NRC INSPECTION MANUAL ARCB

INSPECTION MANUAL CHAPTER 1245, APPENDIX C3

HEALTH PHYSICS INSPECTOR 

TECHNICAL PROFICIENCY 

TRAINING AND QUALIFICATION JOURNAL

Effective Date: 02/20/2017
Issue Date: 02/20/17 i IMC 1245, Appendix C3

TABLE OF CONTENTS

Introduction.................................................................................................................................1

Required Health Physics Inspector Training Courses..................................................................1

Prerequisites and Documentation ...............................................................................................1

Post-Qualification and Refresher Training...................................................................................2

Health Physics Inspector Individual Study Activity.......................................................................3

(ISA-HP-1) Code of Federal Regulations (CFR) ..............................................................3

(ISA-HP-2) Licensee Documents for Health Physics Inspectors ......................................6

(ISA-HP-3) Radiological Hazard Assessment and Exposure Controls .............................9

(ISA-HP-4) Occupational ALARA Planning and Controls...............................................17

(ISA-HP-5) In-Plant Airborne Radioactivity Control and Mitigation .................................24

(ISA-HP-6) Occupational Dose Assessment..................................................................30

(ISA-HP-7) Radiation Monitoring Instrumentation ..........................................................34

(ISA-HP-8) Radioactive Gaseous and Liquid Effluent Treatment...................................40

(ISA-HP-9) Radiological Environmental Monitoring Program .........................................46

(ISA-HP-10) Radioactive Solid Waste Processing and Radioactive Material Handling, 

Storage, and Transportation ..........................................................................................51

(ISA-HP-11) Significance Determination Process - Occupational Radiation Safety .......56

(ISA-HP-12) Significance Determination Process - Public Radiation Safety: Radioactive 

Material Control .............................................................................................................60

(ISA-HP-13) Significance Determination Process Public Radiation Safety: Effluent 

Release Program ..........................................................................................................63

(ISA-HP-14) Significance Determination Process - Public Radiation Safety: 

Environmental Monitoring Program ...............................................................................67

(ISA-HP-15) Significance Determination Process - Public Radiation Safety: 

Transportation Branch...................................................................................................70

(ISA-HP-16) Performance Indicators - Occupational Radiation Safety Cornerstone –

Occupational Exposure Control Effectiveness and Public Radiation Safety Cornerstone -

Radiological Effluent Technical Specifications (RETS)/Offsite Dose Calculation Manual 

(ODCM) Radiological Effluent Occurrences...................................................................75

Health Physics Inspector On-the Job Activities .........................................................................78

(OJT-HP-1) Radiological Hazard Assessment and Exposure Controls ..........................78

(OJT-HP-2) Occupational ALARA Planning and Controls..............................................83

(OJT-HP-3) In-Plant Airborne Radioactivity Control and Mitigation ................................87

(OJT-HP-4) Occupational Dose Assessment.................................................................90

(OJT-HP-5) Radiation Monitoring Instrumentation .........................................................92

(OJT-HP-6) Radioactive Gaseous and Liquid Effluent Treatment..................................95

(OJT-HP-7) Radiological Environmental Monitoring Program ........................................99

(OJT-HP-8) Radioactive Solid Waste Processing and Radioactive Material Handling, 

Storage, and Transportation ........................................................................................103

Health Physics Inspector Technical Proficiency Level Signature Card and Certification..........107

Form 1: Health Physics Inspector Technical Proficiency Level Equivalency Justification .......109

Revision History Table ........................................................................................................ Att1-1
Issue Date: 02/20/17 1 IMC 1245, Appendix C3

Introduction

Each inspector will be assigned a supervisor or a qualified inspector to assist with this training. 

Each inspector must compete the Inspector Manual Chapter (IMC) 1245, Appendix A, Basic 

Level Training and Qualification Journal before beginning this IMC 1245, Appendix C3, with the 

exception of the courses identified below. However, an inspector may complete the General 

Proficiency requirements contained in IMC 1245, Appendix B together with the Technical 

Proficiency requirements outlined in Appendix C3.

There are 16 Individual Study Activities (ISA) and eight on-the-job (OJT) study assignments. 

Several of the ISAs have an associated OJT activity (e.g., ISA-HP-3 and OJT-HP-1). You must 

complete the individual ISA before beginning the corresponding OJT activity. However, it is not 

necessary to complete all the ISAs before beginning the OJTs. Consider discussing potential 

strategies for completing your qualification activities with your supervisor. 

The reference documents listed in the Reference section of the ISA may be found on the 

Radiation Protection and Consequence Branch (ARCB) SharePoint site 

(http://fusion.nrc.gov/nrr/team/dra/arcb/radprotection/default.aspx). The ANSI Standards may 

be obtained through the NRC electronic Library at https://login.ihserc.com/login/erc. 

Required Health Physics Inspector Training Courses

For the two courses listed below, completion of IMC 1245, Appendix A, is recommended, but 

the trainee’s branch chief can override this recommendation based on the trainee’s experience.

R-104B, GE Technology

R-104P, Westinghouse Technology

The courses listed below do not require the completion of Appendix A, but one must complete

course prerequisites.

E-110S, Power Plant Engineering, Self-study of chapters: 1.0, 2.0, 3.0, 4.0, 6.0, 14.0, 15.0, and 

16.0

H-308, Transportation of Radioactive Materials

H-311, Respiratory Protection

H-111, Environmental Monitoring for Radioactivity

H-201, Health Physics Technology

H-202S, Radioactive Waste Management, Self-Study Course

Prerequisites and Documentation

Some of the references listed in each ISA are underlined for emphasis because they are directly 

associated with the ISA training module’s evaluation criteria or are needed to complete a 

task. The inspectors are expected to review these references as necessary to adequately 

address the topics covered in the evaluation module. Mentors are expected to provide guidance 

on how best to use these references during the performance of the ISA.

References that are not underlined are provided to give inspectors additional information related 

to the topic of the ISA. Inspectors should discuss with their mentors which of these references 

may provide additional insights into the performance of the ISA.
Issue Date: 02/20/17 2 IMC 1245, Appendix C3

Independent Study Activities should be completed (in office) prior to performing the 

corresponding OJT at a licensee site. However, there are in-office aspects of the OJTs that 

could be completed concurrently with the related ISA.

OJT tasks are to be performed concurrent with an inspection at an operational nuclear power 

plant under the direction of a qualified inspector. Any unexpected findings or questions that you 

may have should be brought to the attention of the inspector rather than to a licensee’s 

attention.

An inspector may be given credit for equivalent training or experience. The approval process is 

provided in Section 05.02 of IMC 1245.

Post-Qualification and Refresher Training

This section has been moved to Appendix D-1
Issue Date: 02/20/17 3 IMC 1245, Appendix C3

Health Physics Inspector Individual Study Activity

TOPIC: (ISA-HP-1) Code of Federal Regulations (CFR)

PURPOSE: This ISA will help the inspector develop knowledge of the appropriate 

radiation protection regulatory requirements and learn how to apply these

requirements to radiation protection programs at nuclear power plants. 

The Code of Federal Regulations (CFR) requires that licensees comply 

with those Parts of the CFR that pertain to the possession, use, storage, 

disposal and transportation of radioactive materials. Nuclear power 

reactor licensees, for example, are required to develop, document, and 

implement a radiation protection program commensurate with the scope 

and extent of licensed activities within the plants. The CFR provide the 

content and scope that licensees must comply with or receive an 

exemption from the NRC to deviate from the requirements. 

COMPETENCY

AREA: REGULATORY FRAMEWORK

LEVEL

OF EFFORT: 40 hours

REFERENCES: 1. 40 CFR Part 190

2. 29 CFR Part 1910

3. 49 CFR Parts 170-189

4. 10 CFR Part 19

5. 10 CFR Part 20

6. 10 CFR Part 21

7. 10 CFR Part 30

8. 10 CFR Part 31

9. 10 CFR Part 37

10. 10 CFR Part 34

11. 10 CFR Part 36

12. 10 CFR Part 40

13. 10 CFR Part 50

14. 10 CFR Part 51
Issue Date: 02/20/17 4 IMC 1245, Appendix C3

15. 10 CFR Part 52

16. 10 CFR Part 61

17. 10 CFR Part 70

18. 10 CFR Part 71

19. 10 CFR Part 72

20. 10 CFR Part 74

21. 10 CFR Part 100

22. 40 CFR Part 141

23. NRC Enforcement Policy

24. IMC 0308, “Reactor Oversight Process”

25. IMC 0608, “Performance Indicator Program”

26. IMC 0609, “Significance Determination Process”

27. IP 71124, “Radiation Safety – Public and Occupational”

28. IP 71151, “Performance Indicator Verification”

EVALUATION

CRITERIA: Upon completion of this ISA, the inspector should be able to:

1. Identify, recognize and locate radiation protection topics presented 

in the CFR that are relevant to radiation safety at nuclear power 

plants.

2. Discuss the content of radiation protection requirements identified in 

the CFR.

3. Discuss the definitions of radiation protection terms identified in the 

CFR.

4. Recognize and discuss the regulatory bases for the Safety Analysis 

Report and the Technical Specifications as found in 10 CFR 50. 

Discuss Part 50, Appendix I and discuss how these numerical 

criteria are used in the Reactor Oversight Process (ROP)

5. Relate the requirements of the CFR from Evaluation Criterion 1 to 

the ROP. Discuss how the enforcement process as described in the 

Enforcement Policy is reflected in the implementation of the health 

physics aspects of the ROP. (See NRC Enforcement Policy on the 

NRC Office of Enforcement Web site).
Issue Date: 02/20/17 5 IMC 1245, Appendix C3

6. Relate the occupational and public radiation safety Performance 

Indicators to the requirements in the CFR.

7. Correlate the requirements of the CFR to the Radiation Safety 

cornerstone (See IMC 0308) and the inspection objectives of the 

occupational and public radiation safety inspection procedures 

(71151 and 71124.01, .02, .03, .04, .05, .06, .07, and .08). Cross 

reference the applicable CFR requirements to the inspection 

findings.

TASKS: 1. Locate general and specific radiation protection activities described 

in the CFR. 

2. Compare and contrast the requirements contained in the health

physics inspection procedure IP 71124 to the requirements of the 

CFR.

3. Review the CFR to identify the regulatory bases for the radiation 

protection programs at nuclear power plants. Be aware of the 

Supplementary Information (previously called Statements of 

Consideration) that are published in the Federal Register when a 

rule is promulgated or revised. The Supplementary Information is 

the information that serves as the background for the rule and is 

published with the rule in the Federal Register. An inspector should 

be aware that at the end of each section in the regulations, there are 

citations that reference the federal register notice when the rule was 

published (and corresponding “supplemental information”). The 

Federal Register Notices contain additional information explaining 

how the rule was developed in case there may be questions 

regarding interpretation of a particular section of the rule by the 

licensee. For HP related regulations, one can find most of the 

applicable federal register notices at the Radiation Protection and 

Consequence Branch (ARCB) SharePoint site.

http://fusion.nrc.gov/nrr/team/dra/arcb/radprotection/SharedDocume

ntsLibrary/FederalRegisterNotices. Alternatively, one can search

https://www.federalregister.gov/articles/search.

4. Review the CFR and IMC 0608, “Performance Indicator Program” to 

identify the regulatory significance of Performance Indicators. 

5. Review how the health physics significance determination 

processes are related to the Enforcement Policy. 

6. Meet with your supervisor or a qualified inspector to discuss any 

questions you may have as a result of this ISA. Discuss the 

answers to the questions listed under the Evaluation Criteria section 

of this study guide with your supervisor.

DOCUMENTATION: Document completion on the Health Physics Inspector Technical 

Proficiency Level Signature Card and Certification Item ISA-HP-1.
Issue Date: 02/20/17 6 IMC 1245, Appendix C3

Health Physics Inspector Individual Study Activity

TOPIC: (ISA-HP-2) Licensee Documents for Health Physics Inspectors

PURPOSE: This ISA will help Health Physics Inspectors develop a working knowledge 

of the contents of the Updated Final Safety Analysis Report (UFSAR), the 

Technical Specifications (TS), the Technical Requirements Manual (TRM) 

(or equivalent), the Process Control Program (PCP), and the Offsite Dose 

Calculation Manual (ODCM).

Three other important programs for an inspector to be familiar with are the 

Radiation Protection Program, the Emergency Plan, and the Ground Water 

Protection Program. These documents also implement regulatory 

requirements.

The NRC requires that licensees maintain an Updated Final Safety 

Analysis Report (UFSAR). These are available on the NRC internal 

SharePoint site at 

http://fusion.nrc.gov/nrr/team/dorl/FSARs/Forms/AllItems.aspx. The 

UFSAR describes the design features, systems, programs, and design 

basis operation of the facility. For this reason, it is important that health 

physics inspectors gain a detailed knowledge of certain sections of the

UFSAR that are relevant to Radiation Safety (e.g., Chapters 11, 12, and 

15). Additionally, the NRC requires that licensees operate their facilities in 

compliance with the Technical Specifications (TS) that are available at

Open ADAMS P8 Folder (TECH SPECS). The TS can also be found at 

http://fusion.nrc.gov/nrr/team/dss/stsb/default.aspx.

Each licensee maintains additional required documents such as the TRM

(or equivalent), the PCP, and the ODCM. It is important that Health 

Physics inspectors become knowledgeable of the ODCM, TRM (or 

equivalent), PCP and applicable sections of the TS. 

COMPETENCY

AREA: TECHNICAL AREA EXPERTISE

REGULATORY FRAMEWORK

LEVEL

OF EFFORT: 40 Hours

REFERENCES: 1. Plant Updated Final Safety Analysis Report

2. Technical Specifications

3. Technical Requirements Manual (or site-specific equivalent)

4. Process Control Program

5. Radiation Protection Program

6. Emergency Plan
Issue Date: 02/20/17 7 IMC 1245, Appendix C3

7. Groundwater Protection Plan

8. NEI-07-07, Industry Groundwater Protection Initiative – Final 

Guidance Document,” August 2007

9. Generic Letter 89-01, “Implementation of Programmatic Controls for 

Radiological Effluent Technical Specifications in the Administrative 

Controls Section of the TS and the Relocation of Procedural Details 

of RETS to the ODCM or to the PCP,” January 31, 1989.

10. Licensee Basis Documents

EVALUATION

CRITERIA: Upon completion of this ISA, the inspector should be able to:

1. Identify the applicable sections of the designated facility UFSAR 

(meteorology, engineered safety features ventilation/filtered 

systems, radiation monitoring instrumentation, radiation protection, 

radioactive waste management, technical specifications, and quality 

assurance) and discuss the content.

2. Identify the applicable sections of the designated facility TS 

(definitions, radiation monitoring instrumentation, engineered safety 

features ventilation, radioactive effluents, solid radioactive waste 

(PCP), radiological environmental monitoring, administrative 

controls) and TRM (definitions, limiting conditions for operation and 

surveillance requirement applicability, instrumentation, radioactive 

effluents, radiological environmental monitoring, and administrative 

controls) and discuss the content and basis for the requirements.

3. Discuss the content and basis for the requirements of the 

designated facility ODCM.

4. Discuss definitions and terms found in the designated facility 

UFSAR, TS, TRM, PCP and ODCM.

5. Discuss the legal basis, purpose, license conditions, and how these 

documents (UFSAR, TS, TRM, PCP and ODCM) can be changed.

6. Discuss the delineated programs, processes, equipment, and limits, 

and the reasons they are required.

7. Discuss the requirements for surveillances, action statements, and 

reporting.

8. Discuss the administrative controls section of the TS and the types 

of information located in this section. Focus on high radiation area 

alternative controls and radioactive materials effluent requirements.
Issue Date: 02/20/17 8 IMC 1245, Appendix C3

9. Discuss the TRM (or equivalent) purpose, legal basis, and how it 

can be changed.

10. Discuss the Emergency Plan’s purpose, legal bases, Emergency 

Classifications, Protective Action Guidelines, Post Accident 

Sampling and Radiological Dose Assessment.

11. Discuss the Radiation Protection Plan’s purpose, legal bases and its 

contents.

12. Discuss the Groundwater Protection Program’s purpose, legal 

bases and its contents as recommended in NEI 07-07.

TASKS: 1. Request or suggest a facility be designated by your supervisor or 

mentor.

2. For the facility designated by your supervisor or mentor, locate a 

copy of the following documents and review the various sections as 

listed in the Evaluation Criteria section:

a. UFSAR

b. TS

c. RM (or site-specific equivalent)

d. ODCM

e. PCP

f. Emergency Plan

g. Radiation Protection Plan

h. Ground Water Protection Plan

3. Which license amendment(s) removed c, d, and e from the TS?

4. Meet with your supervisor or qualified inspector to discuss any 

questions you may have as a result of this ISA. Discuss the 

answers to the questions listed under the Evaluation Criteria section 

of this study guide with your supervisor.

DOCUMENTATION: Document completion on the Health Physics Inspector Technical 

Proficiency Level Signature Card and Certification Item ISA-HP-2.
Issue Date: 02/20/17 9 IMC 1245, Appendix C3

Health Physics Inspector Individual Study Activity

TOPIC: (ISA-HP-3) Radiological Hazard Assessment and Exposure Controls 

PURPOSE: The purpose of this ISA is to familiarize you with the regulatory basis and 

historical agency position that provide adequate protection of occupational 

workers from exposure to radiation and radioactive materials. The 

information is addressed in generic correspondence, regulatory guides and 

regulations that pertain to the assessment of occupational radiological 

hazards, controlling access to radiologically significant areas, maintaining 

control over occupational radiation exposures and controlling the release 

of radioactive materials. The related inspection procedure is IP 71124.01, 

“Radiological Hazard Assessment and Exposure Controls.”

The inspection procedure has three main objectives:

1. To review and assess licensee performance in assessing the 

radiological hazards in the workplace associated with license 

activities and the implementation of appropriate radiation monitoring 

and exposure control measures for both individual and collective 

exposures.

2. To verify that the licensee is properly identifying and reporting PIs 

for the Occupational Radiation Safety Cornerstone.

3. To identify those performance deficiencies that were reportable as a 

PI and which may have represented a substantial potential for 

overexposure of the worker.

