Issue Date: XX/XX/XX 95003 -1-

NRC INSPECTION MANUAL IPAB 

INSPECTION PROCEDURE 95003 

SUPPLEMENTAL INSPECTION FOR REPETITIVE DEGRADED 

 CORNERSTONES, MULTIPLE DEGRADED CORNERSTONES, 

 MULTIPLE YELLOW INPUTS OR ONE RED INPUT 

PROGRAM APPLICABILITY: 2515 

CORNERSTONES: ALL 

INSPECTION BASIS: The NRC=s revised inspection program includes three parts: 

baseline inspections; generic safety issues and special 

inspections; and supplemental inspections performed as a 

result of risk significant performance issues. The inspection 

program is designed to apply NRC inspection assets in an 

increasing manner when risk significant performance issues 

are identified, either by inspection findings evaluated using 

the significance determination process (SDP) or when 

performance indicator thresholds are exceeded. 

Accordingly, following the identification of an inspection 

finding categorized as risk significant (i.e., white, yellow, or 

red) via Table 2 of the SDP, or when a performance 

indicator exceeds the @licensee response band@ threshold, 

the NRC regional office will perform supplemental 

inspection(s). The scope and breadth of these inspections 

will be based upon the guidance provided in the NRC=s 

AAssessment Action Matrix@ and the Supplemental 

Inspection Selection Table (included in 2515 Appendix B). 

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. 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 
 

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require that some sample reviews be performed for all key 

attributes of the affected strategic performance areas. The 

rational behind this is that additional NRC assurance is 

required to ensure public health and safety, beyond that 

provided by the baseline inspection program and the 

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. 

This procedure was developed with consideration of the 

following boundary conditions: 

1. The NRC is performing the inspection however the 

staff’s determination of the appropriate graded safety 

culture assessment scope to satisfy the completion of 

the inspection requirements will be based on the 

results of their evaluation 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 that may be 

developed into a finding will be evaluated using the 

significance determination process during the course 

of the inspection as time permits. For new issues that 

have a high likelihood to be greater than Green the 

issue will be fully developed during the inspection. 

For new issues with a high likelihood to be Green, the 

use of Unresolved Items (URIs) with a separate 

follow-up inspection is acceptable to resolve technical 

concerns identified by the team to allow the team to 

concentrate on performing their diagnostic 

assessment; 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. The licensee has completed a root-cause, extent-ofcause, 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. Flexibility is afforded to 

perform inspections and safety culture evaluations in 

Deleted: e

Comment [rag1]: Recommendation 

PV-20 to add flexibility. 

Comment [rag2]: Recommendation 

PV-20. 
 

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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 

exists. 

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

by the licensee to identify, evaluate, and correct performance issues. 

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 

graded safety culture assessments 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 

95001 or 95002) have been conducted to evaluate the licensee=s root cause, extent-ofcause, 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 typically include inspection of 

the licensee=s evaluation of those issues. Discretion is allowed for the 95001 or 95002 

supplemental inspections to be performed as part of an associated Confirmatory Action 

Letter (or other regulatory action identified in IMC0305 section 06.05) follow-up inspection 

to evaluate the licensee’s resolution (which may not be ready at the time the 95003 is 

conducted) of the specific performance indicator or inspection finding issue. 

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. 

Comment [rag3]: Recommendation 

PV-20. 

Deleted: independently assess

Deleted: evaluate the licensee=s 

assessment of its

Comment [rag4]: Recommendations 

PV-20 and PV-24.

Comment [rag5]: Recommendation 

PV-22. 

Deleted: The procedure is not 

intended to duplicate the scope of 

previously performed baseline and 

supplemental inspections; however, 

some repetition may be necessary 

where previous inspections were not 

sufficient to fully scope the breadth 

and depth of licensee performance 

issues. 

Comment [rag6]: Recommendation 

PV-21. 
 

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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. When Attachment 95003.01 is implemented it 

supplants the EP related inspection requirements contained in the body of this 

procedure. 

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. 
 

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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. 

(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 modification to the system for review and determine if the 

system is capable of functioning as specified by the current design and 
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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 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 section 06.07 of Inspection Manual Chapter (IMC) 

0305 AOperating Reactor Assessment Program.@

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

accordance with Inspection Procedure 82001, 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 
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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. 

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. 
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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. 

(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. Self-assessment 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 Inspection Procedure 71114.03, AEmergency Response 

Organization Augmentation.@ If this inspection procedure has been 

implemented recently, the inspector may exercise judgement as to the need 

to implement the inspection procedure as part of the 95003 inspection effort. 

 If Attachment 95003.01 is being implemented, there are additional 

requirements under this key attribute to consider. 
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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 2515 Appendix B 

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 team members reviewing this aspect 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. 

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. 

The facilities 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 this aspect of the key attribute should coordinate their activities to ensure 

that the following inspection requirements are addressed: 
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1. Determine adequacy/implementation of the radiological source term controls 

and quality of related chemistry controls. 

