of this 

procedure.

No

C1 10/26/06

ML062970393

CN 06-030

Revised procedure to incorporate safety 

culture enhancements as required by the 

Safety Culture Initiative (ref: Staff 

Requirements - SECY-04-0111 -

Recommended Staff Actions Regarding 

Agency Guidance In The Areas Of Safety 

Conscious Work Environment And Safety 

Culture.)

Yes 07/01/06 ML062980489

N/A 01/15/09

ML080040267

CN 09-002

Procedure revised to incorporate the 

results of the ROP safety culture lessons 

learned evaluation (reference SECY 06-

0122). Changes made for ROP feedback 

forms 95003-1233, 95003-1234, 95003-

1235, 95003-1261, 95003-1238.

No N/A ML083430521

N/A 11/09/09

ML092680677

CN 09-026

Added reference to IP 93002, ―Managing 

Fatigue‖

No N/A N/A

N/A 02/09/11

ML102020551

CN 11-001

Defined procedure completion criteria

and add reference section.

No N/A ML110120516