COMPETENCY

AREA: TECHNICAL AREA EXPERTISE

REGULATORY FRAMEWORK

LEVEL

OF EFFORT: 40 hours

REFERENCES: 1. 10 CFR 19.12

2. 10 CFR 50.120

3. NUREG-1736, “Consolidated Guidance: 10 CFR Part 20- Standards 

for Protection Against Radiation” Review the sections that address 

the following

a. 10 CFR 20.1003

b. 10 CFR 20.1101

c. 10 CFR 20.1203 through 1204

d. 10 CFR 20.1501 through 1502
Issue Date: 02/20/17 10 IMC 1245, Appendix C3

e. 10 CFR 20.1601 through 1602

f. 10 CFR 20.1701 through 1704

g. 10 CFR 20.1902 through 1906

h. 10 CFR 20.2207

4. Regulatory Guide 8.2, “Guide for Administrative Practices in 

Radiation Monitoring,” May 2011, and other revisions as applicable

5. Regulatory Guide 8.25, “Air Sampling in the Workplace,” June 1992, 

and other revisions as applicable

6. Regulatory Guide 8.38, “Control of Access to High and Very High 

Radiation Areas of Nuclear Plants,” May 2006, and other revisions 

as applicable

7. Information Notice 80-22, "Breakdowns in Contamination Control 

Programs,” May 28, 1980

8. Information Notice 82-31, “Overexposure of Diver During Work in 

Fuel Storage Pool,” July 28, 1982

9. Information Notice 84-61, “Overexposure of Diver in Pressurized 

Water Reactor (PWR) Refueling Cavity,” August 8, 1984

10. Information Notice 84-82, “Guidance For Posting Radiation Areas,” 

November 19, 1984

11. Information Notice 85-06, “Contamination of Breathing Air Systems,” 

January 23, 1985

12. Information Notice 88-79, “Misuse of Flashing Lights for High 

Radiation Area Controls,” October 7, 1988

13. Information Notice 90-33, “Sources of Unexpected Occupational 

Radiation Exposures at Spent Fuel Storage Pools,” May 9, 1990

14. Information Notice 90-47, “Unplanned Radiation Exposures to 

Personnel Extremities Due to Improper Handling of Potentially 

Highly Radioactive Materials,” July 27, 1990

15. Information Notice 92-75, “Unplanned Intakes of Airborne 

Radioactive Materials by Individuals at Nuclear Power Plants,”

November 12, 1992

16. Information Notice 97-36, “Unplanned Intakes by Workers of 

Transuranic Airborne Radioactive Materials and External Exposure 

Due to Inadequate Control of Workers,” June 20, 1997
Issue Date: 02/20/17 11 IMC 1245, Appendix C3

17. Information Notice 97-68, “Loss of Control of Diver in a Spent Fuel 

Storage Pool,” September 3, 1997

18. Information Notice 85-92, “Surveys of Wastes before Disposal from 

Nuclear Reactor Facilities,” December 2, 1985

19. Information Notice 86-44, “Failure to Follow Procedures when 

Working in High Radiation Areas,” June 10, 1986

20. Information Notice 88-63, “High Radiation Hazards from Irradiated 

In-core Detectors and Cables,” August 15, 1988

21. Information Notice 90-44, “Dose Rate Instruments Under 

responding to the True Radiation Fields,” June 29, 1990

22. IE Circular 81-07, “Control of Radioactively Contaminated Material,”

May 14 1981

23. IE Circular 76-03, “Radiation Exposures in Reactor Cavities,” 

September 10, 1976

24. IE Bulletin 78-08, “Radiation Levels From Fuel Element Transfer 

Tubes,” June 12, 1978

25. NUREG/CR-5569, Rev. 1, “Health Physics Positions Data Base,”

HPPOS-016, 018, 020, 021, 022, 217, 221, 245, and 250

26. NUREG/CR-6204, “Questions and Answers Based on Revised 10 

CFR Part 20,” Questions 118, 134, 380, 447 and 448

27. Regulatory Guide 1.8, “Qualification and Training of Personnel for 

Nuclear Power Plants,” May 2000, and other revisions as applicable

28. ANSI/ANS 3.1, “Selection, Qualification and Training of Personnel 

for Nuclear Power Plants,” and other revisions as applicable

29. ANSI 18.1, “Selection and Training of Nuclear Power Plant 

Personnel,” and other revisions as applicable

EVALUATION

CRITERIA: Upon completion of this ISA, the inspector should be able to:

1. Discuss the regulatory requirements associated with airborne 

radioactivity areas, radiation areas, high radiation areas, and very 

high radiation areas.

2. Discuss the regulatory requirements associated with surveys and be 

able to describe what may constitute a survey other than 

documented direct meter measurements.
Issue Date: 02/20/17 12 IMC 1245, Appendix C3

3. Discuss the acceptance criteria for training and qualifications for 

radiation workers, radiation protection technicians, and contract 

radiation protection personnel.

4. Discuss and identify free release criteria for contaminated materials 

from RCA to the public domain.

5. Discuss the answers to the questions associated with each scenario 

given in the context of regulatory requirements without using 

licensee procedures.

TASKS: 1. Review each underlined reference and discuss its application with a 

qualified inspector. 

2. Read the scenarios and answer the associated questions.

3. Meet with your supervisor or a qualified inspector to discuss any 

questions you may have as a result of this ISA. Discuss the 

answers to the questions.

Scenario A:

The following is provided in a pre-job ALARA briefing to the work crew and support personnel.

A job has been scheduled that entails the consolidation of highly radioactive filters (stored 

individually in canisters) into a high integrity container (HIC) for disposal. A radiological pre-job 

survey was done earlier this week. The Radiation Protection (RP) tech providing job coverage 

will confirm survey results on entry.

There are approximately 110 used filters that originally had dose rates of up to 40 R/hr contact 

and 3 filters that had dose rates up to 600 R/hr contact. The dose rate in the general area 

where the filters are stored is approximately 25 R/hr with the lowest dose rate in the room being 

1.8 R/hr at the exit door. The room is 8 by 12 feet with 8 feet ceiling. There are 2 installed 

CVCS filters with the housing of the one in service reading 4 R/hr contact on the side away from 

the old filters and 9 R/hr on the side toward the old filters. The out of service filter housing is 2.2 

R/hr contact on the side away from the old filters and about 9 R/hr on the side towards the old 

filters. (The filter element in the out of service filter is unused but full of water.)

Most of the filters were originally stored in polyethylene bags and the bags have generally 

degraded and many are open. The filters are dry. In the room, general area contamination 

levels on horizontal surfaces range up to 200 mrad/hr smearable and vertical surfaces up to 50 

mrad/hr smearable. Alpha contamination levels up to 200,000 dpm/100 cm2 are common. 

Airborne activity in the room is typically about 0.2 DAC. The airborne activity increases to about 

0.4 DAC when someone enters the room to perform a dose rate survey on the in-service filter. 

Room ventilation is poor, air from the room communicates with the corridor through grating used 

to support lead plate shielding on the top of the room. The room is at the same relative 

pressure as the adjacent clean corridor. 

The consolidation and removal of the filters will allow the area to be decontaminated to a 

reasonable level so that valve repairs can be done. The consolidation will be done using 3 
Issue Date: 02/20/17 13 IMC 1245, Appendix C3

personnel: a RP Tech, a Rad Waste Tech and a QC inspector. It is estimated that the task can 

be done in 10 minutes with a total dose expenditure of approximately 1 rem. 

Questions:

1. What are the legally required postings for the room?

2. What are the significant radiological risks associated with the job?

3. What surveys are appropriate given the above scenario?

4. Where should air sampling be done? Why?

5. How will workers be monitored for radiation/internal dose?

6. What protective clothing and equipment will be used for this job?

Scenario A Continued: 

Because of the limited room and high dose rates it was decided to create a contaminated area 

in the corridor outside the room to allow moving to a low dose area(<5 mR/hr) to undress. 

Concerns with high airborne activity potentials led the radiation protection staff to set up a 

portable HEPA ventilation system with the suction “elephant trunk” (flexible 6 inch ventilation 

hose) run in through the open door and the discharge was routed using an “elephant trunk” to 

an area that had a continuous air monitor (Particulate, Iodine, Noble Gas) at the end of the 

corridor. Since there was no practical way to control the breathing zone airborne activity of the 

workers, a powered air purifying respirator (PAPR) (when blower is running, it is a positive 

pressure air supplied full face respirator and when blower is not running, then it is a negative 

pressure air purifying respirator) was chosen to provide an appropriate protection factor, provide 

cooling and reduce worker fatigue.

Questions:

1. Does the door blocked open with the “elephant trunk” pose a potential regulatory 

compliance problem?

2. What are the licensee options with regard to the “elephant trunk?”

3. What would be the best location for the suction of the “elephant trunk”?

4. Can flashing lights be used under these circumstances?

5. What is the protection factor allowed for a PAPR when the blower is running?

6. What is the protection factor allowed for a PAPR when the blower is not running? 

7. If the room is not expected to exceed 1.5 DAC for the 10 minute duration of the job is 

posting of the area as an airborne radioactivity area required?

Assume: 
Issue Date: 02/20/17 14 IMC 1245, Appendix C3

The filters have been removed and the decontamination effort has begun with an expected 

duration of 12 hours. The airborne radioactivity is expected to average 1 DAC over the interval 

of work. 

Questions:

1. Does the room have to be posted airborne?

2. Can credit be taken for the PAPR respirators to avoid posting the room as an airborne 

radioactivity area?

3. Can credit be taken for the PAPR respirator when assigning DAC-Hours for exposure 

tracking? 

Assume:

A painter in an adjacent previously clean area is found to be contaminated, after 

decontamination it is determined by whole body counting that he has received an intake of 13 

DAC-hours (32.5 mrem). 

There is no air sample in the area to confirm the air concentration. It is believed that air diffused 

out of the top of the room into the corridor and was subsequently drawn into the adjacent room 

by normal ventilation flow. 

Questions:

1. What ALARA considerations should have been made in order to reduce the Total 

Effective Dose Equivalent?

2. Is this a Performance Indicator (PI) occurrence?

3. What are the potential violations that could be identified?

4. If potential violations are identified, construct legally valid citations for each identified

potential violation. Present the constructs to either a Senior Health Physics Inspector or 

Supervisor for critique.

Scenario B

The licensee is performing a 10 year in-service inspection of the reactor core support barrel and 

as part of the inspection the Stellite mounting lugs and snubbers will be removed using EDM 

(electrical discharge machining). Due to the dose rates (up to 10,000 R/hr) on the core support 

barrel the work will be done under water in the refueling cavity using vendor provided divers. 

The licensee has built diving platforms which will serve as a tool rest/ support and restrict the 

motions of the diver to surveyed areas that have less than 100 mrem/hr dose rates. The reactor 

vessel is defueled. 

The work will be covered by a special RWP. Diver will be provided with two alarming 

dosimeters in his helmet, and has the full multi-badge complement including wearing extremity 

dosimetry on his fingers and toes. Pocket ion chambers are to be placed on wrists and ankles. 
Issue Date: 02/20/17 15 IMC 1245, Appendix C3

A small underwater survey instrument probe is to be attached to his dominant wrist to provide 

continuous dose rate readings to the RP technician. The RP technician will be maintaining 

constant line of sight on the diver by the use of viewing boxes (i.e., Floating Plexiglas windows 

that eliminate the visual distortion caused by ripples on the water.) All communications with the

diver will be through the Dive Supervisor. The Dive Supervisor has wired 2 way 

communications with the diver. The RP technician can listen in on the circuit. The diver shall 

not enter the water until there is agreement between the RP technician and the Dive Supervisor. 

Both have stop work authority. 

The work area will be surveyed using two independent remote probe instruments at the 

beginning of each shift. The rescue diver will be ready to enter the water with the exception of 

putting on his helmet prior to the diver entering the water. The diver will survey the dive 

platform using the remote probe attached to his wrist at the beginning of each working dive and 

the RP technician shall ensure the diver, rescue diver and Dive Supervisor are aware of the 

dose rates. The survey shall extend, in all directions, as far from the edge of the platform as the 

diver can reach without lifting his feet from the platform. 

Chemistry will sample the water once a shift to determine nuclide content and activity. Tritium

concentration will be analyzed daily. Pool clarity shall be maintained consistent with refueling 

visibility requirements by the use of two 600 gpm underwater filters. One will have a vacuum 

hose attached that can be used to remove any sediment that may accumulate on the dive 

platform. Chemistry controls can be used to supplement the filter system if needed 

(precipitation of colloidal iron with hydrogen peroxide).

Questions:

1. What are the radiological hazards associated with this job?

2. What is the purpose of the dive platforms?

3. Does the RP organization have to post signs under water?

4. Does the RP organization have to post the entire cavity as a Very High Radiation Area 

while diving operations are in progress?

5. Why is the alarming dosimeter in the diver’s helmet instead of on his chest?

6. Why is the water analyzed periodically? Is it a regulatory requirement?

7. Why is extremity dosimetry required? 

8. As it relates to diving operations, define radiation dose gradient and explain and discuss 

the impact of this on the needed controls and monitoring to ensure a diver is adequately 

protected. Compare the dose gradients from a 100 Ci Co-60 point source in air and 

under water.

9. Does the RP technician performing continuous coverage from the cavity walkway meet 

requirements?

11. Are bioassays required and if so what type should be performed?

Scenario C
Issue Date: 02/20/17 16 IMC 1245, Appendix C3

On March 20, 2003, a whole body count vendor operator was conducting whole body counting 

of contract personnel as part of their station in-processing in the General Training and 

Orientation Center (GTOC). The GTOC is located outside of the station’s Protected Area but is 

within the Owner Controlled Area (OCA) of the station. At approximately 11:39 a.m., the first 

contract individual was identified as having a positive whole body count (WBC) with Mn-54, Co58, and Cs-137 (6.3 nanocuries (nCi), 6.4 nCi, and 4.2 nCi, respectively). The vendor operator 

instructed the contractor to remove his lanyard and conducted a second WBC which was 

negative for the presence of measurable radioactivity. Subsequently, the vendor operator 

returned the lanyard to the contractor with the instructions that he discard the old lanyard, since 

it may be mildly contaminated. At approximately 12:12 p.m., a second contractor’s WBC was 

positive for the presence of radioactivity (approximately 9.8 nCi Co-60). The vendor operator 

had the contractor remove his fleece vest and repeated the WBC. The follow-up WBC was 

negative for the presence of measurable radioactivity, and, similarly, the vendor operator 

returned the fleece vest to the contractor with a suggestion that he launder the vest. At 

approximately 1:59 p.m., a third contractor’s WBC indicated possible positive activity with a 143 

keV peak. The contractor indicated that his boots may have been previously contaminated, 

thus the vendor operator performed another WBC of the contractor with his boots removed; the 

follow-up WBC was negative for the presence of measurable radioactivity.

At approximately 2:15 p.m., the Radiation Protection (RP) Instrument Supervisor and a principle 

radiological engineer went out to the GTOC to check on the vendor operator and they were 

subsequently informed about the recent positive WBCs and apparent external contaminations. 

The RP Instrument Supervisor initiated an investigation and was able to take control of the 

contaminated materials/clothing from two of the individuals who were still within the OCA by 

approximately 5:30 p.m. However, the first contractor had apparently left the OCA with the 

externally contaminated lanyard. Radiation Protection management was able to contact the first 

contractor later that evening; the contractor indicated that he was still in possession of the 

lanyard, and RP management requested that he place the lanyard in a bag and bring it into the 

station the following morning.

The licensee’s investigation revealed that the vendor operator was apparently not cognizant of 

the procedural and regulatory requirements to take control of any measurable radioactive 

material outside of the radiologically restricted areas. The licensee additionally identified that 

there was less than adequate vendor oversight by the RP department and procedure 

deficiencies which contributed to the occurrence.

Questions:

1. What Code of Federal Regulations applies?

2. If applicable, what were the violations and by whom?

DOCUMENTATION: Document completion on the Health Physics Inspector Technical 

Proficiency Level Signature Card and Certification Item ISA-HP-3.
Issue Date: 02/20/17 17 IMC 1245, Appendix C3

Health Physics Inspector Individual Study Activity

TOPIC: (ISA-HP-4) Occupational ALARA Planning and Controls

PURPOSE: The purpose of this ISA is to familiarize you with concepts that will be 

important to conducting the related inspection requirements contained in 

IP 71124.02, “Occupational ALARA Planning and Controls.”

As Low As Reasonably Achievable (ALARA) is the central concept of 

occupational radiation protection. In the United States the linear nonthreshold dose response model was selected for its ease of application in 

a regulatory environment. This model assumes that any exposure to 

ionizing radiation constitutes a risk and risk increases proportionally with 

exposure. An outgrowth of this is the concept of maintaining collective 

dose ALARA, thereby minimizing the total risk associated with the use of 

radioactive materials. In order to minimize individual risk, individual 

exposures are also maintained ALARA with the collective exposure taking 

the precedent. ALARA is implicitly an optimization process that attempts 

to maximize the benefit for a given level of total risk. Performance in this 

area is judged on whether the licensee has taken appropriate measures to 

track, and if necessary, to reduce exposures and not whether each 

individual exposure and dose represent an absolute minimum or whether 

the licensee has used all possible methods to reduce exposures. In 

addition, an ALARA analysis that includes engineering controls and health 

and safety risks is performed when elevated concentrations of radioactive 

material in air exist. This analysis should be consistent with maintaining 

the total dose equivalent ALARA. 

The objective of IP 71124.02 is to assess performance with respect to 

maintaining individual and collective radiation exposures as low as is 

reasonably achievable. The inspection will determine whether the licensee 

has an adequate program, including administrative, operational, and 

engineering controls, to maintain occupational exposure ALARA.

COMPETENCY

AREA: TECHNICAL AREA EXPERTISE

REGULATORY FRAMEWORK

LEVEL

OF EFFORT: 40 hours

REFERENCES: 1. 10 CFR 20.1101, 20.2206

2. NUREG 1736, “Consolidated Guidance: 10 CFR 20 - Standards for 

Protection Against Radiation,” Section 3.20.1101

3. Regulatory Guide 8.8, “Information Relevant to Ensuring that 

Occupational Radiation Exposures at Nuclear Power Plants Will Be 

as Low As Is Reasonably Achievable,” June 1978, and other 

revisions as applicable
Issue Date: 02/20/17 18 IMC 1245, Appendix C3

4. Regulatory Guide 8.10, “Operating Philosophy for Maintaining 

Occupational Radiation Exposures as Low As Is Reasonably 

Achievable,” May 1977, and other revisions as applicable

5. NUREG/CR-5569, Rev. 1, “Health Physics Positions Data Base,”

HPPOS-091.

6. NCRP Report 116, “Limitation to Exposure of Ionizing Radiation,”

1993

7. NCRP Report 120, “Dose Control at Nuclear Power Plants,” 1994

8. NUREG-0713, “Occupational Radiation Exposure at Commercial 

Nuclear Power Reactors and Other Facilities”

9. NUREG/CR-6204, “Questions and Answers Based on Revised 10 

CFR Part 20,” Questions 62, 133, 381 and 384

10. Supplementary Information, 10 CFR 20 Final Rule, Subpart B, 

“Radiation Protection Programs," Federal Register Vol. 56, No. 98, 

pages 23366-23367, May 21, 1991

11. EPRI TR-107991, “Radiation Field Control Manual,” October 1997

12. EPRI TR-1003390, Radiation Field Control Manual,” December, 

2004

13. EPRI TR-108737, “BWR Iron Control Monitoring Interim Report,”

December 1998

14. EPRI TR-107566, “Evaluation of PWR Radiation Fields: 1991 –

1996,” March 1997

EVALUATION

CRITERIA: Upon completion of this ISA, the inspector should be able to: 

1. Explain ALARA using a working definition that would be 

understandable to someone not affiliated with nuclear power or its 

regulation

2. Describe the considerations to be taken in the determination if an 

activity or process is ALARA. (Example: Risk from other hazards 

should not be increased substantially in the application of ALARA.)

3. Name the activities which account for the greatest percentage of the 

collective exposure received in a nuclear power plant.

4. Explain where in the life cycle of a nuclear plant the most significant 

ALARA impacts can be achieved.
Issue Date: 02/20/17 19 IMC 1245, Appendix C3

5. Explain why corrosion control is an ALARA consideration and how it 

can affect the overall radiation exposure of the plant staff.

6. Explain the key elements to an ALARA program as described in 

Regulatory Guide 8.8.

7. Explain the NRC philosophy with regard to assigning a set dollar 

value for ALARA planning purposes.

8. Explain the interrelationship between economics and other factors 

such as other risk.

9. Explain the concept of “TEDE ALARA.”

10. Explain the 4 factors that determine the strength of a radiation field.

11. Explain the principles of time, distance and shielding with respect to 

ALARA.

12. Explain the potential impact from use of temporary shielding on 

reactor system safety, and the licensee actions and evaluations 

needed to properly use this shielding.

13. Explain the modes by which radioactive material can be deposited 

in the body.

14. Explain why, it may not be ALARA to use protective gear or perform 

decontamination when work is being performed in a contaminated 

area with the risk of an intake.

15. Describe the role and qualifications of the Radiation Protection 

Manager with regard to the ALARA and radiation protection 

programs. This includes management access, training program 

approval and the program implementation.

16. Describe how the ALARA work planning is a vital, key component in 

the identification of hazards.

17. Describe the benefits of early HP and working craft involvement and 

coordination and how this affects station collective exposures.

18. Describe and discuss the benefits of using employee ALARA 

suggestions, feedback and lessons learned in the planning and 

evaluation process.

19. List the elements of information provided in an RWP, and how this 

information can be effectively used to brief workers before the start 

of the work.

20. Identify the information that should be available during an inspection 

of an ALARA package.
Issue Date: 02/20/17 20 IMC 1245, Appendix C3

21. Discuss the qualifications and requirements for HP technician shift 

coverage and its purpose.

22. Explain how ALARA planning depends on extrapolating the future 

from past experience where successes and failures can be 

identified to aid in future planning.

23. State the regulatory basis of ALARA and identify what condition 

must exist in order to cite a violation on the basis of ALARA. 