2. Determine adequacy/implementation of work planning and controls, focusing 

on outage maintenance periods. 

3. 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. 

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.
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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 section 06.07 of Inspection Manual Chapter (IMC) 

0305 AOperating Reactor Assessment Program.@

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. 

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. 
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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 section 06.07 of Inspection Manual Chapter (IMC) 

0305 AOperating Reactor Assessment Program.@

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 

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

process used to verify trustworthiness of personnel prior to granting unescorted 

access to the protected area. The failure in the Access Authorization Program can 

compromise the licensee=s ability to adequately protect against the insider threat of 

radiological sabotage that may result in core damage. 

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

personnel screening 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. 
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b. Key Attribute - Access Control (Searches of personnel, packages, and vehicles). 

The Access Control program=s function is to prevent the introduction of contraband 

(i.e. firearms, explosives, incendiary devices) into the plant that could aid in the 

attempt to commit radiological sabotage. The failure of the Access Control 

program could compromise 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 into the protected area that could be used to 

commit sabotage. 

2. Verify that the identification and 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. 

c. 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 reasonable 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, properly armed, appropriately trained and within the appropriate time to 

protected positions to interdict and defeat the design basis threat in order to protect 

target sets, and thus prevent core damage. 

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. 

d. 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, properly armed, appropriately trained 

and within the appropriate time to protected defensive positions to interdict and 

defeat the design basis threat in order to protect against radiological sabotage and 

prevent core damage. 

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

appropriate target sets are identified and have an associated protective 

strategy 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. 

e. 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. 
Issue Date: XX/XX/XX 95003 -14-

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 Part 50.54(p). 

02.07 Evaluate the Licensee=s Third-Party Safety Culture Assessment 

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 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. 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. If engagement before or 

during the assessment implementation is possible, interact with the 

licensee to be informed of the status of safety culture assessment 

activities. 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 

that is available. 

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 and licensee 
Issue Date: XX/XX/XX 95003 -15-

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) and individuals from any contract 

organizations performing those functions; 

(b) The assessment included all levels of management with line 

responsibility for operating the plant, 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. Determine whether 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 

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; 
Issue Date: XX/XX/XX 95003 -16-

(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 

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. 

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 
Issue Date: XX/XX/XX 95003 -17-

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

issue(s). 

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, co-addressed 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. 

Flexibility is afforded in this procedure to treat the inspection as more diagnostic in nature. 

Rather than having the team expend resources to fully develop the regulatory basis of 

inspection issues and their associated risk significance, the team is allowed to make 

greater use of Unresolved items. This allows follow-up inspections to perform the requisite 

activities to fully characterize the performance deficiency and its associated significance. 

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. 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 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 

Deleted: Group the safety 

performance deficiencies identified 

during the inspection by apparent root 

and contributing causes. 

Comment [rag7]: Recommendation 

PV-6. 

Comment [rag8]: Recommendation 

PV-12. 

Comment [rag9]: Recommendation 

PV-11. 

Comment [rag10]: Recommendation 

PV-26. 

Deleted: arrange
Issue Date: XX/XX/XX 95003 -18-

contact the Branch Chief of the Performance Assessment Branch of the Division of 

Inspection and Regional Support of the Office of Nuclear Reactor Regulation to verify 

completion. 

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 opinion, 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. 

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)

Deleted: safety culture assessor

Deleted: , typically from 

headquarters

Deleted: safety culture assessor

Deleted: Thre

Deleted: safety culture assessor

Deleted: shall be assigned to the 

inspection team

Comment [rag11]: Integrate 

resolution from Recommendation PV-28. 
Issue Date: XX/XX/XX 95003 -19-

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 

education 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 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 data-gathering 

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) conducting content 

analyses of 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; 

$ 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 
Issue Date: XX/XX/XX 95003 -20-

$ 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 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, 

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 sub-group 

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 during the planning evolutions for 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 

Deleted: Qualification Requirements 

for Safety Culture Assessors [June]¶

¶

The team leader should verify that the 

safety culture assessors collectively 

have appropriate credentials 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) 

questionnaires and surveys, (4) 

behavioral observations and 

checklists, and (5) case studies;¶

¶

$ Ability to determine the 

applicability and likely usefulness of 

various data-gathering 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) conducting content analyses 

of 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, forcedchoice, 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;¶

¶

$ 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 

Comment [rag12]: Recommendation 

PV-10. 

... [1]
Issue Date: XX/XX/XX 95003 -21-

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 

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 the Emergency Preparedness cornerstone was degraded, 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 inspection procedure 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 Inspection Procedure 71152 

AIdentification and Resolution of Problems@. 

Deleted: T

Deleted: assessment

Deleted: ssessment

Deleted: assessors

Deleted: assessors

Deleted: assessors

Deleted: safety culture assessor
Issue Date: XX/XX/XX 95003 -22-

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, Inspection Procedure 71114.05 ACorrection of 

Emergency Preparedness Weaknesses and Deficiencies,@ 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 multiple workarounds 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 Inspection Procedure 40001, 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. 
Issue Date: XX/XX/XX 95003 -23-

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. 

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: 
Issue Date: XX/XX/XX 95003 -24-

(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? 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? 

(3) Have modified structures surrounding safety equipment, components, 

or structures been evaluated for seismic 2-over-1 considerations? Have 
Issue Date: XX/XX/XX 95003 -25-

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 time-frame 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? 

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 involves engineering in determining the operability of 

degraded safety systems and components (SSC=s). 
Issue Date: XX/XX/XX 95003 -26-

(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 

(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. 
Issue Date: XX/XX/XX 95003 -27-

(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; 

(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 Inspection Procedure 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; 
Issue Date: XX/XX/XX 95003 -28-

(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 cross-cut 

organizational boundaries. 

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

assess whether 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. 

(2) Operational decisions and their bases are communicated. 

(3) Interdisciplinary input and reviews of safety-significant or risk-significant 

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 
Issue Date: XX/XX/XX 95003 -29-

equipment of plant conditions, resulting in an operator workaround. 

(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. 

(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 Inspection Procedure 41500, 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. 
Issue Date: XX/XX/XX 95003 -30-

(aa) Review the licensee=s process for controlling overtime. 

(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. 

(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 inspection Procedure 

82001 Attachments 01 & 02 may be used to construct drill scenarios and 

evaluate performance. 

(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 
Issue Date: XX/XX/XX 95003 -31-

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 inspection procedure. 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, 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. 
Issue Date: XX/XX/XX 95003 -32-

(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 Part 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, "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 

Inspection Procedure 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. The NRC position concerning control of 
Issue Date: XX/XX/XX 95003 -33-

procedural adherence is described in NRC Inspection Manual Part 9900, 

"Technical Guidance, Operations -- Procedural Adherence." 

(g) NRC Inspection Procedure (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) Inspection Procedure 82001.05 contains guidance for the inspection of 

emergency plan implementing procedures. 

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. 
Issue Date: XX/XX/XX 95003 -34-

(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 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. Inspection Procedure 82001 Attachment 4 contains guidance for the inspection 

of EP related equipment and facilities that may be useful. 

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

to determine whether the decisions appropriately and conservatively considered 

safety. 
Issue Date: XX/XX/XX 95003 -35-

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. 

(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. 
Issue Date: XX/XX/XX 95003 -36-

(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 

(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? 
Issue Date: XX/XX/XX 95003 -37-

(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. 

(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. 
Issue Date: XX/XX/XX 95003 -38-

(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 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 air-operated 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. 

6. No specific guidance provided. 

g. Key Attribute - Emergency Response Organization Readiness.

The guidance of Inspection Procedure 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
Issue Date: XX/XX/XX 95003 -39-

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 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. 

Plant staffing levels have been reduced during the 1990's, so 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. 
Issue Date: XX/XX/XX 95003 -40-

(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 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 person-rem). 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, 
Issue Date: XX/XX/XX 95003 -41-

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 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. 

The techniques above involve cobalt source reduction, preconditioning of out-ofcore surfaces, control of crud transport (water chemistry control), and chemical 

decontamination. 
Issue Date: XX/XX/XX 95003 -42-

Licensees should not be expected to implement a method for reducing out-ofcore 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 (personrem) 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. 

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. 
Issue Date: XX/XX/XX 95003 -43-

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 a 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 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 SECY89-326 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, 10CFR 50.120 (training rule) requires HPTs 

(including contractors) to be task qualified for their assigned normal and outage 

duties. 
Issue Date: XX/XX/XX 95003 -44-

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 -arounds, 

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 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). 
Issue Date: XX/XX/XX 95003 -45-

(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 decision-making 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. 

(2) Operational decisions and their bases are communicated. 

(3) Interdisciplinary input and reviews of safety-significant or risk-significant 

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 
Issue Date: XX/XX/XX 95003 -46-

(with or without approval) to evaluate whether these personnel feel free 

to report whether they or others are fatigued. 

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 Part 

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: 

(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. 
Issue Date: XX/XX/XX 95003 -47-

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 Part 

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. 

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. 
Issue Date: XX/XX/XX 95003 -48-

(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. 

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 
Issue Date: XX/XX/XX 95003 -49-

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 Inspection Procedure 42700, 

APlant Procedures.@

(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 Part 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

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 
Issue Date: XX/XX/XX 95003 -50-

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. By discussion with selected personnel, review of 

training lesson plans, and completed training records, determine if the 

requirements of 10 CFR Part 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. 
Issue Date: XX/XX/XX 95003 -51-

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 - Access Authorization (AA) 

1. Using data from the licensee=s corrective action system, and Inspection 

Procedures IP 71130.01, AAccess Authorization Program@ and 71152, 

AIdentification and Resolution of Problems,@ 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 Part 50.54(p). See IP 

71130.01 for further guidance. 

b. Key Attribute - Access Control

1. Using IP 71130.02, AAccess Control (Search of Personnel, Packages, and 

Vehicles: Identification and Authorization),@ 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. 