TASKS: 1. Read Regulatory Guide 8.8 and complete the following:

2. Define ALARA. 

a. Identify the factors that must be taken into account when 

making ALARA recommendations.

b. Identify the types of activities typically accounting for the 

majority of the exposures.

c. Identify where in the life cycle of a power plant ALARA 

considerations should be taken.

d. Identify how initial material selection and subsequent 

chemistry control regimens to reduce reactor, feed and steam 

generator systems corrosion contribute to the ALARA concept.

e. Identify the key elements of a radiation control program.

f. Determine the NRC defined dollar per rem amount to be used 

in these analyses for occupational exposure (if any)?

g. Determine if an ALARA cost benefit analysis indicates that for 

a given action there is a net benefit does that mean that a 

particular action must be performed?

h. Determine if ALARA applies equally to internal and external 

exposures.

i. List the 4 factors that determine the intensity of a radiation field 

at a given point.

j. List the 3 principles/techniques that are commonly used to 

reduce exposure to radioactive material.

k. List the 4 modes of entry for radioactive materials into the 

body.

l. Determine how the risk of worker skin contamination fits into 

ALARA decision making.
Issue Date: 02/20/17 21 IMC 1245, Appendix C3

m. Determine the roles of the Radiation Protection Manager:

- Where does he normally fit in the reporting chain?

- Why does he normally fit there?

- What are his specific responsibilities with regard to the

ALARA program?

- What is his responsibility with respect to the sites 

radiation protection training program?

n. Identify the purpose of the ALARA briefings and how they 

contribute to the ALARA program.

o. List the typical contents of a radiation work permit as 

described in Regulatory Guide 8.8.

p. List the typical contents and considerations of an “ALARA 

Package.”

q. Determine the purpose of having a Health Physics (Radiation 

Safety or Radiation Protection) technician assigned to each 

operating shift.

r. Determine the purpose of a post-operational debriefing with 

regard to ALARA.

3. Review the regulatory basis, including the supplementary 

information, for the application of the ALARA regulations.

4. Read the scenario provided and answer the associated questions.

5. Meet with your supervisor or a qualified inspector and discuss any 

questions you may have as a result of this ISA. Discuss the 

answers to the questions and your work products listed under the 

Evaluation Criteria section of this study guide with your supervisor.

Scenario A:

The following is provided as part of a post-refueling outage ALARA inspection.

The inspector found that doses for some jobs conducted during Refueling Outage 15 were not 

maintained as low as was reasonably achievable. From the licensee’s Refuel 15 ALARA 

Outage Report, the following examples were noted:

Job/RWP

Estimated 

Dose 

(Rems)

Actual Dose

(Rems)

Scaffolding in the reactor building 13.000 26.125

Remove and install steam generator manway covers 

and inserts

4.234 7.675
Issue Date: 02/20/17 22 IMC 1245, Appendix C3

Steam generator Eddy current testing/tube plugging 18.985 42.295

Health physics support for Refueling Outage 15 3.450 6.279

Spent fuel pool cleanup/diving operations 3.198 6.723

Reactor coolant pump seal removal and 

replacement

3.745 7.176

The licensee experienced fuel pin leakage over the last year that had increased reactor coolant 

activity and the level of transuranic isotopes. The cause of the fuel pin damage had been 

attributed to several foreign objects that were left in the reactor coolant system after Refueling 

Outage 14.

However, the licensee acknowledged that the increase in reactor coolant activity was 

responsible for only approximately 30 percent of the dose overrun. The licensee conducted 

post job reviews and identified additional causes for higher-than-projected doses. Some of the 

causes were common to more than one job. The inspector reviewed the post job reviews, 

received additional explanation of the licensee’s findings from the ALARA supervisor, and 

reached the following conclusions:

Some activities were not scheduled or sequenced optimally to reduce personnel dose. In an 

effort to advance the outage schedule, steam generator work was started three to four days 

earlier than normal, providing less time for radioactive decay. The licensee set up platforms 

around the steam generators while reactor coolant system cleanup was still in progress and 

before steam generator bowl drains were flushed. 

In the original outage schedule, all reactor coolant pumps (RCP) seal work was to occur when 

the steam generator secondary sides were full. However, because all four RCP seals had to be 

worked, this was not possible. To support the revised schedule, some seal work was performed 

with the generators empty. In past outages when this work was conducted, “an orderly process”

was followed by moving from pump to pump. This process resulted in lower personnel dose by 

minimizing tool movement. In Refueling Outage 15, work crews moved from pump to pump as 

the other work allowed. This forced the crews to move their tooling multiple times.

Insufficient mockup training was conducted to familiarize the workers with plant equipment, use 

of tools, and techniques to reduce dose. Workers spent more than the expected staff-hours in 

high dose areas because “the crews were inexperienced” and “used poor ALARA practices.”

Additional mockup training should have been provided to individuals that installed and removed 

steam generator manways and inserts and those that used robotic eddy current equipment.

Communication between radiation protection personnel and contractor personnel was “poor.” 

Radiation protection personnel “seldom” knew job status or the schedule for the upcoming shift 

work. Therefore, they could not plan their activities to reduce dose.

There was a “lack of involvement and ownership” of the scaffolding program by craft 

supervisors. Reviews of scaffolding packages were not completed in a timely manner. 

Alternatives to erecting scaffolding were not pursued. Scaffolding was allowed to be erected 

during times in the outage when dose rates were high, such as during reactor coolant system 

cleanup.

The inspector also found that high collective radiation dose has been a continuing problem. 

Dose information obtained from the licensee is shown in the following chart.
Issue Date: 02/20/17 23 IMC 1245, Appendix C3

3 years 

ago

2 years 

ago

Last year This year

Annual Collective Dose 237 21.5 220.2 332

Outage Dose 212 NA 197 313

3-Year Average Collective 

Dose

142.6 149.2 159.6 191.2

Questions:

1. What are acceptable reasons for giving credit for exceeding dose projections? 

2. What are not legitimate reasons for giving credit for exceeding dose projections?

3. What constitutes an ALARA finding?

4. How many ALARA findings can be identified in the above scenario? 

5. When would an ALARA violation be warranted?

DOCUMENTATION: Document completion on the Health Physics Inspector Technical 

Proficiency Level Signature Card and Certification Item ISA-HP-4.
Issue Date: 02/20/17 24 IMC 1245, Appendix C3

Health Physics Inspector Individual Study Activity

TOPIC: (ISA-HP-5) In-Plant Airborne Radioactivity Control and Mitigation

PURPOSE: The purpose of this ISA is to familiarize you with various regulatory 

requirements that support Inspection Procedure 71124.03, In-Plant 

Airborne Radioactivity Control and Mitigation. The objectives of this 

procedure are to verify that the licensee is employing, to the extent 

practicable, process and engineering controls such that in-plant airborne 

concentrations are being controlled consistent with ALARA, to the extent 

necessary to validate plant operations as reported by the PI, and to verify 

that the practices and use of respiratory protection devices on site do not 

pose an undue risk to the wearer.

Title 10 of the Code of Federal Regulations (10CFR) Part 20, Subpart H, 

Respiratory Protection and Controls to Restrict Internal Exposure in 

Restricted Areas,” requires licensees to use, to the extent practical, 

process or other engineering controls (e.g., permanent and installed 

filtered ventilation, temporary HEPA units, downdraft tables, glove boxes, 

tents and other enclosures, limitation of exposure times, and 

decontamination) to control the concentration of radioactivity in the air. If 

process or other engineering controls alone are not able to maintain 

airborne concentrations of radionuclides below those defined as an 

airborne radioactivity area, then licensees must take other actions, 

consistent with the ALARA principle, to limit the intake of these 

radionuclides. The use of respiratory protection devices is one of the 

optional measures to limit intake.

COMPETENCY

AREA: TECHNICAL AREA EXPERTISE

REGULATORY FRAMEWORK

LEVEL

OF EFFORT: 19 hours

REFERENCES: 1. 10 CFR 20.1101(b)

2. 10 CFR Part 20, Subpart H, “Respiratory Protection and Controls to 

Restrict Internal Exposure in Restricted Areas”

3. 10 CFR Part 20, Appendix A, “Assigned Protection Factors for 

Respirators”

4. 29 CFR Part 1910, Subpart I, “Personal Protective Equipment”

5. Regulatory Guide 8.15, “Acceptable Programs For Respiratory 

Protection,” October 1999, and other revisions as applicable

6. NUREG-1400, “Air Sampling in the Workplace”
Issue Date: 02/20/17 25 IMC 1245, Appendix C3

7. NUREG/CR-0041, Rev. 1, “Manual of Respiratory Protection 

Against Airborne Radioactive Materials”

8. NUREG-0938, “Information for Establishing Bioassay 

Measurements and Evaluations of Tritium Exposures”

9. ANSI/Compressed Gas Association Commodity Specification for 

Air, G-7.1-1989

10. Information Notice 85-87, “Hazards of Inerting Atmospheres,”

November 18, 1985

11. Information Notice 98-20, “Problems with Emergency Preparedness 

Respiratory Protection Programs,” June 3, 1998

12. Information Notice 99-05, “Inadvertent Discharge of Carbon Dioxide 

Fire Protection System and Gas Mitigation,” March 8, 1999

13. Information Notice 00-07, “National Institute for Occupational Safety 

and Health Respirator User Notice: Special Precautions for Using 

Certain Self-Contained Breathing Apparatus Air Cylinders,” April 13, 

2000

14. NUREG/CR-5569, Rev. 1, “Health Physics Positions Data Base,”

HPPOS-94, 103, and 147 

15. NUREG 1736, “Consolidated Guidance: 10 CFR 20 - Standards for 

Protection Against Radiation, Section 3.20.1703

16. Supplementary Information, 10 CFR Part 20, “Respiratory 

Protection and Controls to Restrict Internal Exposures,” Final Rule 

Summary, Federal Register Vol. 64, No. 194, page 54543, 

October 7, 1999

17. Information Notice 94-35, “NIOSH Respirator User Notices,” 

“Inadvertent Separation of the Mask Mounted Regulator (MMR) 

from the Face piece of the Mine Safety Appliances (MSA) Company 

Self Contained Breathing Apparatus (SCBA) and Status Update,”

May 16, 1994

18. Information Notice 95-01, “DOT Safety Advisory: High Pressure 

Aluminum Seamless and Aluminum Composite Hoop Wrapped 

Cylinders,” January 4, 1995

19. NUREG/CR-6204, “Questions and Answers Based on Revised 10 

CFR Part 20,” Questions 91, 124 and 418

20. ANSI Z88.2-1992, “Practices for Respiratory Protection,” and other 

revisions as applicable
Issue Date: 02/20/17 26 IMC 1245, Appendix C3

21. ANSI Z88.10-2001, “Fit Testing Methods,” and other revisions as 

applicable

22. IE Circular 80-03, “Protection From Toxic Gas Hazards,” March 6, 

1980

23. Information Notice 81-26, Part 4, “Personnel Entry Into Inerted 

Containment”

24. Information Notice 86-46, “Improper Cleaning and Decontamination 

of Respiratory Protection Equipment”

25. Information Notice 83-68, “Respirator User Warning: Defective SelfContained Breathing Apparatus Air Cylinders,” October 11, 1983

26. Information Notice 85-48, “Respirator User Warning: Defective SelfContained Breathing Apparatus Air Cylinders,” June 19, 1985

27. Information Notice 86-103, “Respirator Coupling Nut Assembly 

Failures,” December 16, 1986

28. Information Notice 89-47, “Potential Problems with Worn or 

Distorted Hose Clamps on Self-Contained Breathing Apparatus,” 

May18, 1989

29. Information Notice 97-66, “Failure to Provide Special Lenses for 

Operators Using Respirator or SCBA During Emergency 

Operations,” August 20, 1997

EVALUATION

CRITERIA: Upon completion of this ISA, the inspector should be able to:

1. Discuss the application of the relevant portions of 10 CFR 20 which 

pertain to respiratory protective equipment.

2. Discuss the regulatory philosophy for providing engineering controls 

to minimize airborne radioactivity (i.e., describe why Part 20 gives 

priority to engineering controls over the use respiratory protection 

devices.)

3. Discuss the importance of providing adequate monitoring for 

changing airborne radioactivity concentrations in the workplace.

4. Describe the essential elements of a basic respiratory protection 

program that would be acceptable to the NRC.

5. Describe the characteristics of Grade D air quality, and where, 

when, and how air quality is tested. 
Issue Date: 02/20/17 27 IMC 1245, Appendix C3

6. Explain the potential consequences of not using SCBA in IDLH 

environments, and describe the testing, maintenance, and 

replacement cycles for the various SCBA cylinders.

7. Describe deficiencies that are commonly found in respiratory 

protection training programs for the use of SCBA and how they 

might impact emergency preparedness.

8. Describe the relationship of NIOSH certification of SCBA designs to 

NRC regulatory compliance and explain how part substitution can 

invalidate the NIOSH certification.

TASKS: 1. Read 10 CFR 20, Subpart H portions applicable to SCBA.

2. Read Regulatory Guide 8.15, “Acceptable Programs for Respiratory 

Protection,” October 1999.

3. Read Information Notice 85-87, “Hazards of Inerting Atmospheres,”

November 18, 1985.

4. Read Information Notice 98-20, “Problems with Emergency 

Preparedness Respiratory Protection Programs,” June 3, 1998 (This 

one is particularly important from an inspection standpoint.)

5. Read HPPOS-147, “Respirator User’s Notice – Use of Unapproved 

Subassemblies” (This is a recurrent problem.)

6. Read the scenario provided and answer associated questions.

7. Meet with your supervisor or a qualified Health Physics inspector 

and discuss any questions you may have as a result of this ISA. 

Discuss the answers to the questions and your work products listed 

under the Evaluation Criteria of this study guide with your 

supervisor.

Scenario A

During a review of self-contained breathing apparatus maintenance and surveillance records, an 

inspector identified that 36 self-contained breathing apparatus air bottles were beyond the 3-

year hydrostatic test dates. Hydrostatic testing had expired in April 2001 for 31 of the selfcontained breathing apparatus air bottles that were in service. According to the NIOSH, selfcontained breathing apparatus units with expired hydrostatic testing are no longer certified.
Issue Date: 02/20/17 28 IMC 1245, Appendix C3

Questions: 

1. Describe the relationship of NIOSH certification to SCBA designs to NRC regulatory 

compliance.

2. Could this be a violation? Why or why not.

3. Other than Radiation Protection, what NRC cornerstone would also need to be 

evaluated?

Scenario B

A plant is in a refueling outage in which the steam generators are scheduled to be replaced. As 

part of the steam generator project, the plant will have to cut out a large opening in the 

containment building to support the removal of the old generators and the installation of the new 

generators. One of the activities involves vacuuming out the “A” steam generator cold leg to 

remove debris prior to the pipe end decontamination work. 

While performing the debris removal activity using a vacuum cleaner unequipped with a HEPA 

filter and designated “wet use”, the exhaust from the vacuum cleaner caused airborne 

radioactivity levels to increase throughout containment. When various portable air monitoring 

systems alarmed alerting personnel to the presence of airborne radioactivity, Radiation 

Protection personnel initiated the evacuation of about 175 workers from containment. All of the 

individuals were monitored for external and internal contamination using personnel 

contamination monitors and whole body counting equipment. Of the 175 workers, 145 workers 

were determined to have sustained either low-level external radioactive contamination or lowlevel intakes of airborne radioactivity. The highest calculated internal committed effective dose 

equivalent to one individual was slightly over 10 mrem.

Low level contamination was detected outside the Personnel Hatch. Airborne environmental 

monitoring samples collected from two ODCM continuous monitoring stations identified 

statistically detectable radioactivity above background. Estimates of the projected maximum 

annual dose to the critical receptor ranged from 0.02 to 0.05 mrem (organ) in a year.

Questions

1. Assess the event for performance deficiencies, possible violations and significance of 

the event.

2. With respect to the occupational radiological safety aspects associated with the event, 

discuss what should be reviewed as part of your assessment.

3. What specific radiological hazard that can have a significant impact in the evaluation of 

occupational internal exposure does the licensee need to consider in their evaluation?

4. Has a violation of 10 CFR 20 occurred? If so, what regulation is applicable to the event?

5. What do you believe the safety significance of this event to be and why? Identify any 

cross cutting aspects that may be applicable.
Issue Date: 02/20/17 29 IMC 1245, Appendix C3

6. With respect to the public radiological safety aspects associated with the event, discuss 

what should be reviewed as part of your assessment.

7. Are doses to the public significant? Explain answer.

8. What other possible violations could have occurred from this event? (Hint: Consider 

controls for the large additional opening of containment – did they work)

DOCUMENTATION: Document completion on the Heath Physics Inspector Technical 

Proficiency Level Signature Card and Certification Item ISA-HP-5.
Issue Date: 02/20/17 30 IMC 1245, Appendix C3

Health Physics Inspector Individual Study Activity 

TOPIC: (ISA-HP-6) Occupational Dose Assessment

PURPOSE: The purpose of this ISA is to familiarize you with various regulatory 

requirements that support Inspection Procedure 71124.04, “Occupational 

Dose Assessment.” The objectives of this inspection procedure are to 

determine the accuracy and operability of personal monitoring equipment, 

determine the accuracy and effectiveness of the licensee’s methods for 

determining total effective dose equivalent, and ensure that occupational 

dose is appropriately monitored.

COMPETENCY

AREA: TECHNICAL AREA EXPERTISE REGULATORY FRAMEWORK

LEVEL

OF EFFORT: 19 hours

REFERENCES: 1. 10 CFR 20.1003 (Definitions associated with study activity area)

2. 10 CFR 20.1201 through 20.1208

3. 10 CFR 20.1501(c)

4. IP 71124.04, Section 02.04(a)

5. Regulatory Issue Summary 2003-04, “Use of the Effective Dose 

Equivalent in Place of the Deep Dose Equivalent in Dose 

Assessments,” February 13, 2003

6. Regulatory Issue Summary 2004-01, “Method for Estimating 

Effective Dose Equivalent from External Radiation Sources Using 

Two Dosimeters,” February 17, 2004

7. Regulatory Issue Summary 2009-09, “Use of Multiple Dosimetry and 

Compartment Factors in Determining Effective Dose Equivalent 

from External Radiation Exposures,” July 13, 2009

8. Regulatory Guide 8.40, “Methods for Estimating Effective Dose 

Equivalent from External Exposure,” July 2010, and other revisions 

as applicable

9. ANSI N13.52-1999, “Personnel Neutron Dosimeters (Neutron 

Energies Less Than 20 MeV),” and other revisions as applicable

10. ANSI N13.6-1999, “Practice for Occupational Radiation Exposure 

Records System,” and other revisions as applicable

11. ANSI N13.11-2001, “Personnel Dosimetry Performance – Criteria 

for Testing,” and other revisions as applicable 
Issue Date: 02/20/17 31 IMC 1245, Appendix C3

12. ANSI N13.30-1996, “Performance Criteria for Bioassay,” and other 

revisions as applicable 

13. NUREG/CR-6581, “Considerations in the Application of the 

Electronic Dosimeter to Dose of Record”

14. Regulatory Guide 8.13, “Instructions Concerning Prenatal Radiation 

Exposure,” June 1999, and other revisions as applicable

15. Regulatory Guide 8.20, “Applications of Bioassay for Radioiodine,”

September 2014, and other revisions as applicable

16. Regulatory Guide 8.34, “Monitoring Criteria and Methods to 

Calculate Occupational Radiation Doses,” July 1992, and other 

revisions as applicable

17. Regulatory Guide 8.7, “Instructions for Recording and Reporting 

Occupational Radiation Exposure Data,” November 2005, and other 

revisions as applicable

18. NUREG/CR-4884, “Interpretation of Bioassay Measurements”

19. Regulatory Guide 8.36, “Radiation Dose to the Embryo/Fetus,” July 

1992, and other revisions as applicable

20. NCRP Report 128, “Radionuclide Exposure of the Embryo/Fetus,” 

1998 

21. NUREG/CR-5631, “Contribution of Maternal Radionuclide Burdens 

to Prenatal Radiation Doses--Interim Recommendations”

EVALUATION

CRITERIA: Upon completion of this ISA, the inspector should be able to:

1. Describe the responsibility of a licensee for limiting, monitoring, and 

tracking the exposure of a declared pregnant worker. Specifically, 

discuss the following:

a. the timing of licensee actions,

b. the embryo/fetus exposure limits, and

c. the inspection considerations detailed in IP 71124.04, Section 

02.05(a)

2. Describe the regulatory requirements for the accreditation of 

personnel dosimeters used for legal dose records.