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

design, and support elements)

1. Using IP 71130.03, AResponse to Contingency Events,@ 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: XX/XX/XX 95003 -52-

d. Key Attribute - Response to Contingency Events (performance-based force-on-force 

Exercises and Target Set Evaluation)

Using IP 71130.05, Response to Contingency Events (Force on Force Exercises 

and Target Set Evaluation), the inspectors should request that the licensee explain 

their target set analysis and conduct force-on-force evaluated exercises. See IP 

71130.05 for guidance. 

e. 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. 

Without the use of an NRC endorsed industry 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 of this type. 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 effects of NRC’s 

presence during assessment activities on participants’ behavior and 

consequently the results. 

 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: 

Comment [rag13]: Recommendation

s PV-8 and PV-9. 
Issue Date: XX/XX/XX 95003 -53-

 (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 terms 

such as safety culture attributes or principles, but the concepts addressed should 

be similar. 

 2. (a) It is important to verify that adequate samples of functional groups and 

organizational levels was 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. 
Issue Date: XX/XX/XX 95003 -54-

 (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 

fall between 60% to 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 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 section 06.07 of IMC 

0305. 

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. 
Issue Date: XX/XX/XX 95003 -55-

(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 licensee=s observations in proper context. (For 

related guidance, see Enclosures 95003.02-D.) 

(f) If the assessment included a structured survey, then evaluate the survey 

instrument used, a sampling of raw survey data including write-in 

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 third-party contractors 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 safety culture, in particular at nuclear facilities, 

and appropriate knowledge and experience in implementing safety 

culture/organizational assessment activities. 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: XX/XX/XX 95003 -56-

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. If the licensee and 

third-party contractor appropriately address NRC concerns and comments on the 

methodology, the scope and depth of the NRC’s graded safety culture assessment 

can be adjusted accordingly in recognition of the resolution of the NRC concerns. 

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). 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 
Issue Date: XX/XX/XX 95003 -57-

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 focus on 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 response rates, verify the thirdparty assessment team’s conclusions about the reasons for the low 

participation and evaluate the licensee’s response, such as 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 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, 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, 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) 
Issue Date: XX/XX/XX 95003 -58-

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 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. 
Issue Date: XX/XX/XX 95003 -59-

 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 communication methods 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 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-to-know 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 
Issue Date: XX/XX/XX 95003 -60-

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 third-party assessment of safety culture are 

consistent with 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 the different 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 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 should have 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’s data and the licensee’s 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 
Issue Date: XX/XX/XX 95003 -61-

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). 

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 

IP95001 or IP95002) that may have been performed by the licensee or others, the team 

should group the identified 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. 

The senior reactor analyst should perform a detailed assessment of the individual and 

collective risk associated with team’s findings. 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. The senior reactor analyst should 

use the guidance of LIC-504, Integrated Risk-Informed Decision-making Process for 

Emergent Issues (ADAMS ML070440213), Steps 4.4.3 and 4.4.4, to perform these 

evaluations. 

If the team utilizes Unresolved items to document the issues, a more complete assessment 

of the issues regarding characterization as performance deficiencies and significance 

determination will be performed at a later date. 

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 

Significance Determination Process (SDP) for risk significance, and was this 

data appropriately entered into the NRC action matrix. 

03.12 Document Inspection Results

Deleted: 03.07 Review the 

Licensee=s Independent Safety 

Culture Assessment [June]¶

¶

The review of the licensee=s safety 

culture assessment will occur in two 

steps. The first step is primarily 

focused on how the licensee=s team 

completed their assessment (e.g., the 

methods used, sampling strategies, 

team qualifications, and the use of 

standardized safety culture 

assessment protocols that have been 

endorsed by the NRC). [integrate 

results from ROP feedback form 

95003-1234] The second step of the 

review involves independently 

verifying the validity of the findings 

and conclusions of the licensee=s 

assessment. This second step is 

performed concurrently with the 

NRC=s independent assessment of 

the licensee=s safety culture, by 

conducting interviews and/or focus 

groups with some of the individuals 

who also participated in the licensee=s 

assessment or other representatives 

of the same functional groups. After 

the NRC team has collected and 

analyzed the data from its 

independent assessment, the findings 

and conclusions from the NRC=s 

assessment will be compared to 

those of the licensee=s assessment. ¶

¶

a. Preparation¶

¶

1. The tools and instruments used to 

conduct the licensee=s independent 

safety culture assessment(s) could 

include (but are not limited to) survey 

forms and questionnaires, interview 

guides, checklists, and the charter for 

the assessment(s).¶

¶

2. Documents produced by the 

assessment team that conducted the 

licensee=s most recent safety culture 

assessment could include (but are not 

Comment [rag14]: Recommendation 

PV-5 and PV-6. 

Deleted: Using the results from this 

inspection, in conjunction with 

information obtained from the NRC=s 

review of previous root cause 

analyses that may have been 

performed by the licensee or others, 

attempt to group the apparent causes 

of the risk significant performance 

issues. The issues should be 

Deleted: ¶

The senior reactor analyst should 

perform a detailed assessment of the 

individual and collective risk 

associated with team=s findings. 

Comment [rag15]: Recommendation 

PV-12.

Deleted: F

Deleted: were

... [2]

... [3]
Issue Date: XX/XX/XX 95003 -62-

Due to the diagnostic nature of this inspection, a thorough documentation of the 

team=s observations, Unresolved Items, 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. Any apparent violations or findings that are 

potentially greater than green should receive priority treatment for SDP risk analysis. 

Development of green findings may be deferred until after the cross-cutting aspects 

and diagnostic conclusions have been formulated by the team. If time does not 

permit complete development of the green findings, they may be turned over to the 

Regional staff as unresolved items (URIs) for completion and incorporation into a 

baseline inspection report. 

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 

cross-cutting aspect is assigned in accordance with IMC0612 to findings and URIs 

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 cross-cutting 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 and URIs which were identified during this assessment. 

3. Assessment Result 

Document a summary assessment of licensee performance in each 

degraded SPA, with reference to the observations, findings, and URIs

which support the assessment. 

Deleted: s

Deleted: include the team=s 

assessment of licensee performance 

for each inspection requirement of the 

key attributes contained in the section 

of this procedure.

Deleted: T

Deleted: documentation 

Deleted: s

Deleted: While 

Comment [rag16]: 

Recommendation PV-12. 

Deleted: 

Deleted: each

Deleted: each 

Deleted: each 

Deleted: 

Deleted: and
Issue Date: XX/XX/XX 95003 -63-

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 and URIs 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, findings, and URIs 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. 

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 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/URIs 

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 any 

safety culture components. If the team’s assessment of a safety culture 

Deleted: and 

Deleted: Assessment 

Deleted: Describe the assessment 

methods used, and the documents 

and records reviewed, personnel 

interviewed, walkdowns conducted, 

activities observed, etc., to complete 

this assessment.¶

Deleted: Document the aggregated 

results relative to the consistencies 

derived from the various datacollection tools (described in the 

Enclosures), in conjunction with 

inspectors/assessors insights.

Deleted: , along with 

Deleted: the results of the 

performance deficiency causal 

analysis and evaluation of the 

associated

Deleted: assigned

Deleted: cross-cutting aspects 

assigned to the team’s observations 

and findings; and, the accompanying 

insights about the licensee’s safety 

culture 

Deleted: 

Deleted: components 

Deleted: that were obtained from 

those consistencies. ¶
Issue Date: XX/XX/XX 95003 -64-

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 onsite 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 95003 inspection per year.

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 0-40 

 

Review of Assessment Process 40 (not direct inspection) 

* Including a remedial exercise in the scope of Attachment 95003.01 will require an 

additional 40 hours, resulting in a total of 200. 

 END 

Attachment: 

95003.01 Emergency Preparedness 

95003.02 Guidance for Conducting a Full NRC Safety Culture Assessment

Attachment 1 Revision History For IP 95003

Attachment 95003.02 

Deleted: Document a summary 

assessment of safety culture

Deleted: based on insights from 

assessments of each safety culture 

component

Deleted: , highlighting significant 

weaknesses that are found to exist in 

any safety culture component or 

functional/organizational area that is 

supported by the observations, 

findings and Unresolved Items 

revealed during the assessment. 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), 

the safety culture assessment can 

document the component 

assessment by reference. ¶

Deleted: ¶

80

Deleted: ¶

360

Comment [rag17]: Recommendation 

PV-28. 
Issue Date: XX/XX/XX 95003 -65-

Guidance for Conducting an Independent NRC Safety Culture Assessment 

Introduction

The purpose of this assessment is for the NRC to assess the licensee=s safety culture. 

Safety culture is defined as Athat assembly of characteristics and attitudes in 

organizations and individuals which establishes that, as an overriding priority, nuclear 

plant safety issues receive the attention warranted by their significance.@ Therefore. an 

organization=s characteristics (i.e., safety culture components that comprise the visible 

aspects of a safety culture) can be assessed by evaluating the extent to which its 

policies, programs, and processes ensure that nuclear safety issues receive the 

attention warranted by their significance. For example, the effectiveness of the 

licensee=s corrective action program at identifying, prioritizing, and resolving issues with 

nuclear safety impacts provides important insights into the licensee=s safety culture. An 

organization=s members= shared attitudes and behaviors with respect to nuclear safety 

also provide important insights into a licensee=s safety culture and can be assessed 

through behavioral observations, interviews, and focus groups. 

The guidance in this attachment is intended to enable inspectors to identify those 

consistencies in attitudes and behavior that are indicative of safety culture. 

Relevant attitudes with respect to nuclear safety include: 

• specific attitudes (i.e., an individual=s tendencies to react favorably or 

unfavorably) towards different characteristics of the organization, which may 

include general attitudes about the organization as a whole, the effectiveness of 

the job performance evaluation system in encouraging taking responsibility for 

nuclear safety, the effectiveness of the work control system in scheduling work 

activities safely and efficiently, or the procedures and work packages the 

individual is given to assist in performing tasks; • perceptions, which are how an individual interprets information about the 

organization to form beliefs; and • values, which are an individual=s judgments about what is important, meaningful, 

and worthwhile at work both to the individual and to the organization. 