3. Describe the NRC-approved methods for assessing effective dose 

equivalent.

4. Describe the NRC-approved methods for assessing internal doses

described in RG 8.34.
Issue Date: 02/20/17 32 IMC 1245, Appendix C3

5. Discuss TEDE, CEDE, LDE, SDE and how they are measured and 

where the values are recorded in NRC Forms 4 and 5.

TASKS: 1. Review RG 8.13 related to a declared pregnant worker (DPW) and 

discuss the technical basis for the DPW dose limit.

2. Review each underlined reference and discuss its application with a 

qualified inspector.

3. Read the scenarios and answer the associated questions.

Scenario A: Adequate Surveillance of Workers in High Radiation Area

A pipe fitter (Worker A) received an unintended radiation exposure while working in the reactor 

water cleanup (RWCU) heat exchanger room. The plant was in refueling outage RFO9 and the 

RWCU system was out of service for chemical decontamination. Radiation protection (RP) 

personnel were supporting work in the RWCU heat exchanger room which involved rotating the 

remaining three (of 16) spectacle flanges, which were located close to the heat exchangers, as 

part of the chemical decontamination process. Substantial radiation dose gradients existed in 

the room as a result of hot spots on the heat exchangers and associated piping. RP support 

personnel for this portion of the work consisted of an ALARA specialist, a lead RP technician 

(RPT) and a RP specialist (RPS). An Infrequently Performed Test or Evolution briefing and an 

ALARA briefing for the flange work were conducted. The workers were issued telemetry 

dosimetry (wireless remote monitoring system) to track their dose as the licensee had 

determined that telemetry dosimetry would be the only method for tracking worker exposure for 

this job. The electronic dosimeter set-points were 1000 mrem total dose and 15000 mrem/hour 

dose rate. Approximately 10 workers were on the telemetry dosimetry monitoring system. A 

television camera was located inside the RWCU room and provided visual monitoring of the 

workers. However, this camera was controlled by licensee personnel on the refueling floor and 

not by the RP staff controlling this job. The ALARA specialist and the RPT were outside of the 

RWCU room and the lead RPT was stationed just inside the room. One worker (A) noted to the 

RP staff that his telemetry screen entry was a different color than all of the other workers. 

Worker A then noticed that his name was no longer on the telemetry computer screen, he 

informed the RPS. He was told that his name would come back later. A short time later Worker 

A noted to the RP staff that his name was still not on the telemetry read out. The worker was 

told that his name would probably show up again when he entered the RWCU room and he was 

told to talk to the lead RPT about the issue. The RPS did not expect Worker A to enter the 

RWCU room for work because he believed that the worker would notify the lead RPT inside the 

RWCU room that he was having a dosimetry problem and would be replaced. Worker A did not 

discuss his telemetry problem with the lead RPT and entered the RWCU room with the other 

two workers. However, the RP staff outside of the RWCU room did not realize that all three 

workers entered the area. Once inside, Workers A and B were in close proximity to each other 

but Worker A was not at his assigned flange and was unable to locate his correct work site. 

After some time he returned to the entrance of the room. The third worker (C) climbed a 

scaffold to upper elevations of the room to work on his flange. After Worker A returned to the 

alcove just inside the RWCU room to ask about the location of his work area, Worker B was 

reaching his limit of 800 mrem for the entry and all workers were told to return to the low dose 

waiting area just inside of the room. Worker B told Worker A where the flange was that Worker 

A was to work on, and Worker B left the area. Worker C asked for his dose and was told 57 
Issue Date: 02/20/17 33 IMC 1245, Appendix C3

mrem by the RPT who was outside of the room acting as the LHRA gate guard. Worker A also 

asked for his dose and was told 57 mrem, which he questioned. He was told that his reading 

was low and that he could return to work. Both Workers A and C returned to work. Worker C 

then finished his task and exited the area. The RPS, who periodically had been observing the 

workers on a TV monitor that had a view of the lower area, had assumed that the worker who 

was showing increased dose on the telemetry (Worker C) was the same person he had seen on 

the monitor. When the RPS noticed that a worker was still in the area with no one now in a 

dose field on the telemetry screen, he told the lead RPT to remove the worker from the RWCU 

room. When Worker A exited the room his electronic dosimeter registered 1834 mrem. Worker 

A was in a significant radiological gradient at his assigned work site due to hot spots on the HX 

and on piping close to his work area. The licensee’s dose reconstruction determined that the 

portion of the body that was in the highest radiation field was the worker’s head. However, the 

licensee did not place the worker’s dosimetry on his head, but placed the dosimetry on the 

worker’s chest, which was the licensee’s standard practice.

Questions:

1. Did the licensee provide adequate surveillance of workers in the high radiation area?

2. Was the location of the dosimetry appropriate? Why or why not?

3. Could this be a violation of TSs?

Scenario B

Refer to Scenario A from (ISA-HP-3), Radiological Hazard Assessment and Exposure Controls

Assume: 

One of the filters is dropped, and promptly picked up by a worker wearing a properly functioning 

PAPR, his breathing zone air sampler results shows 8 DAC beta-gamma and 110 DAC alpha. 

Questions:

1. Calculate the estimated uptake by the individual with a respirator and without a 

respirator.

2. What engineering controls were considered?

3. Was there a significant potential for an over exposure?

DOCUMENTATION: Document completion on the Health Physics Inspector Technical 

Proficiency Level Signature Card and Certification Item ISA-HP-6.

 
Issue Date: 02/20/17 34 IMC 1245, Appendix C3

Health Physics Inspector Individual Study Activity

TOPIC: (ISA-HP-7) Radiation Monitoring Instrumentation

PURPOSE: The purpose of this ISA is to familiarize you with various regulatory 

requirements that support Inspection Procedure 71124.05, “Radiation 

Monitoring Instrumentation.” The objective of this inspection procedure is 

to verify that the licensee is ensuring the accuracy and operability of 

radiation monitoring instruments that are used to monitor areas, materials, 

and workers to ensure a radiologically safe work environment. The 

instrumentation subject to this inspection procedure includes equipment 

used to monitor radiological conditions related to normal plant operations, 

including anticipated operational occurrences, and conditions resulting 

from postulated accidents.

COMPETENCY 

AREA: TECHNICAL AREA EXPERTISE

REGULATORY FRAMEWORK

LEVEL

OF EFFORT: 19 hours

REFERENCES: 1. 10 CFR Part 50, Appendix I

2. 10 CFR Part 50, Appendix A, Criterion 64

3. 10 CFR Part 20, Subpart F

4. 10 CFR Part 61

5. NUREG/CR-5569, Rev. 1, “Health Physics Positions Data Base,” 

HPPOS-001, 088, 279, and 328

6. NUREG-1736, “Consolidated Guidance: 10 CFR 20 - Standards for 

Protection Against Radiation, Section 3.20.1501”

7. Regulatory Guide 4.15, “Quality Assurance for Radiological 

Monitoring Programs,” July 2007, and other revisions as applicable

8. Regulatory Guide 1.21, “Measuring, Evaluating and Reporting 

Radioactivity in Solid Wastes and Releases of Radioactive Materials 

in Liquid and Gas from LWRs,” June 2009, and other revisions as 

applicable

9. Regulatory Guide 1.33, “Quality Assurance Program Requirements 

(Operations),” June 2013, and other revisions as applicable

10. IE Circular 81-07, “Control of Radioactive Contaminated Material,”

May 14, 1981
Issue Date: 02/20/17 35 IMC 1245, Appendix C3

11. Information Notice 85-92, “Surveys of Wastes Before Disposal from 

Nuclear Reactor Facilities,” December 2, 1985

12. Information Notice 93-30, “NRC Requirements for Evaluation of 

Wipe Test Results; Calibration of Count Rate Survey Instruments,” 

April 12, 1993

13. IE-Bulletin 97-001, “Potential for Erroneous Calibration, Dose Rate, 

or Radiation Exposure Measurements with Certain Victoreen Model 

530 and 53OSI Electrometer/Dose-Meters”

14. NUREG-0737, “Clarification of TMI Action Plan Requirements”

15. ANSI 42.14-1991, “Calibration and Use of Germanium 

Spectrometers for the Measurement of Gamma-Ray Emission 

Rates of Radionuclides,” and other revisions as applicable

16. ANSI N323D-2002, “Installed Radiation Protection Instrumentation,”

and other revisions as applicable 

17. Regulatory Guide 8.6, “Standard Test Procedure for Geiger-Muller 

Counters,” May 1973, and other revisions as applicable

18. ANSI N42.17A-1989, “Performance Specifications for Health 

Physics Instrumentation-Portable Instrumentation for Use in Normal 

Environmental Conditions,” and other revisions as applicable

19. ANSI N42.17B-1989, “Performance Specifications for Health 

Physics Instrumentation-Occupational Airborne Radioactivity 

Monitoring Instrumentation,” and other revisions as applicable

20. ANSI N42.17C-1989, “Performance Specifications for Health 

Physics Instrumentation-Portable Instrumentation for Use in 

Extreme Environmental Conditions,” and other revisions as 

applicable

21. ANSI N323A-1997, “Radiation Protection Instrumentation and 

Calibration, Portable Survey Instruments,” and other revisions as 

applicable

22. ANSI N13.4-1971, “American National Standard for the 

Specification of Portable X- or Gamma-Radiation Survey 

Instruments,” and other revisions as applicable

23. NUREG/CR-6204, “Questions and Answers Based on Revised 10 

CFR Part 20,” Questions 147 and 209

24. Regulatory Guide 1.97, “Criteria for Accident Monitoring 

Instrumentation for Nuclear Power Plants,” June 2006, and other 

revisions as applicable
Issue Date: 02/20/17 36 IMC 1245, Appendix C3

25. NUREG-1400, “Air Sampling in the Workplace”

26. IEEE-N323C-2009, “American National Standard for Radiation 

Protection Instrumentation Test and Calibration – Air Monitoring 

Instruments”

27. IEEE-N42.12-1994-R2004, “ANSI Calibration and Usage of 

Thallium, Activated Sodium Iodide Detector Systems for Assay of 

Radionuclides”

28. IEEE-42,25-1997-R2005, “ANSI Calibration and Usage of AlphaBea Proportional Counters

29. IEEE-N42.22-1995, “ANSI Traceability of Radioactive Sources to 

NIST and Associated Instrument Quality Control”

EVALUATION

CRITERIA: Upon completion of this ISA, the inspector should be able to:

1. Describe the effects of operating temperature and pressure on the 

accuracy of instruments.

2. Describe and discuss the calibration techniques used for portable 

radiation detection instruments including where electronic source 

usage is appropriate and typical calibration frequencies.

3. Describe and discuss the calibration techniques for environmental 

monitoring equipment e.g., air sample, environmental TLD, 

composite samplers, and typical calibration frequencies.

4. Describe and discuss the measurement and calibration techniques 

used for RP and Chemistry laboratory instruments such as 

proportional counter, gamma spectroscopy, and liquid scintillation 

counter, and typical calibration frequencies. Identify the types of 

sources to be used and expected MDA or LLDs.

5. Discuss laboratory analytical techniques for radionuclides and 

radiations such as gamma, Sr-89/90, H-3, and gross alpha/beta. 

Identify best methods.

6. Discuss Minimum Detectable Activity (MDA) and Lower Limits of 

Detection (LLDs).

7. Describe the appropriate uses for alarming dosimeters and their 

limitations.

8. Explain the NRC requirements associated with calibration interval.

9. Describe the guidance in place to assess the implementation of 10 

CFR 20.1501(c).
Issue Date: 02/20/17 37 IMC 1245, Appendix C3

10. Discuss the calibration technique for contamination monitors used to 

free release material (e.g., small article monitor (SAM), bag monitor, 

frisker) in order to conform to guidance in IE Circular 81-07 and 

Information Notice 85-92.

TASKS: 1. Read HPPOS-088 which pertains to calibration of air sampling 

equipment. (The following equation is provided to help with 

understanding the issue. See NUREG 1400 for further details.)

( )( )

c

s

s

c

s c

T

T

P

P

V V

Gas Law Correction

Where:

a. Vs = volume under field conditions (appropriate volume unit)

b. Vc = volume under calibration conditions (appropriate volume 

unit)

c. Ps = absolute pressure during sampling (mm Hg)

d. Pc = absolute pressure during calibration (mm Hg)

e. Ts = absolute temperature during sampling (Kelvin)

f. Tc = absolute temperature during calibration (Kelvin)

Certain survey meters are subject to the same type and magnitude 

of errors. Typically in nuclear power situations the instrument of 

choice is a vented ion chamber. It will typically be calibrated at 

atmospheric pressure and can be used in a sub-atmospheric 

containment which could result in a reduced response due to lower 

contained air mass in the chamber. Conversely in a large dry 

containment the instrument could be over responding to increased 

pressure. The tolerance allowed on survey meters is typically large 

enough that most common situations do not result in correction of 

survey meter readings. 

2. Read HPPOS-279. With the exception of certain high range 

instruments, use of an electronic calibration is generally not 

considered sufficient. 

3. Read HPPOS-328 (Alarming Dosimeter)

4. Read Q&A 209 and 147, (Calibration frequency)

5. Read the portion of NUREG-1736 that applies to 10 CFR 20.1501(c)
Issue Date: 02/20/17 38 IMC 1245, Appendix C3

Note: The NUREG predates the change to 10 CFR 20.1501 that 

added a new paragraph changing the existing 20.1501(b) to 

20.1501(c). The intended section is titled “Instrument Calibration.”

6. Review and evaluate the characteristics of radiation monitoring 

systems (RMS), such as gamma-scintillation detector, betascintillation detector, GM, and ion chamber used by the licensee. 

Identify capabilities and limitations

7. Review and evaluate the different measuring techniques, such as 

proportional counter, gamma spectroscopy, and liquid scintillation 

counter.

8. Review and evaluate the calibration techniques and results for the 

measurement of laboratory’s gamma spectroscopy, proportional 

counter and liquid scintillation counter.

9. Review, evaluate, and understand the laboratory’s analytical 

techniques (e.g., gamma, Sr-89/90, H-3, and gross alpha/beta.)

10. Review and evaluate the MDA and LLDs used by the licensee.

11. Discuss with a Senior Health Physics Inspector the types of 

instruments commonly used, their application and calibration 

techniques. Include both portable survey instruments (e.g., gamma 

and neutron, and installed equipment (e.g., area radiation monitors 

and portal monitors).

12. Meet with your supervisor or a qualified inspector and discuss any 

questions you may have as a result of this training activity. Discuss 

the answers to the questions and your work products listed under 

the Evaluation Criteria section of this study guide with your 

supervisor.

Scenario: Calibration Instrumentation

During a tour of the radiologically controlled area on July 18, 2001, an inspector identified a 

continuous air monitor in the radioactive waste truck bay with an expired calibration. The 

calibration due date was May 31, 2001. The licensee had identified on June 4, 2001, a survey 

instrument was out of calibration. The calibration due date was also May 31, 2001. The 

licensee had not properly marked the instruments out of calibration or removed them to the 

designated holding area. Radiation Protection Procedure requires that instruments be properly 

marked out of calibration and/or placed in a proper holding area. 

Questions:

1. Is there a technical specification that requires written procedures? 

2. What regulatory guide recommends procedures for area, portable, and airborne 

radiation monitor calibrations? 
Issue Date: 02/20/17 39 IMC 1245, Appendix C3

3. Would this be a violation? Why?

DOCUMENTATION: Document completion on the Health Physics Inspector Technical 

Proficiency Level Signature Card and Certification Item ISA-HP-7.
Issue Date: 02/20/17 40 IMC 1245, Appendix C3

Health Physics Inspector Individual Study Activity

TOPIC: (ISA-HP-8) Radioactive Gaseous and Liquid Effluent Treatment 

PURPOSE: The purpose of this ISA is to familiarize you with the regulatory bases and 

historical agency positions in the area of radioactive gaseous and liquid 

effluent treatment and provide you the technical knowledge to conduct 

inspections using procedure 71124.06, “Radioactive Gaseous and Liquid 

Effluent Treatment.” The objectives of procedure 71124.06, “Radioactive 

Gaseous and Liquid Effluent Treatment,” are:

1. To ensure that the gaseous and liquid effluent processing systems 

are maintained so that radiological discharges are properly 

mitigated, monitored, and evaluated with regard to public exposure. 

The principal design Criterion for these systems are found in 

General Design Criteria 60 and 64 of Appendix A to 10 CFR Part 

50, Radiological Effluent Technical Specifications (RETS). Generic 

Letter 89-01, NUREG-1301, NUREG-1302, Standard Radiological 

Effluent Controls (SREC), and the Offsite Dose Calculation Manual 

(ODCM) provide acceptable methods of implementation to meet 

these design criteria.

2. To ensure that abnormal radioactive gaseous or liquid discharges 

and conditions, when effluent radiation monitors are out-of-service, 

are controlled in accordance with applicable regulatory requirements 

and licensee procedures.

3. To verify that the licensees’ quality control program ensures that the 

radioactive effluent sampling and analysis requirements are 

satisfied so that discharges of radioactive materials are adequately 

quantified and evaluated.

4. To verify the adequacy of public dose calculations and projections 

resulting from radioactive effluent discharges.

 

The NRC requires that licensees’ ensure adequate protection of public 

health and safety and the environment from exposure to radioactive 

materials released to the public domain. Radiation exposure to the public 

must be below the 10 CFR Part 20 and 40 CFR Part 190 limits. Doses 

below the design objectives of Appendix I to 10 CFR Part 50 and 40 CFR 

Part 190 dose values are considered ALARA. Radioactive effluent 

treatment systems are required by Criteria 60 and 64 of Appendix A to 

10 CFR Part 50.

COMPETENCY

AREA: TECHNICAL AREA EXPERTISE

LEVEL 

OF EFFORT: 80 HOURS

REFERENCES: 1. 40 CFR Part 190
Issue Date: 02/20/17 41 IMC 1245, Appendix C3

2. 10 CFR 50.75(g)

3. 10 CFR 50.59

4. 10 CFR Part 50, Appendix A, Criterion 60

5. 10 CFR Part 50, Appendix A, Criterion 64

6. 10 CFR 50.34(a)

7. 10 CFR 72.44(d) 

8. 10 CFR 72.104

9. Regulatory Guide 1.33, “Quality Assurance Program Requirements 

(Operation),” February 1978, and other revisions as applicable

10. Regulatory Guide 1.112, “Calculation of Annual Releases of 

Radioactive Material in Gaseous and Liquid Effluents from Light 

Water Cooled Power Reactors,” March 2007, and other revisions as 

applicable

11. Regulatory Guide 4.1, “Radiological Environmental Monitoring for 

Power Plants,” June 2009, and other revisions as applicable

12. NUREG-1301, “Offsite Dose Calculation Manual Guidance: 

Standard Radiological Effluent Controls for Pressurized Water 

Reactors”

13. NUREG-1302, “Offsite Dose Calculation Manual Guidance: 

Standard Radiological Effluent Controls for Boiling Water Reactors”

14. ANSI N42.18-2004, “Specification and Performance of On-Site 

Instrumentation for Continuously Monitoring Radioactivity in 

Effluents,” and other revisions as applicable

15. Plant Updated Final Safety Analysis Report

16. NUREG/CR-5569, Rev. 1, “Health Physics Positions Data Base”

17. NUREG/CR-6204, “Questions and Answers Based on Revised 10 

CFR Part 20”

18. NUREG-1736, “Consolidated Guidance: 10 CFR Part 20- Standards 

for Protection Against Radiation”