Consistencies in attitudes are typically identified by interviewing an organization=s 

members to elicit their specific attitudes, perceptions, and values, and their views of the 

organization=s values and attitudes as they relate to nuclear safety. Consistencies in 

attitudes may also be identified by asking individuals to provide examples of situations 

and experiences that are consistent with the attitudes, perceptions, and values they 

describe. If a large proportion of an organization=s members share the same specific 

attitudes, perceptions, and values, these become a social Afact@ within the safety culture 

of the organization and can influence individuals= subsequent decisions and actions. 

Behavioral consistencies with respect to nuclear safety include: 

• observable behaviors (e.g., how often supervisors mention safety considerations 

during pre-job briefs for jobs that may impact nuclear safety, how often personnel 

use procedures in the manner required), 

• written communications (e.g., how often do annual performance reviews mention 

individuals= decisions and actions related to nuclear safety in the past year; how 

often do the forms from a management feedback program note unsafe acts or 

favorably record desirable safety behaviors; how often do email communications 

from managers and supervisors emphasize production or safety goals), and 
Issue Date: XX/XX/XX 95003 -66-

• verbal communications (e.g., how often do supervisors and managers mention 

deadlines and schedules during conversations or in meetings compared to the 

number of times they mention nuclear safety considerations). 

Consistencies in behavior with respect to nuclear safety are typically identified by 

observing an organization=s members going about their daily work as well as reviewing 

written records of decisions made and work that was previously performed. Behavioral 

consistencies can also be identified by asking an organization=s members questions that 

focus on their perceptions of the organization=s norms. Norms are an organization=s 

underlying, usually unwritten, rules for behavior that establish Ahow we do things around 

here,@ and they may or may not coincide with the organization=s stated policies and 

procedures. 

Therefore, the safety culture assessment will focus on the use of tools and instruments 

(i.e., information-collection methods) that will help the inspection team identify these 

consistencies in organizational attitudes and behavior through the use of interviews, 

structured behavioral observations, document reviews and analysis, and case studies. 

1. Preparation for an Independent NRC Safety Culture Assessment 

a. From the list of inspection requirements in Enclosure 95003.02-A, select the 

requirements that relate to the performance deficiencies that prompted this 

inspection, and, to the extent possible, adapt the selected requirements to focus on 

those performance deficiencies. Ensure that the selected requirements include at 

least one requirement associated with each safety culture component. A specific 

performance deficiency may or may not be the result of a weak safety culture. As 

input to the safety culture assessment, inspectors/SCAs should independently 

determine whether weaknesses in one or more safety culture components played a 

causal or contributing role in each performance deficiency. 

b. Identify the SCA lead and SCA subteam who will conduct the safety culture 

assessment, and ensure that at least two different SCAs (or other inspection team 

members as appropriate) independently collect information from each 

functional/organizational group. The safety culture components detailed in section 

06.07 of IMC 0305 describe organizational characteristics and consistencies in 

attitudes and behavior with respect to nuclear safety that are indicative of safety 

culture. 

 (1). When developing the tools and instruments and selecting the methods for 

this assessment, include opportunities to look for evidence of all of these 

components. 

(2). When planning focus group interviews, assign two different SCAs to lead the 

interviews with different groups of participants from the same functional area. 

 If there is an insufficient number of licensee staff within a functional area to 

form two separate focus groups (e.g., chemistry), consider combining focus 

group participants from more than one functional area or use individual 

interviews instead of focus groups. 

(3). Assign two different SCAs to conduct structured behavioral observations of 

the selected work activities within a functional area. 

(4). Establish a plan for communication between and coordination among the 

SCAs and with the other inspection team members. 
Issue Date: XX/XX/XX 95003 -67-

c. Obtain access to the following documents (ensure they are not duplicates of that 

which has already been requested by team leader): 

(1). The root cause investigations of the performance deficiencies that prompted 

this inspection. 

(2). Reports of any self or third-party assessments related to the components of 

safety culture from the past five years. These reports may include: 

 -safety culture assessments completed either by the licensee or a third-party. 

 -safety conscious work environment assessments. 

 -leadership assessments (redacted, as necessary). Information gathered 

about an individual manager that provides evidence of leadership 

ineffectiveness or staff dislike of a particular leadership style should not be 

considered when developing safety culture insights. However, if there is 

evidence of generally ineffective leadership or style which demonstrates 

organizational attitudes and behaviors that are inconsistent with those 

described in the safety culture components, this should be noted. 

 -employee morale/job satisfaction assessments. Information gathered that 

provides evidence of low employee morale or low job satisfaction should not 

be considered when developing safety culture insights, except if the low 

morale or job satisfaction significantly contributes to weaknesses in the 

safety culture components. 