19. NUREG/CR-4757, “Line-Loss Determination for Air Sampler 

Systems”
Issue Date: 02/20/17 42 IMC 1245, Appendix C3

20. Environmental Measurement Laboratory, HASL-300 Procedures 

Manual U.S. Department of Energy, New York, NY

21. NCRP Report 58, “A Handbook of Radioactivity Measurements 

Procedures,” 2nd Edition, 1984

22. Regulatory Guide 1.21, “Measuring, Evaluating, and Reporting

Radioactive Material in Liquid and Gaseous Effluents and Solid 

Waste,” June 2009, and other revisions as applicable

23. Regulatory Guide 1.52, “Design, Inspection, and Maintenance 

Criteria for Air Filtration and Adsorption Units of Post-Accident 

Engineered-Safety-Feature Atmosphere Cleanup Systems of LightWater-Cooled Nuclear Power Plants,” June 2001, and other 

revisions as applicable

24. Regulatory Guide 1.109, “Calculation of Annual Doses to Man from 

Routine Releases of Reactor Effluents for the Purpose of Evaluating 

Compliance with 10 CFR 50, Appendix I,” October 1977, and other 

revisions as applicable

25. Regulatory Guide 1.111, “Methods for Estimating Atmospheric 

Transport and Dispersion of Gaseous Effluents in Routine Releases 

from Light-Water-Cooled Reactors,” July 1977, and other revisions 

as applicable

26. Regulatory Guide 1.140, “Design, Inspection, and Testing Criteria 

for Air Filtration and Adsorption Units of Normal Atmosphere 

Cleanup Systems in Light-Water-Cooled Nuclear Power Plants,” 

August 2016, and other revisions as applicable

27. Regulatory Guide 4.15, “Quality Assurance for Radiological 

Monitoring Programs (Inception through Normal Operations to 

License Termination) - Effluent Streams and the Environment,” July 

2007, and other revisions as applicable

28. ANSI N13.1-1999, “Sampling and Monitoring Releases of Airborne 

Radioactive Substances from the stacks and Ducts of Nuclear 

Facilities,” and other revisions as applicable

29. ANSI N45.2.23-1978, “Qualification of Quality Assurance Program 

Audit Personnel for Nuclear Power Plant Personnel,” and other 

revisions as applicable

30. ANSI/ANS N55.4-1979, “Gaseous Radioactive Waste Processing 

Systems for Light-Water-Cooled Reactor Plants,” and other 

revisions as applicable

31. ANSI/ANS N55.6-1979, “Liquid Radioactive Waste Processing 

System for Light-Water-Cooled Reactor Plants,” and other revisions 

as applicable
Issue Date: 02/20/17 43 IMC 1245, Appendix C3

32. ANSI/ASME N509-1980, “Nuclear Power Plant Air Cleaning Units 

and Components,” and other revisions as applicable

33. ANSI/ASME N510-1980, “Testing of Nuclear Air-Cleaning Systems,”

and other revisions as applicable 

34. IE Bulletin 80-10, “Contamination of Nonradioactive System and 

Resulting Potential for Unmonitored, Uncontrolled Release of 

Radioactivity to Environment,” May 6, 1980

35. IE Circular No. 80-18, “10 CFR 50.59 Safety Evaluations for 

Changes to Radioactive Waste Treatment Systems,” August 22, 

1980

36. Information Notice No. 82-43, “Deficiencies in LWR Air 

Filtration/Ventilation Systems,” November 16, 1982

37. Information Notice No. 82-49, “Correction for Sample Conditions for 

Air and Gas Monitoring,” December 16, 1982

38. Information Notice No. 13-01, “Emergency Action Level Thresholds 

Outside the Range of Radiation Monitors,” February 13, 2013

39. Information Notice No. 13-12, Improperly Sloped Instrument 

Sensing Lines,” July 13, 2013

40. Information Notice No. 13-13, “Deficiencies with Effluent Radiation 

Monitoring System Instrumentation,” July 12, 2013

41. Generic Letter 89-01, “Implementation of Programmatic Controls for 

Radiological Effluent TS in the Administrative Controls Section of 

the TS and the Relocation of Procedural Details of RETS to the 

ODCM or to the PCP,” January 31, 1989

42. Generic Letter 83-13, “Clarification of Surveillance Requirements for 

HEPA Filters and Charcoal Absorber Units in Technical 

Specifications on ESF Cleanup Systems,” March 2, 1983

EVALUATION 

CRITERIA: Upon completion of this ISA, the inspector should be able to:

1. Discuss and identify effluent sampling techniques that may be used 

by the licensee for sampling effluents such as radioactive waste 

water, air iodine, tritium and air particulate.

2. Describe and discuss the calibration techniques used for effluent 

monitoring instruments, alarm set points and typical calibration 

frequencies.
Issue Date: 02/20/17 44 IMC 1245, Appendix C3

3. Discuss laboratory QA/QC Policy and its implementation, including 

inter-laboratory and intra-laboratory comparisons. Discus NRC 

expectations in this area.

4. Discuss calculation of projected public dose calculation 

methodologies (all pathways) listed in Regulatory Guide 1.109 and 

the use of the results from the annual land use census in the dose 

assessment process.

5. Discuss principle of air cleaning systems, such as air capacity test, 

in-place testing, and laboratory test to determine the iodine 

collection efficiency. Discuss applicable guidance documents and 

acceptable values.

6. Discuss the differences between Engineered-Safety-Feature (ESF)

atmosphere cleanup systems, evaluated using the guidance of RG 

1.52 versus the normal atmosphere cleanup systems using the 

guidance of RG 1.140.

7. Review the air cleaning system section from your assigned facility’s 

UFSAR and Technical Specification. Discuss the principal 

components and their use. 

8. Identify the various sources of liquid and gaseous radioactive 

wastes at your assigned facility by review of the UFSAR. Discuss 

the UFSAR described waste streams, and technologies associated 

with liquid and gaseous radioactive waste processing for the facility.

9. Discuss considerations that would factor into collecting a 

“representative sample.”

10. Discuss the licensee’s 10 CFR Part 61 report for determining the 

amount of radionuclides released.

TASKS: 1. Under the direction of your mentor, select a facility for review.

2. Review and evaluate the effluent sampling techniques, such as 

radioactive waste water, air iodine, tritium, and air particulate.

3. Review and evaluate the laboratory’s QA/QC Policy and its 

implementation, including inter-laboratory and intra-laboratory 

comparison.

4. Review and evaluate the projected public dose calculation 

methodologies (all pathways as available) listed in Regulatory Guide 

1.109.

5. Review and evaluate principle air cleaning system tests, such as air 

capacity test, in-place testing, and laboratory test to determine the 

iodine collection efficiency. Compare with applicable regulatory 

limits.
Issue Date: 02/20/17 45 IMC 1245, Appendix C3

6. Identify the various sources of liquid and gaseous radioactive waste, 

waste streams, and technologies associated with liquid and 

gaseous radioactive waste processing. Compare that with the 

descriptions presented in the UFSAR.

7. Identify the gaseous and liquid effluent monitors for your facility 

using the FSAR and ODCM. Locate in the ODCM the required 

functional tests and channel calibrations specified for these effluent 

radiation monitors.

8. Find the equations for the alarm set points for the effluent monitors 

in the ODCM.

9. Explain how accident radiation monitors are calibrated and how the 

calibration would differ from routine radiation monitors.

DOCUMENTATION: Document completion on the Health Physics Inspector Technical 

Proficiency Level Signature Card and Certification Item ISA-HP-8.
Issue Date: 02/20/17 46 IMC 1245, Appendix C3

Health Physics Inspector Individual Study Activity

TOPIC: (ISA-HP-9) Radiological Environmental Monitoring Program

PURPOSE: The purpose of this ISA is to familiarize you with the regulatory bases and 

historical agency positions in the areas of radiological environmental

monitoring and provide you the technical knowledge to conduct 

inspections using procedure 71124.07, “Radiological Environmental 

Monitoring Program (REMP).” The objectives of procedure 71124.07, 

“Radiological Environmental Monitoring Program,” are:

1. To verify that the REMP quantifies the impact of radioactive effluent 

releases to the environment and sufficiently validates the integrity of 

the radioactive gaseous and liquid effluent release program.

2. To verify that the REMP is implemented consistently with the 

licensee’s TS and/or ODCM and to validate that the radioactive 

effluent release program meets the design objective in Appendix I to 

10 CFR Part 50.

3. To ensure that the REMP (1) monitors non-effluent exposure 

pathways (e.g., onsite spills or leaks, exposures from direct and 

scattered (skyshine) radiation from plant facilities and components), 

(2) is based on sound principles and assumptions, and (3) validates 

that doses to members of the public are within the dose limits of 

10 CFR 20, “Standards for Protection against Radiation,” and 40 

CFR Part 190, “Environmental Radiation Protection Standards for 

Nuclear Power Operations,” as applicable.

4. To verify that the licensee is continuing to implement the voluntary 

NEI/Industry Groundwater Protection Initiative (GPI).

The NRC requires that licensee’s ensure adequate protection of public 

health and safety from exposure to radioactive material released to the 

public domain as a result of routine operations. The REMP is required by 

Criterion 64 of Appendix A to 10 CFR Part 50. The REMP supplements 

the effluent monitoring program by verifying that the measurable 

concentrations of radioactive materials and levels of radiation in the 

environment are in agreement with the values calculated by the 

radioactive effluent monitoring program. The licensee is required to 

implement the REMP in accordance with its Technical Specifications 

and/or Offsite Dose Calculation Manual, which are based on the design 

objectives contained in Appendix I of 10 CFR Part 50, as required by 

10 CFR 50.34a.

COMPETENCY

AREA: TECHNICAL AREA EXPERTISE

LEVEL 

OF EFFORT: 30 HOURS
Issue Date: 02/20/17 47 IMC 1245, Appendix C3

REFERENCES: 1. 10 CFR 50.34 (a)

2. 10 CFR Part 50, Appendix A, Criterion 64

3. 40 CFR 141

4. Plant Updated Final Safety Analysis Report and Offsite Dose 

Calculation Manual for the facility designated by your supervisor

5. Regulatory Guide 1.23, “On-site Meteorological Programs for 

Nuclear Power Plants,” March 2007, and other revisions as 

applicable or its predecessor Safety Guide 23, “Onsite 

Meteorological Programs,” 1972, depending on the licensee 

commitments

6. Regulatory Guide 1.111, “Methods of Estimating Atmospheric 

Transport and Dispersion of Gaseous Effluents in Routine Release 

from Light-Water-Cooled Reactors,” July 1977, and other revisions 

as applicable

7. Regulatory Guide 4.1, “Programs for Monitoring Radioactivity in the 

Environs of Nuclear Power Plants,” June 2009, and other revisions 

as applicable

8. Regulatory Guide 4.13, “Testing, and Procedural Specifications for 

Thermoluminescence Dosimetry: Environmental Application,” July 

1977, and other revisions as applicable

9. Regulatory Guide 4.15, “Quality Assurance for Radiological 

Monitoring Programs (Normal Operations), Effluent Streams and the 

Environment,” July 2007, and other revisions as applicable

10. NUREG-1301, “Offsite Dose Calculation Manual Guidance: 

Standard Radiological Effluent Controls for Pressurized Water 

Reactors”

11. NUREG-1302 “Offsite Dose Calculation Manual Guidance: Standard 

Radiological Effluent Controls for Boiling Water Reactors”

12. NRC Branch Technical Position, “An Acceptable Radiological 

Environmental Monitoring Program,” November 1979

13. NUREG/CR-5569, Rev. 1, “Health Physics Positions Data Base”

14. NUREG/CR-6204, “Questions and Answers Based on Revised 10 

CFR Part 20”

15. NUREG-1736, “Consolidated Guidance: 10 CFR Part 20, Standards 

for Protection Against Radiation” November 1979
Issue Date: 02/20/17 48 IMC 1245, Appendix C3

16. NEI-07-07, “Industry Groundwater Protection Initiative - Final 

Guidance Document,” August 2007

17. NEI-08-08A, “Guidance for Life Cycle Minimization of Contamination 

– with SER (ML093220530)

18. EPRI TR-1016099, “Groundwater Protection Guidelines for Nuclear 

Power Plants,” 2008

19. Liquid Radioactive Release Lessons Learned Task Force Final 

Report September 2006 (ADAMS Accession Number 

ML062650312)

20. TI 2515/173, “Review of the Implementation of the Industry Ground 

Water Protection Voluntary Initiative”

21. TI 2515/185, “Follow up on the Industry’s Groundwater Protection 

Initiative”

22. Environmental Measurement Laboratory, HASL-300 Procedures 

Manual U.S. Department of Energy, New York, NY

23. NCRP Report 47, “Tritium Measurement Techniques,” 1975

24. NCRP Report 50, “Environmental Radiation Measurements,” 1976

25. NCRP Report 58, “A Handbook of Radioactivity Measurements 

Procedures,” 2nd Edition, 1984

26. NCRP Report 94, “Exposure of the Population in the United States 

and Canada from Natural Background Radiation (Supersedes 

NCRP Report 45),” 1987

27. ANSI N545-1975, “Performance, Testing, and Procedural 

Specification for TLD, Environmental Application,” and other 

revisions as applicable

28. ANSI/HPS N13.37-2014, “Environmental Dosimetry – Criteria for 

System Design and Implementation,” and other revisions as 

applicable

29. Regulatory Guide 1.109, “Calculation of Annual Doses to Man from 

Routine Releases of Reactor Effluents for the Purpose of Evaluating 

Compliance with 10 CFR 50, Appendix I,” October 1977

30. ANSI N13.1-1999, “Sampling and Monitoring Releases of Airborne 

Radioactive Substances from the Stacks and Ducts of Nuclear 

Facilities”
Issue Date: 02/20/17 49 IMC 1245, Appendix C3

31. NUREG-0800 BTP 11-3, “Design Guidance for Solid Radioactive 

Waste Management Systems Installed in Light-Water-Cooled 

Nuclear Power Reactor Plants”

EVALUATION 

CRITERIA: Upon completion of this ISA, the inspector should be able to:

1. Discuss the specific regulatory requirements that require a licensee 

to have a REMP and the process for making revisions to the REMP.

2. Discuss the specific environmental sampling techniques (water, 

milk, air iodine, air particulate, vegetation, fish, and soil/sediment) 

required to be collected in accordance with the REMP and ODCM.

3. Discuss and show how to calculate standard data reduction 

techniques, including MDA and LLD highlighted in NRC and industry 

publications. 

4. Discuss the NRC expectations for a licensee’s QA program for 

REMP including the use and bases for inter-laboratory and intralaboratory comparisons.

5. Discuss the Land Use Census. List the principle uses of land 

around the assigned facility.

6. Discuss the NRC requirements for a Meteorological Monitoring 

Program, including calibration methodology for wind direction, wind 

speed, and delta temperature. Compare this with that described in 

the licensee’s ODCM.

7. Show how to determine X/Q and D/Q, and annual average data and 

discuss when the values of X/Q and D/Q in the ODCM should be 

updated.

8. Discuss the principle parts of a REMP Annual Report.

9. Discuss the contents of the licensee’s ODCM and explain and 

compare the contents to NUREG-1301/1302 (or Branch Technical 

Position, November 1979).

10. Describe the man-made and natural radiation exposure pathways 

(fission/activated products and the source of the natural background 

radiation) that are present at your assigned facility.

11. Explain the Fundamentals of Laboratory QA/QC Policy and its 

implementation.

12. Identify the specific radiological dose limits for your assigned facility 

as described in the REMP and ODCM.
Issue Date: 02/20/17 50 IMC 1245, Appendix C3

13. Identify the specific environmental sample requirements for your 

assigned facility as described in the REMP and ODCM.

14. Discuss the sequential relationship between Radiological Effluent 

Control and the REMP.

15. Discuss the elements of a ground water protection program.

TASKS: 1. Locate a copy of the REMP section from the UFSAR and ODCM for 

your assigned facility.

2. Review the requirements of the REMP, including review of the 

REMP Annual Report.

3. Review NUREG-1301/1302 (or Branch Technical Position, 

November 1979) and bases.

4. Review the radiological measurement instrument data (proportional 

counter, gamma spectroscopy, liquid scintillation counter, and TLD 

reader).

5. Review counting statistics and data reduction, including minimum 

detectable activity (MDA) and lower limits of detection (LLDs).

6. Review meteorological monitoring requirements and calibration 

results for wind direction, wind speed, and delta temperature.

7. Review the sequential relationship between Radiological Effluent 

Controls and the REMP.

8. Review man-made and natural radiation exposure pathways 

(fission/activated products and the source of the natural background 

radiation).

9. Review the Fundamental of Laboratory QA/QC Policy and its 

implementation.

10. Locate the acceptance criteria to implement the voluntary 

Groundwater Protection Initiative in NEI 07-07.

DOCUMENTATION: Document completion on the Health Physics Inspector Technical 

Proficiency Level Signature Card and Certification Item ISA-HP-9.
Issue Date: 02/20/17 51 IMC 1245, Appendix C3

Health Physics Inspector Individual Study Activity

TOPIC: (ISA-HP-10) Radioactive Solid Waste Processing and Radioactive Material 

Handling, Storage, and Transportation

PURPOSE: The purpose of this ISA is to familiarize you with the regulatory bases and 

historical agency positions in the area of radioactive material processing 

and transportation and provide you the technical knowledge to conduct 

inspections using procedure 71124.08, “Radioactive Solid Waste

Processing and Radioactive Material Handling, Storage, and 

Transportation.” The objective of this inspection procedure is to verify the 

effectiveness of the licensee’s programs for processing, handling, storage, 

and transportation of radioactive material.

The NRC requires that licensee’s ensure adequate protection of public 

health and safety from exposure to radioactive material released to the 

public domain as a result of routine operations which include processing 

and shipment. The licensee’s radioactive material processing and 

shipping programs are required by Criterion 60 of Appendix A to 10 CFR 

Part 50 and must comply with the requirements of 10 CFR Parts 20, 61, 

and 71 and Department of Transportation (DOT) regulations contained in 

49 CFR Parts 100-189.

COMPETENCY

AREA: TECHNICAL AREA EXPERTISE

LEVEL 

OF EFFORT: 40 HOURS

REFERENCES: 1. 10 CFR Parts 20, 61, and 71

2. 10 CFR 20.1801-1802

3. 10 CFR 20.1904-1905

4. 10 CFR Part 20, Appendix G

5. 10 CFR Part 50, Appendix A, Criterion 60

6. 10 CFR 30.41

7. 10 CFR Part 37

8. 49 CFR Parts 100-189, including Part 172, Subpart H, “Training”

9. Radioactive waste system and facility description from the UFSAR 

for a facility designated by your supervisor

10. Plant Annual Radioactive Effluent Release Report

11. Offsite Dose Calculation Manual
Issue Date: 02/20/17 52 IMC 1245, Appendix C3

12. NRC Branch Technical Position, Waste Form Technical Position, 

Rev. 1, January 1991 (ADAMS Accession Number ML033630746)

13. NRC Branch Technical Position, Waste Form Technical Position,

Rev. 1, February 2015 (Vol, 1 - ADAMS Accession Number 

ML12254B065, Vol.2 - ADAMS Accession Number ML12326A611)

14. NUREG-1608, “Categorizing and Transporting Low Specific Activity 

Materials and Surface Contaminated Objects”

15. NUREG-1660, “U.S. - Specific Schedules of Requirements for 

Transport of Specified Types of Radioactive Material Consignments”

16. NUREG/BR-0204, “Instructions for Completing NRC’s Uniform Lowlevel Radioactive Waste Manifest”

17. IE Bulletin 79-19, “Packaging of Low Level Radioactive Waste for 

Burial,” August 10, 1979

18. Information Notice 89-13, “Alternative Waste Management 

Procedures in Case of Denial of Access to Low Level Waste 

Disposal Sites,” February 8, 1989

19. Information Notice 85-92, “Surveys of Waste Before Disposal from 

Nuclear Reactor,” December 2, 1985

20. Information Notice 86-20, “Low-Level Radioactive Waste Scaling 

Factors, 10 CFR Part 61,” March 28, 1986

21. Information Notice 90-50, “Minimization of Methane Gas in Plant 

Systems and Radioactive Waste Shipping Containers,” August 8, 

1990

22. Information Notice 92-62, “Emergency Response Information 

Requirements for Radioactive Material Shipments,” August 24, 1992

23. Information Notice 92-72, “Employee Training and Shipper 

Registration Requirements for Transporting Radioactive Materials,”

October 28, 1992

24. Information Notice 04-13, “Registration, Use and Quality Assurance 

Requirements for NRC Certified Transportation Packages,” June 30, 

2004

25. Generic Letter 95-09, “Monitoring and Training of Shippers and 

Carriers of Radioactive Materials,” November 3, 1995

26. Generic Letter 81-38, “Storage of Low-Level Radioactive Wastes at 

Power Reactor Sites,” November 10, 1981 
Issue Date: 02/20/17 53 IMC 1245, Appendix C3

27. IE Bulletin 80-10, “Contamination of Nonradioactive System and 

Resulting Potential for Unmonitored, Uncontrolled Release of 

Radioactivity to Environment,” May 6, 1980

28. Regulatory Issue Summary 2008-32, “Interim Low Level Radioactive 

Waste Storage at Reactor Sites,” December 30, 2008

29. Regulatory Issue Summary 2014-03, “Notice of 10 CFR Part 37 

Implementation Deadline for NRC Licensees,” March 13, 2014

30. ANSI/ANS-40.37-1993, “Mobile Radioactive Waste Processing 

Systems,” and other revisions as applicable

31. NUREG/CR-5569, Rev. 1, “Health Physics Positions Data Base”

32. NUREG/CR-6204, “Questions and Answers Based on Revised 10 

CFR Part 20”

33. NUREG-1736, “Consolidated Guidance: 10 CFR Part 20, Standards 

for Protection Against Radiation”

34. Regulatory Guide 1.143, “Design Guidance for Radioactive Waste 

Management Systems, Structures and Components Installed in 

Light Water- Cooled Nuclear Power Plants,” November 2001

35. SECY-94-198, “Review of Existing Guidance Concerning the 

Extended Storage of Low Level Radioactive Waste,” August 1, 1994

EVALUATION

CRITERIA: Upon completion of this ISA, the inspector should be able to:

1. Discuss the station’s recent radiological effluent release report and 

the types and amounts of radioactive waste that the licensee has 

disposed in the past year. 