 -Nuclear Quality Assurance/Oversight assessments related to the 

components of safety culture. 

 -any focused or broad-scope assessments of organizational factors. 

(3). Review corporate and site safety policy statements as they relate to safety 

culture. 

(4). A sample of redacted job performance reviews from each functional group in 

the organization (e.g., operations, maintenance, security, engineering) and 

any (redacted) agreements or documents related to the bases for 

management compensation and incentives. Review the sample to determine 

if there is evidence that the actual reviews implemented the guidance, 

especially with regard to the balance between safety and production. 

 (5). Meeting minutes from the past year for site senior management team 

meetings, nuclear oversight review group meetings, and corrective action 

review group meetings; meetings to develop and amend site financial plans 

and budgets, including operating, maintenance, and capital improvement 

plans; meetings that focus on decision-making with nuclear safety 

implications; and other meetings held to plan and discuss mitigating any 

potential or actual chilling effects from disciplinary actions. Review for 

evidence that safety has priority, e.g., for safety over costs in making 

improvements or maintaining systems. 

 (6). Documents describing any reward or incentive programs focused on 

promoting nuclear safety behaviors and documentation pertaining to the 

implementation of the program(s) (redacted, if necessary). 
Issue Date: XX/XX/XX 95003 -68-

 (7). Lesson plans used to train site personnel on safety culture and/or safety 

conscious work environment, and records that show when the training was 

presented and attendance. Review training lesson plans and records to 

determine what was presented, when it was presented, and who attended. 

 (8). Summaries of documents from the corrective action program that relate to 

the components of safety culture and were identified or resolved within the 

previous year. Review and code them to the applicable safety culture 

components. 

 (9). Complete and current organizational charts, including the names and site 

contact information for the individuals listed. 

 (10). Written communications (e.g. memoranda, e-mails) between management 

and staff related to any significant organizational changes within the past 

year. Significant organizational changes could include changes in 

organizational structures and functions, leadership, policies, programs, 

procedures, and resources. 

 (11). Documents that correspond to the selected inspection requirements in 

Enclosure 95003.02-A. 

d. From NRC sources, obtain and review: 

 (1). Allegations related to the site=s safety culture received within the previous 

year, at a minimum; and 

 (2) Inspection reports that document the performance deficiencies which 

prompted this inspection. 

e. Review the results of the licensee=s third-party safety culture assessment to 

determine if particular functional/organizational groups were identified by the SCAs 

as having problems in any of the safety culture components or if the results indicated 

that there were weaknesses in any of the safety culture components across the site. 

f. From the review of the documentation, determine if any functional groups, 

management levels, or safety culture components should receive more emphasis in 

the assessment based on the licensee=s assessment having identified safety culture 

or performance issues related to them. 

g. Determine the assessment methods that are best suited to the perceived situation at 

the site, ensuring that each safety culture component will be assessed with at least 

two different methods, and develop sampling plans for each method. Assessment 

methods shall include individual and/or focus group interviews (see guidance in 

Enclosure 95003.02-B and Enclosure 95003.02-C); structured behavioral 

observations (see guidance in Enclosure 95003-D.02); and event follow-up studies 

(see guidance in Enclosure 95003.02-E). 

h. Prepare the selected data-collection tools, which may include interview and focus 

group guides and behavioral observation checklists. 

i. Work with the licensee to identify the appropriate means to disseminate a 

communication plan to site personnel that addresses the purpose of the 

assessment; states the team will meet with groups, observe meetings and work 

activities, and talk with individuals; states anyone who wants to talk with the NRC 

should contact the team (provide appropriate instructions); describes, to the extent 
Issue Date: XX/XX/XX 95003 -69-

possible, information obtained during the assessment will not be attributed to 

individual participants who are interviewed or observed by NRC inspectors; and 

requests sensitivity to information shared between personnel during participation in 

NRC activities, such as focus groups. 

2. Conduct the NRC=s Independent NRC Safety Culture Assessment 

a. Complete this assessment by applying the methods and sampling plans, using the 

data-collection tools developed during the preparation phase. 

b. As the assessment progresses, the lead SCA should adjust the assessment plan as 

required to: 

 

(1) ensure that the information collected will be adequate to reach valid and 

reliable insights about the safety culture components at each management 

level and in each functional group that falls within the scope of the 

assessment; 

(2) resolve inconsistencies identified in NRC assessment results; and 

(3) address emergent issues identified during other inspection or assessment 

activities. 

Adjustments to the assessment plan may include increasing the number of 

interviewees or focus group participants, conducting additional individual interviews, 

adding documents to be reviewed, increasing the number and focus of behavioral 

observations, re-directing resources to complete a case study, and/or increasing the 

number of team members collecting related information. 

c. Ensure that each safety culture component is assessed using at least two datacollection methods, by at least two SCAs (or other inspection team members as 

appropriate) independently, and that the data-collection methods are applied 

consistently. In this context, valid insights are based on consistent results from 

applying multiple information-gathering methods, and reliable insights are based on 

consistent results from multiple team members who are independently collecting 

related information. 

d. Identify weaknesses in any safety culture component and the functional groups and 

management levels in which any weaknesses appear. As examples, identify any 

performance deficiencies for which safety culture weaknesses were a root or 

contributing cause. 

e. As results from the various data-collection tools are being compiled, and particularly 

after all results have been obtained, aggregate those results to determine whether 

any consistency regarding a particular safety culture component exists among 

results obtained from various data-collection tools and inspectors. From this 

determination, develop insights about the various safety culture components and 

how they are reflected in the attitudes and behaviors within the various licensee 

functional groups. 

 The insights of most concern include the following: 

 $ Corporate and/or senior site management demonstrates attitudes and 

behavior with respect to nuclear safety that are substantively inconsistent 

with the expectations in any of the safety culture components. 
Issue Date: XX/XX/XX 95003 -70-

 $ A single critical functional group, including operations, engineering, 

maintenance, or security, demonstrates weaknesses across multiple safety 

culture components. 

 $ The majority of functional groups demonstrate some weaknesses in multiple 

safety culture components (i.e., weaknesses are widespread throughout the 

organization). 

If any of these statements are supported by consistent results obtained through 

application of multiple data-collection tools in conjunction with inspector/SCA insight, 

then: 

 $ Use this information to inform the assessment of the contributors to 

degraded performance in the affected SPA(s). For instance, the results of 

this assessment can be reflected in statements such as Aweaknesses in a 

safety culture component (or group of components) contributed to 

deficiencies in (some SPA Key Attribute),@ or Aweaknesses exhibited by the 

operations organization related to (some specific list of safety culture 

components) contributed to the degraded Reactor Safety SPA (or Key SPA 

Attribute).@

 $ Document each such statement as a key result of this assessment. Follow 

the requirements in section 02.12. and associated guidance in section 03.12. 

to perform the documentation in coordination with the rest of the inspection 

team. 

Attachment 95003.02 Enclosures 

Enclosure 95003.02-A: “Sample Inspection Requirements for Safety Culture 

Components.” 

Located at: https...., or access through NRC’s Agencywide Document Management 

and Access System (ADAMS) using MLxxxx 

Enclosure 95003.02-B: “Sample Questions for Safety Culture Components.” 

Located at: https...., or access through ADAMS using MLxxxx 

Enclosure 95003.02-C: “Guidance for Focus Groups and Individual Interviews.” 

Located at: https...., or access through ADAMS using MLxxxx 

Enclosure 95003.02-D: “Guidance for Structured Behavioral Observations.” 

Located at: https...., or access through ADAMS using MLxxxx 

Enclosure 95003.02-E: “Guidance for Event Follow-up Studies.” 

Located at: https...., or access through ADAMS using MLxxxx 

Enclosure 95003.02-F: “Guidance for Evaluating Safety Culture Surveys.” 

Located at: https...., or accessed through ADAMS using MLxxxx 

Deleted: Enclosure 95003-A¶

Sample Inspection Requirements 

for Safety Culture Components¶

¶

This enclosure contains sample 

inspection requirements from which 

safety culture assessors may select 

and adapt inspection requirements 

related to performance deficiencies, 

as required by section 95003.02.08.a. 

 This enclosure also identifies 

documents which correspond to the 

requirements, to assist safety culture 

assessors in compiling a list of 

documents to request from the 

licensee. Hence, this enclosure is a 

resource that safety culture assessors 

may use to develop the assessment 

section of the inspection plan.¶

¶

In this enclosure, sample inspection 

requirements and corresponding 

documents are listed in the twocolumn table below, under each of the 

safety culture components: column 

one includes the inspection 

requirements associated with each 

component, while the second column 

describes the corresponding 

documents which should be 

requested from the licensee.¶

¶

¶

ACCOUNTABILITY - Management 

defines the line of authority and 

responsibility for nuclear safety. 

Specifically (as applicable):¶

 ¶

<#>Accountability is maintained for 

important safety decisions in that the 

system of rewards and sanctions is 

aligned with nuclear safety policies 

and reinforces behaviors and 

outcomes that reflect safety as an 

overriding priority. ¶

<#>Management reinforces safety 

standards and displays behaviors that 

reflect safety as an overriding priority. 

 ¶

<#>The workforce demonstrates a 

proper safety focus and reinforces 

safety principles among their peers. ... [4]
Issue Date: XX/XX/XX Att 1-1 95003 

Attachment 1 

Revision History For IP 95003 

Commitment Tracking Number 

Issue Date Description of Change Training Needed Training 

Completion Date 

Comment Resolution Accession Number 

C1 10/26/06 

CN 06-031 

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 July 1, 2006 ML062980489 

TBD XX/XX/XX 

CN 08-XXX 

Procedure revised to incorporate the results of the ROP safety culture lessons learned evaluation (reference SECY XX-XXX). Changes made for ROP feedback forms...... 

TBD TBD MLXXXXXXXXXX 
Issue Date: XX/XX/XX Att 1-2 95003