2. Discuss and highlight the key aspects of the station’s solid and 

liquid radioactive waste processing systems, and their operation, as 

described in the station’s process control program (PCP) and the 

updated final safety analysis report (UFSAR). 

3. Discuss what the NRC expectations are relative to non-operational 

or abandoned radioactive waste processing equipment as described 

in NRC guidance in Bulletin 80-10, and NUREG/CR-5569.

4. Discuss the purpose and principal requirements of 10 CFR Part 61 

and NRC waste classification and characterization guidance. 

5. Discuss NRC guidance in the area of collection of representative 

samples of waste for the 10 CFR Part 61 program and how changes 

in waste streams should be identified for purposes of waste 

characterization and classification. 
Issue Date: 02/20/17 54 IMC 1245, Appendix C3

6. Discuss the packaging, labeling and marking requirements for 

various types of radioactive materials packages expected to be 

shipped from the facility, as presented in NUREG-1660 and 49 

CFR100-189. 

7. Discuss the manifesting, labeling, and placarding requirements for 

non-exempt types of radioactive materials packages relative to 

NUREG-1660 requirements. 

8. Discuss the allowable radiation dose rates (e.g., contact, 1 meter)

and contamination limits for shipment of packages of radioactive 

material specified in regulatory documents 49 CFR100-189. The 

limits include transport vehicle dose limits, cab limits, and package 

limits, as appropriate.

9. Discuss Hazmat training and emergency response program 

requirements relative to guidance in Information Notices 92-62, 92-

72, 95-09, and 49 CFR 100-189.

10. Discuss the principal aspects of classification of low-level 

radioactive waste as outlined in 10 CFR Part 61.

11. Discuss Quality Assurance Requirements for a radioactive materials 

shipping program, as described in 10 CFR 71 including the 

collection of representative samples and 10 CFR 20, Appendix G.

TASKS: 1. Review and familiarize yourself with the underlined references.

Specifically, identify the purpose of each document and what 

guidance the document provides.

2. Locate a copy of the radioactive waste section from the UFSAR of 

your designated facility. Review the design and operation of the 

radioactive waste systems. Identify the various sources of liquid 

and solid radioactive waste, waste streams, and technologies 

associated with liquid radioactive waste processing for the facility.

3. Review the requirements for the transfer and receipt of radioactive 

material as specified in 10 CFR 20 and 10 CFR 71, including 

reporting requirements for problems identified. 

4. Review and highlight the key aspects and requirements for low-level 

radioactive waste disposal as outlined in 10 CFR 61 and 71, and the 

burial site license. 

5. Review the contents of the licensee’s process control program 

(PCP), use of scaling factors for hard to detect nuclides, and the 

waste form and characteristic requirements for disposal of solid 

radioactive waste. 
Issue Date: 02/20/17 55 IMC 1245, Appendix C3

6. Review the requirements in the area of training and emergency 

response as specified in 49 CFR100-189. Identify minimum training 

requirements and minimum emergency response requirements.

7. Discuss how one would verify that a licensee is checking to see if 

another licensee is authorized to receive a radioactive material 

shipment package. 

8. Meet with your supervisor or a qualified inspector to discuss any 

questions you may have as a result of this activity. Discuss 

answers to the questions listed under the Evaluation Criteria section 

of this study guide with your supervisor.

Scenario A

On October 7, 2003, a low level radioactive waste shipment consisting of two box containers 

(volume of approximately 1400 cubic-foot each) was prepared by the licensee and offered to a 

carrier for transport to a waste processing contractor. Each of the sea-land (box) containers 

housed plastic bagged waste with one containing “Green is Clean” potentially contaminated dry 

waste and the other dry active waste (DAW). The sea-land containers were loaded back-toback on an open, flatbed trailer with the DAW container loaded toward the rear of the vehicle. 

The shipment was consigned as exclusive use and categorized as class 7 (radioactive) 

materials, low specific activity, containing a total activity of about 67 millicuries of primarily 

mixed activation products. The shipment departed the site at approximately 2:00 p.m. on 

October 7, 2003, and arrived at the waste processing facility about 10 hours later. 

On October 8, 2003, radiation measurements performed by waste processing contractor 

personnel on the exterior surface of the packages (the sea-land containers) identified a highly 

localized area of elevated radiation on the external surface of the DAW container that exceeded 

the Department of Transportation (DOT) limit provided in 49 CFR 173.441. Specifically, a coinsized (one-inch diameter) spot measuring 250 millirem/hour was identified on the external 

surface of the sea-land container’s rear door, about three and one-half feet up from the bottom 

of the package and one and one-half inches lateral to a vertical metal bar used to latch the 

container’s door. Package and vehicle surveys performed by the licensee prior to the 

shipment’s departure on October 7, 2003, documented a maximum package surface radiation 

level of 33 millirem/hour on the DAW filled container that was located in the same general 

location as the coin-sized “hot spot” identified at waste processing contractor.

Questions:

1. What two Codes of Federal Regulations apply to this condition? (NRC/DOT)

2. What is the DOT limit on external surface of containers?

3. What are the possible reasons for the licensee “missing” the hot spot on the container?

DOCUMENTATION: Document completion on the Health Physics Inspector Technical 

Proficiency Level Signature Card and Certification Item ISA-HP-10.
Issue Date: 02/20/17 56 IMC 1245, Appendix C3

Health Physics Inspector Individual Study Activity

TOPIC: (ISA-HP-11) Significance Determination Process - Occupational Radiation 

Safety

PURPOSE: The purpose of this ISA is familiarize you with the Occupational Radiation 

Significance Determination Process (SDP), explain the bases of each SDP 

branch, provide technical and policy references, and present practical 

exercises/scenarios that challenge you to use the SDP. The risk 

significance of inspection findings are evaluated using the SDP. 

COMPETENCY

AREA: INSPECTION, 

TECHNICAL AREA EXPERTISE

REGULATORY FRAMEWORK

LEVEL 

OF EFFORT: 36 hours

REFERENCES: 1. IMC 0609, Appendix C, “Occupational Radiation Safety Significance 

Determination Process”

2. IMC 0308, Attachment 3, Appendix C, “Technical Basis for 

Occupational Radiation Safety Significance Determination Process”

3. SDP Branch on ALARA

a. NRC Inspection Report (50-483-0017) (ADAMS Accession 

Number ML003757369), and NRC Responses to Licensee 

Contestation for White Findings at Callaway (EA-00-208, May, 

2001) (ADAMS Accession Number ML011270225)

b. NUREG-0713, Latest Rev, “Occupational Radiation Exposure 

at Commercial Nuclear Power Reactors and Other Facilities”

c. Supplementary Information, 10 CFR 20 Final Rule, Subpart B, 

“Radiation Protection Programs,” Federal Register Vol. 56, No. 

98, pages 23366-23367, May 21, 1991 (ADAMS Accession 

Number ML051790366)

4. SDP Branch on Overexposure

a. Regulatory Guide 8.38, “Control of Access to High and Very 

High Radiation Areas in Nuclear Power Plants,” May 2006

b. Nuclear Regulatory Commission Enforcement Manual 

(http://www.nrc.gov/aboutnrc/regulatory/enforcement/guidance.html)

5. SDP Branch on Substantial Potential for Overexposure (SPFO)
Issue Date: 02/20/17 57 IMC 1245, Appendix C3

a. NUREG/CR-5569, Rev. 1, “Health Physics Positions Data 

Base,” HPPOS-232

b. Information Notice 88-63, “High Radiation Hazards from 

Irradiated In-core Detectors and Cables,” August 15, 1988

c. Information Notice 88-63, Supplement 1, “High Radiation 

Hazards from Irradiated In-core Detectors and Cables,” 

October 5, 1990

d. Information Notice 88-63, Supplement 2, “High Radiation 

Hazards from Irradiated In-core Detectors and Cables,” June 

25, 1991

e. Information Notice 02-03, “Highly Radioactive Particle Control 

Problems During Spent Fuel Pool Cleanout,” January 10, 2002

f. Davis-Besse Special Inspection Reports 02-06 and 02-16

(ADAMS Accession Number ML030070606)

6. SDP Branch on Ability to Assess Dose

a. Davis-Besse Special Inspection Reports 02-06 and 02-16

(ADAMS Accession Number ML030070606)

7. IMC 0612, Appendix B, “Issue Screening”

8. IMC 0612, Appendix E, “Examples of Minor Issues,” Section 6, 

“Health Physics”

EVALUATION

CRITERIA: Upon completion of this ISA, the inspector should be able to:

1. Explain the purpose, objectives and applicability of the SDP 

process.

2. Compare and contrast an “observation” and a “finding.”

3. Describe the process for providing feedback and recommendations 

to improve the SDP process.

4. Be aware of the SDP and Enforcement Review Panel procedures, 

and the process for appealing the NRC characterization of 

inspection process.

5. Describe and discuss the objective of the Occupational Radiation 

cornerstone.

6. State the technical bases for the two branches of the Occupational 

SDP (ALARA and Exposure Control).
Issue Date: 02/20/17 58 IMC 1245, Appendix C3

7. Define the safety significance, and give examples of Green, White, 

Yellow and Red findings in each of the branches of this SDP.

8. For the ALARA Branch of the SDP, you should be able to:

9. Process ALARA findings for a job through the branch, using 

example training scenarios.

10. Explain why a WHITE finding may not be an adequate basis for 

issuing a notice of violation against Part 20.1101(b).

11. Explain the basis for the “Greater than 4 Occurrences” logic gate 

and the rationale for issuing a WHITE finding.

12. For the Overexposure Branch of the SDP, you should be able to 

process exposure findings through the branch, using example 

training scenarios. (You may use examples from events at 

Palisades – ML14336A624, Perry – ML11187A121 and the 

examples in given in IMC 0609, Appendix C)

13. For the Substantial Potential for Overexposure (SPFO) Branch of 

the SDP, you should be able to:

14. Discuss and explain the SPFO enforcement policy tool, and give 

examples of SPFO findings.

15. Process exposure findings through the branch, using example 

training scenarios from NRC Information Notices and other sources.

16. Explain how findings involving shallow dose equivalents from 

discrete hot particle exposures are treated in this branch, and 

understand the bases for this position. Explain why, for very high 

activity particles with a significant gamma component (e.g., 100's of 

mCi of Co-60), are treated differently relative to SPFO.

17. Explain why an event with a SPFO that occurs in a Very High 

Radiation area merits a yellow finding.

18. For the Ability to Assess Dose Branch of the SDP, you should be 

able to: 

a. Discuss, and give examples of the systemic type of problems 

that could lead to white findings in this branch.

b. Process findings through the branch, using example training 

scenarios.

TASKS: 1. Review MC 0609 and focus on Appendix C, Section IV, 

“Significance Determination Process for Occupational Radiation 

Safety.”
Issue Date: 02/20/17 59 IMC 1245, Appendix C3

2. Review the underlined references pertinent to each SDP branch. 

3. Determine a facility’s current three year rolling average (TYRA) and 

determine the median collective dose for both PWRs and BWRs

from the latest official cumulative dose reported by the licensee.

4. Find all applicable Part 20 annual dose limits pertinent to the SDP.

5. List and describe areas/locations in BWR and PWR where whole 

body, DDE overexposures have occurred. Specifically focus on 

areas with the potential for rapid, high dose rate increases 

described in Regulatory Guide 8.38, Appendix B.

6. Develop an event scenario of a “near-miss event”, where a worker 

received no dose, but this event could result in a SPFO finding. 

7. Develop a SPFO scenario involving airborne material intake, and 

determine the range of significance determinations. Explain why 

DDE-related SPFOs are considered more risky to nuclear power 

plant workers.

8. Review selected Regional inspection reports selected by your 

mentor that resulted in Green and White finding. Process the 

finding through the SDP. Determine if you agree with the results in 

the inspection report. Discuss any differences you may have on the 

SDP results with a qualified Health Physics inspector.

9. Whenever possible, attend a significance determination and 

enforcement review panel (SERP) related to this SDP. Discuss 

rationale for the outcome/resolution of the panel with a qualified 

Health Physics inspector.

10. Meet with your supervisor or a qualified inspector to discuss any 

questions you may have as a result of this training activity. Discuss 

the answers to the questions and your work products listed under 

the Evaluation Criteria section of this study guide with your 

supervisor.

DOCUMENTATION: Document completion on the Health Physics Inspector Technical 

Proficiency Level Signature Card and Certification Item ISA-HP-11.
Issue Date: 02/20/17 60 IMC 1245, Appendix C3

Health Physics Inspector Individual Study Activity

TOPIC: (ISA-HP-12) Significance Determination Process - Public Radiation Safety: 

Radioactive Material Control

PURPOSE: The purpose of this ISA is to familiarize you with the Radioactive Material 

Control portion of the Public Radiation Safety Significance Determination 

Process (SDP), explain the bases of this SDP branch, provide technical 

and policy references, and present practical exercises/scenarios that 

challenge you to use the SDP. The risk significance of inspection findings 

are evaluated using the SDP. 

COMPETENCY

AREA: INSPECTION

TECHNICAL AREA EXPERTISE

REGULATORY FRAMEWORK,

LEVEL 

OF EFFORT: 24 hours

REFERENCES: 1. IMC 0609, Appendix D, “Public Radiation Safety Significance 

Determination Process,” Section III, “Radioactive Material Control 

Program”

2. IMC 0308, Attachment 3, Appendix D, “Technical Basis for Public 

Radiation Safety Significance Determination Process”

3. IMC 0612, Appendix B, “Issue Screening”

4. IMC 0612, “Appendix E, “Examples of Minor Issues,” Section 6, 

Health Physics

5. 10 CFR Part 20, “Subpart F, Surveys and Monitoring”

6. 10 CFR Part 20, Subpart K, “Waste Disposal”

7. NUREG/CR-5569, Rev. 1, “Health Physics Positions Data Base,” 

HPPOS-42, 43, 44, 45, 48, 71, 72, 73, 79, 106, 138, 171, 189, 190, 

221, 250, 300

8. IE Circular 81-07, “Control of Radioactively Contaminated Material,” 

May 14, 1981

9. Information Notice 83-05, “Obtaining approval for disposing of verylow-level radioactive waste - 10 CFR Section 20.302 (current 10 

CFR 20.2002),” February 24, 1983

10. Information Notice 85-92, “Survey of Wastes Before Disposal from 

Nuclear Reactor Facilities,” December 2, 1985
Issue Date: 02/20/17 61 IMC 1245, Appendix C3

11. Information Notice 86-90, “Requests to dispose of very low-level 

radioactive waste pursuant to 10 CFR 20.302 (current 10 CFR 

20.2002),” November 3, 1986

12. Information Notice 88-22, “Disposal of Sludge from Onsite Sewage 

Treatment Facilities at Nuclear Power Stations,” May 12, 1988

13. NRC Inspection Report (ADAMS Accession Number ML013650258) 

and NRC Responses to Licensee Contestation for White Findings at 

Comanche Peak (ADAMS Accession Number ML022900740)

EVALUATION

CRITERIA: Upon completion of this ISA, the inspector should be able to:

1. Explain the purpose, objectives and applicability of the Radioactive 

Material Control SDP process.

2. Define the safety significance, and give examples of Green, White, 

Yellow and Red findings of this SDP branch.

3. Process a finding through the SDP branch.

4. Explain the NRR “no detectable” licensed radioactive material 

release policy for surface contamination, volumetric contamination, 

and difficult to detect radionuclides.

5. Describe the different ways to assess a finding which involves a “hot 

particle.”

6. Describe the controls the licensee must have to demonstrate that 

licensed radioactive material is still under their control after it left the 

radiation controlled area, but is still in a controlled area.

7. Describe the difference between a release limit and the “no 

detectable” criteria.

8. Describe the minimum detection sensitivity criteria that licensees 

must use for radiation surveys of potentially contaminated material.

TASKS: 1. Review IMC 0609, Appendix D, Section III, “Radioactive Material 

Control Program.”

2. Review each of the underlined references to learn about the NRC’s 

policies and practices for the release of radioactive materials.

3. Go to the ROP web-site and review recent Green and White 

findings in the area of public radiation safety.

4. For those inspection findings related to radioactive material control, 

process the finding through the Radioactive Material Control Branch 

of the Public Radiation Safety SDP. Determine if you agree with the 
Issue Date: 02/20/17 62 IMC 1245, Appendix C3

results in the inspection report. Discuss any differences you may 

have with a qualified health physics inspector.

5. Whenever possible, attend a significance determination and 

enforcement review panel (SERP) related to this SDP. Discuss 

rationale for the outcome/resolution of the panel with a qualified 

health physics inspector.

6. Meet with your supervisor or a qualified inspector to discuss any 

questions you may have as a result of this training activity. Discuss 

the answers to the questions and your work products listed under 

the Evaluation Criteria section of this study guide with your 

supervisor.

DOCUMENTATION: Document completion on the Health Physics Inspector Technical 

Proficiency Level Signature Card and Certification Item ISA-HP-12.
Issue Date: 02/20/17 63 IMC 1245, Appendix C3

Health Physics Inspector Individual Study Activity

TOPIC: (ISA-HP-13) Significance Determination Process Public Radiation Safety: 

Effluent Release Program

PURPOSE: The purpose of this ISA is to familiarize you with the Radioactive Effluent 

portion of the Public Radiation Safety Significance Determination Process 

(SDP), explain the bases of this branch, provide technical and policy 

references, and present practical exercises/scenarios that challenge you to 

use the SDP. The risk significance of inspection findings are evaluated 

using the SDP. 

COMPETENCY

AREA: INSPECTION

TECHNICAL AREA EXPERTISE

REGULATORY FRAMEWORK

LEVEL 

OF EFFORT: 32 hours

REFERENCES: 1. IMC 0609, Appendix D, Section I, “Public Radiation Safety 

Significance Determination,” Radioactive Effluent Release Program”

2. IMC 0308, Attachment 3, Appendix D, “Technical Basis for Public 

Radiation Safety”

3. IMC 0612, Appendix B, “Issue Screening”

4. IMC 0612, Appendix E, “Examples of Minor issues”, Section 6, 

Health Physics

5. 10 CFR Part 50, Appendix A, Criteria 60, 61, 63, and 64

6. 10 CFR Part 20 Subpart D, “Radiation Dose Limits for Individual 

Members of the Public”

7. 10 CFR Part 50, Appendix I, “Numerical Guidelines for Design 

Objectives and Limiting Conditions for Operation to Meet the Criteria 

for ALARA

8. 40 CFR 190, “Environmental Radiation Protection Standards for 

Nuclear Power Plant Operations”

9. NUREG/CR-5569, Revision 1, “Health Physics Positions Data Base” 

HPPOS 4, 6, 7, 8, 40, 88, 102, 122, 170, 171, 212, 223, 229, 326

10. NUREG-0543, “Methods for Demonstrating Compliance with the 

EPA Uranium Fuel Cycle Standard (40 CFR Part 190)”

11. NUREG-0133, “Preparation of Radiological Effluent Technical 

Specifications for Nuclear Power Plants”
Issue Date: 02/20/17 64 IMC 1245, Appendix C3

12. NUREG-1301, “Offsite Dose Calculation Manual Guidance: 

Standard Radiological Effluent Controls for Pressurized Water 

Reactors”

13. NUREG-1302, “Offsite Dose Calculation Manual Guidance: 

Standard Radiological Effluent Controls for Boiling Water Reactors”

14. Regulatory Guide 1.21, “Measuring, Evaluating, and Reporting 

Radioactivity in Solid Wastes and Releases of Radioactive Materials 

in Liquid and Gaseous Effluents From Light-Water-Cooled Nuclear 

Power Plants,” June 2009, and other revisions as applicable

15. Regulatory Guide 1.23, “On-site Meteorological Programs for 

Nuclear Power Plants,” March 2007, and other revisions as 

applicable or its predecessor Safety Guide 23, “Onsite 

Meteorological Programs,” 1972, depending on the licensee 

commitments

16. Regulatory Guide 1.109, “Calculation of Annual Doses to Man From 

Routine Releases of Reactor Effluents for the Purpose of Evaluating 

Compliance with 10 CFR Part 50, Appendix I,” October 1977, and 

other revisions as applicable

17. Regulatory Guide 1.143, “Design Guidance for Radioactive Waste 

Management Systems, Structures, and Components Installed in 

Light-Water-Cooled Nuclear Power Plants,” November 2001, and 

other revisions as applicable

18. Regulatory Guide 4.15, “Quality Assurance for Radiological 

Monitoring Programs (Normal Operations) - Effluent Streams and 

the Environment,” July 2007, and other revisions as applicable

19. Regulatory Guide 1.110, “Cost-Benefit Analysis for Radioactive 

Waste Systems for Light-Water-Cooled Reactors,” March 1976, and 

other revisions as applicable

20. Regulatory Guide 1.111, “Methods for Estimating Atmospheric 

Transport and Dispersion of Gaseous Effluents in Routine Releases 

from Light-Water-Cooled Reactors,” July 1977, and other revisions 

as applicable

21. Regulatory Guide 1.112, “Calculation of Releases of Radioactive 

Materials in Gaseous and Liquid Effluents from Light-Water-Cooled 

Power Reactors,” March 2007, and other revisions as applicable

22. Regulatory Guide 1.113, “Estimating Aquatic Dispersion of Effluents 

from Accidental and Routine Reactor Releases for the Purpose of 

Implementing Appendix I,” April 1977, and other revisions as 

applicable
Issue Date: 02/20/17 65 IMC 1245, Appendix C3

23. IE Bulletin 80-10, “Contamination of Nonradioactive System and 

Resulting Potential for Unmonitored, Uncontrolled Release of 

Radioactivity to Environment,” May 6, 1980

24. IE Circular 79-21, “Prevention of Unplanned releases of 

Radioactivity,” October 19, 1979

25. Information Notice 79-21, “Transportation and Commercial Burial of 

Radioactive Materials,” September 5, 1979

26. Information Notice 91-56, “Potential Radioactive Leakage to Tank 

Vented to Atmosphere,” September 19, 1991

27. Generic Letter 79-03, “Offsite Dose Calculation Manual,” January 

18, 1979

28. Generic Letter 79-06, “Contents of the Offsite Dose Calculation 

Manual,” February 14, 1979

29. Generic Letter 84-12, “Compliance with 10 CFR Part 61 and 

Implementation of Radioactive Effluent Technical Specifications and 

Attendant Process Control Program,” April 30, 1984

30. Generic Letter 89-01, “Implementation of Programmatic and 

Procedural Controls for Radiological Effluent Technical 

Specifications,” January 31, 1989

EVALUATION

CRITERIA: Upon completion of this ISA, the inspector should be able to:

1. Explain the purpose, objectives and applicability of the Effluent 

Release Program SDP process.

2. Define the safety significance, and give examples of Green, White, 

Yellow and Red findings of this SDP branch.

3. Process a finding through the SDP branch.

4. Explain what is meant by “impaired ability to assess dose.”

5. Explain the similarity and differences between Appendix I to 10 CFR 

Part 50 and 40 CFR Part 190.

6. Compare and contrast the dose assessment methodology in 10 

CFR Part 20 to that in Appendix I of 10 CFR Part 50.

7. Explain the technical and operational differences, if any, between a 

licensee who still has the Radiological Effluent Technical 

Specifications and a licensee who implemented the guidance in 

Generic Letter 89-01.
Issue Date: 02/20/17 66 IMC 1245, Appendix C3

TASKS: 1. Review MC 0609, Appendix D, Section I, “Radioactive Effluent 

Release Program.”

2. Review each of the underlined references to become familiar with 

the accepted methodologies for the control and release of effluents.

3. Go to the ROP web-site and review recent Green and White 

findings in the area of public radiation safety.

4. For those inspection findings related to the effluent release program, 

process the finding through the Effluent Release Branch of the 

Public Radiation Safety SDP. Determine if you agree with the 

results in the inspection report. Discuss any differences you may 

have with a qualified health physics inspector.

5. Whenever possible, attend a significance determination and 

enforcement review panel (SERP) related to this SDP. Discuss 

rationale for the outcome/resolution of the panel with a qualified 

health physics inspector

6. Meet with your supervisor or a qualified inspector to discuss any 

questions you may have as a result of this training activity. Discuss 

the answers to the questions and your work products listed under 

the Evaluation Criteria section of this study guide with your 

supervisor.

DOCUMENTATION: Document completion on the Health Physics Inspector technical 

Proficiency Level Signature Card and Certification Item ISA-HP-13.
Issue Date: 02/20/17 67 IMC 1245, Appendix C3

Health Physics Inspector Individual Study Activity

TOPIC: (ISA-HP-14) Significance Determination Process - Public Radiation Safety: 

Environmental Monitoring Program 

PURPOSE: The purpose of this ISA is familiarize you with the Environmental 

Monitoring Program portion of the Public Radiation Safety Significance 

Determination Process (SDP), explain the bases of this SDP branch, 

provide technical and policy references, and present practical 

exercises/scenarios that challenge you to use the SDP. The risk 

significance of inspection findings are evaluated using the SDP.

COMPETENCY 

AREA: INSPECTION

TECHNICAL AREA EXPERTISE

REGULATORY FRAMEWORK

LEVEL

OF EFFORT: 24 hours

REFERENCES: 1. IMC 0609, Appendix D, “Public Radiation Safety Significance 

Determination Process,” Section II, “Radioactive Environmental 

Monitoring Program.”

2. IMC 0308, Attachment 3, Appendix D, “Technical Basis for Public 

Radiation Safety Significance Determination Process”

3. IMC 0612, Appendix B, “Issue Screening”

4. 10 CFR Part 50, Appendix A, Criteria 60 and 64

5. Generic Letter 79-65, “Radiological Environmental Monitoring 

Program Requirements - Enclosing Branch Technical Position, 

Revision 1,” November 27, 1979

6. Generic Letter 89-01, “Implementation of Programmatic and 

Procedural Controls for Radiological Effluent Technical 

Specifications,” January 31, 1989

7. NUREG-1301, “Offsite Dose Calculation Manual Guidance: 

Standard Radiological Effluent Controls for Pressurized Water 

Reactors”

8. NUREG-1302, “Offsite Dose Calculation Manual Guidance: 

Standard Radiological Effluent Controls for Boiling Water Reactors”

9. Regulatory Guide 1.21, “Measuring, Evaluating, and Reporting 

Radioactive Material in Liquid and Gaseous Effluents and Solid 

Waste,” June 2009, and other revisions as applicable
Issue Date: 02/20/17 68 IMC 1245, Appendix C3

10. Regulatory Guide 4.1, “Programs for Monitoring Radioactivity in the 

Environs of Nuclear Power Plants,” June 2009, and other revisions 

as applicable

11. Regulatory Guide 4.8, “Environmental Technical Specifications for 

Nuclear Power Plants,” December 1975. Also, see its revision, 

NRC Branch Technical Position, “An Acceptable Radiological 

Environmental Monitoring Program,” November 1979, and other 

revisions as applicable

12. Regulatory Guide 4.15, “Quality Assurance for Radiological 

Monitoring Programs (Normal Operations) - Effluent Streams and 

the Environment,” July 2007, and other revisions as applicable

EVALUATION

CRITERIA: Upon completion of this ISA, the inspector should be able to:

1. Define the safety significance, and give examples of minor 

performance deficiencies and Green findings of this SDP branch.

2. Process a finding through the SDP branch. 

3. Describe the different environmental pathways the program is 

designed to assess.

4. Describe the significance of the pathways monitored by the 

radiological environmental monitoring program.

5. Describe the minimum detection sensitivity criteria that licensees 

must use to analyze their environmental samples.

6. Describe the process a licensee needs to do to make changes to 

the scope of their radiological environmental monitoring program.

TASKS: 1. Review MC 0609, Appendix D, Section II, Radioactive 

Environmental Monitoring Program.

2. Review the underlined reference materials to become familiar with 

the characteristics of an acceptable environmental monitoring 

program. 

3. Go to the ROP web-site and review recent Green findings in the 

area of environmental monitoring.

4. For those inspection findings related to environmental monitoring 

programs, process the finding through the Environmental Monitoring 

Program branch of the Public Radiation Safety SDP. Determine if 

you agree with the results in the inspection report. Discuss any 

differences you may have with a qualified health physics inspector.
Issue Date: 02/20/17 69 IMC 1245, Appendix C3

5. Meet with your supervisor or a qualified inspector to discuss any 

questions you may have as a result of this training activity. Discuss 

the answers to the questions and your work products listed under 

the Evaluation Criteria section of this study guide with your 

supervisor.

DOCUMENTATION: Document completion on the Health Physics Inspector Technical 

Proficiency Level Signature Card and Certification Item ISA-HP-14.
Issue Date: 02/20/17 70 IMC 1245, Appendix C3

Health Physics Inspector Individual Study Activity

TOPIC: (ISA-HP-15) Significance Determination Process - Public Radiation Safety: 

Transportation Branch

PURPOSE: The purpose of this ISA is to familiarize you with the transportation branch 

of the Public Radiation Safety Significance Determination Process (SDP), 

explain the bases of the branch, provide technical and policy references, 

and present practical exercises/scenarios that will require you to use this 

branch of the SDP. The risk significance of inspection findings are 

evaluated using the SDP. 

COMPETENCY

AREA: INSPECTION

TECHNICAL AREA EXPERTISE

REGULATORY FRAMEWORK

LEVEL

OF EFFORT: 36 hours

REFERENCES: 1. IMC 0609, Appendix D, “Public Radiation Safety Significance 

Determination Process,” Section VII, “Transportation.”

2. IMC 0308, Attachment 3, Appendix D, “Technical Basis for Public 

Radiation Safety Significance Determination Process”

3. NRC Inspection Report and NRC Responses to Licensee 

Contestation for White Findings (VY – ML063110067, 

ML072340014; Prairie Island – ML090410466, ML091270080, 

ML100130231) 

4. NUREG-1660, “U.S.- Specific Schedules of Requirements for 

Transport of Specified Types of Radioactive Material Consignments”

5. NUREG-1608, “Categorizing and Transporting Low Specific Activity 

Materials and Surface Contaminated Objects”

6. NUREG/CR-6407, “Classification of Transportation Packing and Dry 

Spent Fuel Storage System Components According to Importance 

to Safety”

7. NUREG/CR-5569, “Health Physics Positions Data Base,” Section 

2.17 (Transportation and Shipping)

8. NUREG/CR-6204, “Questions and Answers Based on Revised 10 

CFR Part 20,” Sections 20.1904 and 20.1906

9. 10 CFR Part 20, 61, and 71

10. 49 CFR Parts 171-178, Subchapter C
Issue Date: 02/20/17 71 IMC 1245, Appendix C3

11. IMC 0609, Appendix D, Section VII, “Transportation,” Subsection a. 

“Radiation Limits Exceeded”

12. IMC 0609, Appendix D, Section VII, “Transportation,” Subsection b,

“Breach of Package During Transit”

13. IMC 0609, Appendix D, Section VII, “Transportation,” Subsection c. 

“Part 61 Finding”

14. IMC 0609, Appendix D, Section VII, “Transportation,” Subsection d. 

“Failure to Make Notifications or Provide Emergency Information”

a. Information Notice 92-62, “Emergency Response Information 

Requirements for Radioactive Material Shipments” August 24, 

1992

b. Information Notice 93-07, “Classification of Transportation 

Emergencies,” February 1, 1993

15. IMC 0609, Appendix D, Section VII, “Transportation,” Subsection e. 

“Certificates of Compliance”

a. NUREG-0383 (latest revision), “Directory of Certificates of 

Compliance for Radioactive Materials Packages,” Volume 2

EVALUATION

CRITERIA: Upon completion of this ISA, the inspector should be able to:

1. State the bases for the five sub-branches of the Transportation 

Branch of Public Radiation Safety SDP.

2. Define the safety significance, and give examples of Green, White, 

Yellow and Red findings in the sub-branches of this SDP Branch.

3. For the Radiation Limit Sub-Branch of the SDP, you should be able 

to:

a. Process findings for an event through the sub-branch, using 

example training scenarios. (Use the examples listed in the 

reference section)

b. Explain why five times the limit for external radiation is a 

WHITE finding, while five times the surface contamination limit 

gets only a GREEN finding.

4. For the Breach of Package Sub-Branch of the SDP you should be 

able to:

a. Define “breach”, “normal conditions of transport,” and “Type A”

and how these terms factor in when determining whether a 
Issue Date: 02/20/17 72 IMC 1245, Appendix C3

loss of containment of a Type A package is deemed a breach 

finding, for purposes of this SDP.

b. Define “breach”, “beyond normal conditions of transport”, and 

“Type B” and how these terms factor in when determining 

whether a loss of containment of a Type B package is deemed 

a breach finding, for purposes of this SDP.

c. Process findings through the branch, using example training 

scenarios.

5. For the Low Level Burial Ground Access Sub-Branch of the SDP, 

you should be able to:

a. Discuss and explain why waste under-classification merits a 

WHITE finding.

b. Process exposure findings through the branch, using example 

training scenarios.

6. For the Notifications or Emergency Information Sub-Branch of the 

SDP, you should be able to: 

a. For Block N1, give examples of non-compliances that result in 

GREEN and WHITE findings.

b. For Block N4, state the external radiation and surface 

contamination levels (five times the allowable Part 71 limits) 

for a typical spent fuel exclusive-use shipment that if 

exceeded, would result in a WHITE finding if not reported by 

the receiving facility.

c. Process findings through the branch, using example training 

scenarios.

7. For the Certificate of Compliance Sub-Branch of the SDP, you 

should be able to:

a. Define what a Certificate of Compliance (COC) is, what it does 

and discuss typical COC requirements (or conditions) for a 

Type B shipping container.

b. Process findings through the branch, using example training 

scenarios

TASKS: 1. Review MC 0609 and focus on Appendix D, “Public Radiation 

Safety Significance Determination Process.”

2. Review the references pertinent to each SDP sub-branch.
Issue Date: 02/20/17 73 IMC 1245, Appendix C3

3. Identify all applicable DOT radiation limits for exclusive use, Type A 

and B transport packages pertinent to this Sub-Branch.

4. Develop a scenario involving a breach of a Type A package (with 

loss of contents) that would result in only a GREEN finding.

5. Review Part 61 

a. Identify and define Class A, B, and C wastes.

b. Identify the eight minimum waste Part 61 

requirements/characteristics and be able to discuss the most 

likely problems encountered by power plant shipments.

6. Review 10 CFR 71.97 and identify four possible non-compliances 

that would result in GREEN findings. Develop a scenario that 

should result in a WHITE finding for Block N1.

7. Review 49 CFR 172.602 (Block N2), and list the minimum 

emergency information that must be provided by the shipper and be 

able to discuss why and how quickly this information must be 

provided to emergency responders. 

8. Develop a scenario for a Type B shipment of dewatered resins that 

leads to a White finding in the COC sub-Branch, and a Yellow 

finding in the Package Breach sub-Branch. Discuss with your 

supervisor or qualified HP inspector which finding(s) would be 

documented for this one event. 

9. Using an example COC for a typical Type B waste container, and 

using the NUREG-0383, sort and list six components relative to their 

importance to safety. Additional information on COCs may be found 

on http://rampac.energy.gov. 

10. Review selected regional inspection reports that resulted in Green 

and White finding in this area. Process the finding through the 

appropriate branch of the SDP. Determine if you agree with the 

results in the inspection report. Discuss any differences you may 

have on the SDP results with a qualified health physics inspector.

11. Whenever possible, attend a significance determination and 

enforcement review panel (SERP) related to this SDP. Discuss 

rationale for the outcome/resolution of the panel with a qualified 

health physics inspector.

12. Meet with your supervisor or a qualified inspector to discuss any 

questions you may have as a result of this training activity. Discuss 

the answers to the questions and your work products listed under 

the Evaluation Criteria section of this study guide with your 

supervisor.
Issue Date: 02/20/17 74 IMC 1245, Appendix C3

DOCUMENTATION: Document completion on the Health Physics Inspector Technical 

Proficiency Level Signature Card and Certification Item ISA-HP-15.
Issue Date: 02/20/17 75 IMC 1245, Appendix C3

Health Physics Inspector Individual Study Activity

TOPIC: (ISA-HP-16) Performance Indicators - Occupational Radiation Safety 

Cornerstone – Occupational Exposure Control Effectiveness and Public 

Radiation Safety Cornerstone - Radiological Effluent Technical 

Specifications (RETS)/Offsite Dose Calculation Manual (ODCM) 

Radiological Effluent Occurrences

PURPOSE: This ISA will help the inspector to understand the definition of an 

Occupational Radiation Safety Cornerstone PI and a Public Radiation 

Safety Cornerstone PI, to gain the ability to identify an individual 

operational occurrence, and to correctly classify the occurrence as a PI, or 

not.

One of the objectives of the Inspection program is to verify the 

Performance Indicators (PIs) reported by the licensee in each cornerstone 

area. Therefore, it is important for the inspector to be capable of 

determining whether an operational occurrence identified during an 

inspection is, or is not, reportable as a PI. 

COMPETENCY

AREA: INSPECTION

TECHNICAL AREA EXPERTISE

REGULATORY FRAMEWORK

LEVEL 

OF EFFORT: 18 Hours

REFERENCES: 1. NEI 99-02, (Latest Revision), “Regulatory Assessment Performance 

Indicator Guideline,” Section 2.5

2. NEI 99-02, (Latest Revision), “Regulatory Assessment Performance 

Indicator Guideline,” Frequently Asked Questions, pages 112 - 120

3. Standard Technical Specification, Section 6.12, “High Radiation 

Area Access Control”

4. Regulatory Guide 8.38, “Control of Access to High and Very High 

Radiation Areas in Nuclear Power Plants,” May 2006, and other 

revisions as applicable

5. NUREG/CR-6204, “Questions and Answers Based on Revised 10 

CFR Part 20,” Questions Nos. 49, 92, 218, 373, 385, 423, 441, 447, 

448

6. NUREG/CR-5569, Rev.1, “Health Physics Positions Data Base,”

Section 2.5

7. IMC 0608, “Performance Indicator Program”
Issue Date: 02/20/17 76 IMC 1245, Appendix C3

8. IMC 0308, Attachment 1, “Technical Basis for Performance 

Indicators”

9. IP 71151, “Performance Indicator Verification”

EVALUATION 

CRITERIA: Upon completion of this ISA, the inspector should be able to:

1. Identify which attributes of a Radiation Protection Program are 

covered by the PI in this area.

2. Discuss the several types of operational occurrences included in the 

PI definition, including which non-conformance with High Radiation 

Area Technical Specifications is pertinent.

3. Define “unintended occupational exposure” and “radiation safety 

barriers”.

4. Distinguish between an individual occurrence and concurrent 

operational occurrences.

5. Discuss the minimum significance “thresholds” of an occurrence in 

the PI definition. 

6. Discuss the “thresholds” associated with the colored PI significance 

bands.

7. Identify which elements of a Chemistry Program are covered by the 

PI in this area.

8. Discuss the types of radiological effluent occurrences included in 

the PI definition.

9. Discuss how “significant” individual occurrences (i.e., those that 

exceed the reporting criteria in 10 CFR Parts 20.2202 or 20.2203) 

are handled by the Reactor Oversight Process.

TASKS: 1. Locate a copy of the TS for the facility designated by your 

supervisor, and compare the High Radiation Access requirements to 

those in the STS.

2. Review NEI 99-02 section 2.5 for Occupational Exposure Control 

Effectiveness and section 2.6 for RETS/ODCM Radiological Effluent 

Occurrences. 

3. Review the materials in NUREG/CR-6204 and NURG/CR-5569 

identified in the References section.

4. Review the discussion on “accessible” areas and adequate controls 

for High and Very High Radiation Areas in Regulatory Guide 8.38.
Issue Date: 02/20/17 77 IMC 1245, Appendix C3

5. Meet with your supervisor or a qualified inspector to discuss any 

questions you may have as a result of this activity. Discuss the 

answers to the questions listed under the Evaluation Criteria section 

of this study guide with your supervisor.

DOCUMENTATION: Document completion on the Health Physics Inspector Technical 

Proficiency Level Signature Card and Certification Item ISA-HP-16.
Issue Date: 02/20/17 78 IMC 1245, Appendix C3

Health Physics Inspector On-the-Job Activities

Health Physics Inspector On-the-Job Activity

TOPIC: (OJT-HP-1) Radiological Hazard Assessment and Exposure Controls

PURPOSE: The purpose of this activity is to familiarize you with inspection activities in 

inspection procedure 71124.01, “Radiological Hazard Assessment and 

Exposure Controls.” The objectives of this procedure are:

1. To review and assess licensee performance in assessing the 

radiological hazards in the workplace associated with licensed 

activities and the implementation of appropriate radiation monitoring 

and exposure control measures for both individual and collective 

exposures.

2. To verify that the licensee is properly identifying and reporting PIs 

for the Occupational Radiation Safety Cornerstone.

3. To identify those performance deficiencies that were reportable as a 

PI and which may have represented a substantial potential for 

overexposure of the worker.

Upon completion of this on-the-job activity, you will be able to perform 

assessments of reactor licensee’s radiological hazard assessment and 

exposure controls.

COMPETENCY

AREA: TECHNICAL AREA EXPERTISE

ASSESSMENT AND ENFORCEMENT

COMMUNICATION

INSPECTION

LEVEL 

OF EFFORT: 64 hours

REFERENCES: 1. Licensee “Radiation Protection Plan”

2. Licensee “Radiological Access Control” procedure

3. Licensee “Radiological Postings” procedure

4. Licensee “Radiation Work Permit” procedure

5. Licensee Procedures for High and Very High Radiation Areas

6. Licensee Procedures for Controlling High Risk Jobs and Plant 

Evolutions

7. Licensing Procedures for Release of Materials for Unrestricted Use
Issue Date: 02/20/17 79 IMC 1245, Appendix C3

8. Licensee “Radiological Survey and Analysis” procedure

9. Licensee “Technical Specifications” (Administrative controls for high 

radiation areas)

10. Licensee “Final Safety Analysis Report” or “Updated Safety Analysis 

Report”

11. Davis-Besse Special Inspection Reports 02-06 and 02-16 (ADAMS 

Accession Number ML030070606)

12. Licensee PIs for the Occupational Radiation Safety Cornerstone

EVALUATION

CRITERIA: Upon completion of the tasks in this OJT, the inspector should be able to:

1. Describe areas that are considered risk significant and what kind of 

work would increase the risk significance.

2. Describe licensee controls for radiation areas, high radiation and 

very high radiation areas, and airborne radioactivity areas.

3. Describe the licensee’s mechanism for making timely changes to 

controls and postings for radiation and high/very high radiation 

areas, as a result of changing plant conditions.

4. Describe the results of the survey you performed and the survey 

performed by the licensee and explain what could cause the results 

to be different.

5. Describe Technical Specification required controls for High 

Radiation Area and Very High Radiation Areas and how they 

compare to those in 10 CFR 20.

6. Describe the information communicated to the worker by the RWP, 

and the responsibilities of the radiation workers and HP staff.

7. Explain licensee policy on how electronic dosimeter alarm set points 

are established. 

8. Explain the actions expected from workers in response to a dose or 

dose rate alarm and those expected in the event of an instrument 

malfunction.

9. Describe methods that could be employed to reduce or prevent the

uptake of radioactive materials.

10. Describe expected licensee response to suspected or actual uptake 

of airborne radioactive material.
Issue Date: 02/20/17 80 IMC 1245, Appendix C3

11. Describe controls used to protect against accidentally exposing 

highly activated materials stored in the spent fuel pool, transfer 

canal, moisture separator pit or other tank of water for the purpose 

of shielding.

12. Discuss how the licensee performs radiation surveys on coolant 

piping and components, coolant radiochemistry analysis and waste 

radiochemistry analysis to determine changes in radionuclide 

source term and radionuclide mix in various waste streams.

13. Describe the degree of documentation in the corrective action 

program with regards to threshold, detail, thoroughness, and 

timeliness and how the program should be assessed for adequacy.

14. Identify those performance deficiencies that were reportable as a PI 

and which may have represented a substantial potential for 

overexposure of the worker.

15. Describe the licensee controls to ensure that radioactive materials 

are not released to unrestricted areas.

16. Describe the training and qualification requirements for the RP 

technicians and radiation workers.

TASKS: 1. Identify exposure significant work areas within radiation areas, high 

radiation areas and airborne radioactivity areas in the plant.

2. Identify appropriate licensee controls for exposure significant work 

areas.

3. Walk down a work area with a survey instrument and determine if 

licensee surveys are accurate and postings are complete. Prior to 

performing the survey perform the following steps:

a. Select an appropriate instrument. (If you are uncertain contact 

the qualified inspector) 

b. Familiarize yourself with the operations of the instrument and 

any peculiarities/features that it may have. 

c. Turn the instrument on and check the batteries, check the high 

voltage or instrument zero if applicable, check the physical 

condition and perform a response test with an appropriate 

source.

d. Verify that the instrument is within its calibration interval. If 

there is any doubt about the operability of the instrument, do 

not use it.

4. Review plant specific Technical Specifications high radiation area 

requirements to determine necessary barriers for high radiation 
Issue Date: 02/20/17 81 IMC 1245, Appendix C3

areas, and locked high radiation areas. Review licensee 

procedures for very high radiation areas, and identify the “additional 

measures” implemented, as required by 10 CFR 20.1602.

5. Review radiation work permits (RWP) used to access an exposure 

significant work area and identify what work control instructions or 

control barriers are specified.

6. Review licensee procedures which define how electronic dosimeter 

alarm set points are established.

7. Review plant policy and determine the expected actions on an 

electronic dosimeter integrated dose alarm, dose rate alarm and a 

malfunction.

8. In a power plant setting discuss the engineering controls that would 

be expected for an area that was expected to exceed 20 DAC 

(Derived Air Concentration) airborne being produced by surfacedisturbing work. (A qualified inspector can identify an example 

physical setting and work through this with you.)

9. Whenever possible, select several work activities and follow these 

jobs (in radiation and high radiation areas) during the inspection. 

Observe workers during the pre-job briefings, work preparation 

activities, actual in-field work and any post-work debriefings. 

Discuss any observations with a qualified inspector.

10. Perform an inspection of the licensee spent fuel pool/transfer canal 

or moisture separator pit and determine licensee controls for high 

activity items stored in the pool. Determine what controls are in 

place to prevent accidentally exposing high activity items. Note that 

some licensees store high activity items suspended from the side of 

spent fuel pools in which case determine the controls that they 

would use for short term transient storage prior to disposition.

11. Review a licensee’s self-assessment or audit. Select 3 findings 

from the report and determine if the identified problems are 

documented in the corrective action program for tracking and 

resolution.

12. Observe control point operation to verify that personnel (radiation 

workers and RP technicians) are complying with the applicable 

guidance and training regarding survey and release of material.

13. Select three sources from the licensee inventory records. Verify the 

source location is correct and inspect the physical security of the 

source. Review leak test results, if applicable. Verify that any 

transactions involving Category 1 and 2 sources were reported in 

accordance with 10 CFR 20.2207.
Issue Date: 02/20/17 82 IMC 1245, Appendix C3

14. Review the licensee’s reports for dose rate alarms from electronic 

dosimeters in which the measured dose rate was greater than 1000 

mrem/hr. Review the licensee’s reports for dose alarms from

electronic dosimeters in which the dose was greater than 100 mrem

above the alarm dose limit. Explain how this data can be used to 

verify the accuracy of the reported PIs.

15. Meet with your supervisor or a qualified inspector to discuss any 

questions you may have as a result of this activity and to 

demonstrate that you can meet the evaluation criteria for this 

activity.

DOCUMENTATION: Document completion on the Health Physics Inspector Technical 

Proficiency Level Signature Card and Certification Item OJT-HP-1.
Issue Date: 02/20/17 83 IMC 1245, Appendix C3

Health Physics Inspector On-the-Job Activity

TOPIC: (OJT-HP-2) Occupational ALARA Planning and Controls

PURPOSE: The purpose of this activity is to familiarize you with the inspection 

activities in inspection procedure 71124.02, “Occupational ALARA 

Planning and Controls.” The objective of this procedure is to assess 

performance with respect to maintaining individual and collective radiation 

exposures ALARA. This inspection will determine whether the licensee’s

ALARA program, including administrative, operational, and engineering 

controls, is effectively maintaining occupational exposure ALARA.

Upon completion of this on-the-job activity, you will be able to perform 

assessments of reactor licensee’s ALARA planning and controls.

COMPETENCY

AREA: TECHNICAL AREA EXPERTISE

ASSESSMENT AND ENFORCEMENT

COMMUNICATION

INSPECTION

LEVEL 

OF EFFORT: 54 hours

REFERENCES: 1. Licensee ALARA Program procedure

2. Licensee dose printout of individual doses for a work group 

identified by qualified inspector

3. Licensee source term trending documentation/ ALARA review

4. Completed licensee ALARA package for a significant job identified 

by qualified inspector

5. Licensee historical or current outage backbone schedule that 

identifies significant work milestones

6. Licensee temporary shielding procedure

7. Licensee hot spot reduction procedure

8. Licensee source term control/ reduction procedure (May be 

fragmented into several procedures)

9. Licensee “Strategic Primary Water Plan” section describing 

shutdown chemistry or other procedure describing the chemistry 

controls involved in shutting down the plant. (The “Strategic Water 

Plans” are licensee process optimization documents that are 

generated in response to EPRI and NEI initiatives/commitments)
Issue Date: 02/20/17 84 IMC 1245, Appendix C3

EVALUATION

CRITERIA: Upon completion of the tasks in this OJT, the inspector should be able to:

1. Explain why engineering controls are used to reduce or eliminate 

respiratory protection requirements.

2. Explain why there are still some variations in the amount of dose 

received for a population of workers even though skill levels are the

same and work assignments are comparable. 

3. Describe licensee source term trending techniques. 

4. Describe key considerations and components of an ALARA work 

package.

5. Describe how RWPs capture the ALARA planning considerations.

6. Describe considerations in determining whether temporary shielding 

is provided at RP’s request.

7. Describe how the timing of work evolutions affects dose reduction 

opportunities. Discuss the things that may make the timing less 

than optimal.

8. Describe how the licensee ALARA program incorporates lessons 

learned, industry experience, historical data, and employee 

feedback/ recommendations. 

TASKS: 1. Estimate dose reductions achieved by use of engineering controls.

a. Reduction in TEDE (Total Effective Dose Equivalent) dose by 

avoiding having respiratory protection.

b. Reduction in TEDE dose that would have been incurred had 

the licensee used neither engineering controls nor respiratory

protection.

2. Compare individual exposures for a selected workgroup and identify 

probable reasons for significant differences. Population selected 

should be comparable in skill level and duty assignment (i.e. do not 

compare a Senior RP Technician with a Junior RP Technician as 

duties are dissimilar.)

3. Determine source term historical trends and current status using 

licensee records. (Source terms include hot spots, pipe contact 

dose rates, coolant activity, contamination levels and airborne 

activity.) 

4. Determine the elements of the licensee source term control 

strategies. Determine constraints on these elements, (i.e., hotspot 

reduction and shielding may not be dose-justified (more dose 
Issue Date: 02/20/17 85 IMC 1245, Appendix C3

received during mitigation than can reasonably be saved which can 

occur in some infrequently entered areas such as steam generator 

cubicles. Zinc injection, certain chemistry controls and alternative 

resins may be rejected based on fuel warranty or cost concerns or 

both.)

5. Review a completed licensee ALARA job package. Compare 

planned work (man-hours) to actual work (man-hours). Compare 

expected dose (mrem) with actual dose (mrem). Compare expected 

radiological conditions with actual for dose rate, contamination and 

airborne radioactivity. Identify any additional controls employed to 

reduce exposure such as ventilation, keeping work surfaces wet, 

glove bags, and shielding. Identify any in-progress reviews that are 

used to reevaluate expected man-hours and dose. If included, 

review surveys used for original estimate and compare to surveys 

performed while work was in progress.

6. Review the ALARA package to determine how well the RWP 

captured the ALARA requirements as well as any other work 

documentation. Review an example of the licensee’s ALARA Workin-Progress Review for an outage high dose job. If the job was 

expected to exceed its dose estimate, what was the estimate for the 

increased dose and verify that this increase in the estimated dose 

was approved by the Station ALARA Committee.

7. Review the RP Group shielding request with respect to dose rate 

reduction and assigning value. Compare the projected dose 

savings with the dose expended installing shielding. If net dose 

savings did not occur determine if radiological risk was averted such 

as elimination of high dose gradients.

8. Whenever possible, select several work activities and follow these 

jobs (in radiation and high radiation areas) during the inspection. 

Observe workers during the pre-job ALARA briefings, work 

preparation activities, actual in-field work and any post-work lessonlearned debriefings. Discuss any observations with a qualified 

inspector.

9. Review outage schedule to determine if jobs are scheduled to 

reduce exposure (e.g., Major work on reactor coolant pumps and 

steam generator platforms is scheduled after initial shutdown crud 

burst or is it done during the crud burst? Are systems worked on 

when they are filled with water if possible? How is shielding and 

scaffold erection scheduled in relation to other work?).

10. Review the most recent Outage ALARA Report. Review the 

recommendations and lessons learned from the report. Spot check 

three of the most significant from a dose and cost reduction 

perspective to see how the licensee plans to incorporate them into 

future work.
Issue Date: 02/20/17 86 IMC 1245, Appendix C3

11. Meet with your supervisor or a qualified health physics inspector to 

discuss any questions you may have as a result of this activity and 

demonstrate that you can meet the evaluation criteria for this 

activity.

DOCUMENTATION: Document completion on the Health Physics Inspector Technical 

Proficiency Level Signature Card and Certification Item OJT-HP-2.
Issue Date: 02/20/17 87 IMC 1245, Appendix C3

Health Physics Inspector On-the-Job Activity

TOPIC: (OJT-HP-3) In-Plant Airborne Radioactivity Control and Mitigation

PURPOSE: The purpose of this activity is to familiarize you with the inspection 

activities in inspection procedure 71124.03, “In-Plant Airborne 

Radioactivity Control and Mitigation.” The objectives of this procedure are 

to verify that in-plant airborne concentrations are being controlled 

consistent with ALARA to the extent necessary to validate plant operations 

as reported by the PI and to verify that the practices and use of respiratory 

protection devices on site do not pose an undue risk to the wearer.

Upon completion of this on-the-job activity, you will be able to perform 

assessments of reactor licensee’s airborne radioactivity controls.

COMPETENCY

AREA: TECHNICAL AREA EXPERTISE

INSPECTION

LEVEL 

OF EFFORT: 20 hours

REFERENCES: 1. Licensee Respiratory Protection Program procedures

2. Licensee Respiratory Protection Equipment training and 

maintenance procedures

3. Licensee procedures for use of temporary containment/confinement 

and ventilation to control the spread of airborne radioactive 

contamination

4. Licensee procedures for use of systems to continuously monitor air 

for changing airborne concentrations in the plant and applicable 

alarm set points

EVALUATION

CRITERIA: Upon completion of this activity, the inspector should be able to:

1. Describe how the licensee uses engineering controls to minimize 

airborne radioactivity in the workplace.

2. Describe how the licensee uses systems to monitor the air for the 

presence of airborne radioactivity.

3. Describe the potential consequences of the licensee;

a. Providing non-NIOSH approved respirators to workers;

b. Not ensuring that Grade D air is provided to supplied air 

respiratory protective devices;
Issue Date: 02/20/17 88 IMC 1245, Appendix C3

c. Not providing a standby rescue person while using devices 

that are difficult to escape;

d. Not properly determining if the area is considered Immediately 

Deleterious to Life and Health (IDLH).

4. Describe the potential consequences of a defective SCBA being 

used in a fire, a toxic atmosphere or an area with unknown high 

airborne radioactivity. Identify the section in 10 CFR 20 that 

requires that respirators be NIOSH approved.

5. Describe the medical and fit test qualification and training program 

for respirator users.

6. Describe the licensee SCBA air cylinder program for air quality 

testing, hydro-testing, and cylinder replacement schedule.

TASKS: 1. Review licensee’s process and procedures for providing engineering 

controls (i.e., temporary tents, or other containment devices, and/or 

filtered ventilation) to prevent the generation and spread of airborne 

radioactive contamination resulting from onsite licensed activities.

2. Verify that licensees use systems to monitor and warn of changing 

airborne concentrations in the plant and that they have established 

appropriate alarm set points.

3. Verify that respiratory protective devices employed by the licensee 

have been certified by NIOSH or have been specifically approved 

for use by the NRC per the requirements of 10 CFR 20.1703.

4. Verify that air used for respiratory protection devices meets or 

exceeds Grade D air quality.

5. Observe condition of stored respiratory protection devices ready for 

use. 

6. Based on FSAR, Technical Specifications and Emergency 

Operating Procedures requirements, review the status and 

surveillance records of SCBA staged and ready for use in the plant.

7. Determine if 1) control room operators and other emergency 

response personnel (assigned in-plant search and rescue duties or 

as required by EOPs or Emergency Plan) are trained and qualified 

in the use of SCBA (including personal bottle change out) and 2) 

these individuals are provided with appropriate vision correction 

lenses.

8. Review pertinent sections of 29 CFR 1910 and 49 CFR Part 173 

and Part 178 for additional respirator and cylinder maintenance.
Issue Date: 02/20/17 89 IMC 1245, Appendix C3

9. Discuss each of the program elements of an acceptable Respiratory 

Protection Program described in RG 8.15, “Acceptable Programs for 

Respiratory Protection, Rev 1.

10. Observe the licensee inspecting a SCBA, testing the regulator and 

testing the various alarms. Also, if possible observe a worker 

getting a respirator fit test.

11. Meet with your supervisor or a qualified Health Physics inspector to 

discuss any questions you may have as a result of this activity and 

to demonstrate that you can meet the evaluation criteria for this 

activity.

DOCUMENTATION: Document completion on the Health Physics Inspector Technical 

Proficiency Level Signature Card and Certification Item OJT-HP-3.
Issue Date: 02/20/17 90 IMC 1245, Appendix C3

Health Physics Inspector On-the-Job Activity

TOPIC: (OJT-HP-4) Occupational Dose Assessment

PURPOSE: The purpose of this activity is to familiarize you with the inspection 

activities in inspection procedure 71124.04, “Occupational Dose 

Assessment.” The objective of this procedure is to determine the accuracy 

and operability of personal monitoring equipment, determine the accuracy 

and effectiveness of the licensee’s methods for determining total effective 

dose equivalent and verify that occupational dose is appropriately 

monitored.

Upon completion of this on-the-job activity, you will be able to perform 

assessments of reactor licensee’s ability to measure occupational dose.

COMPETENCY

AREA: TECHNICAL AREA EXPERTISE

ASSESSMENT AND ENFORCEMENT

COMMUNICATION

INSPECTION

LEVEL

OF EFFORT: 20 hours 

REFERENCES: 1. Licensee Declared Pregnant Worker Procedure

2. Licensee procedures that address methods for determining if an 

individual is internally or externally contaminated

3. Licensee Internal Dose Assessment procedures

4. Licensee Monitoring of External Dose Procedures

5. Occupational Dosimeter Performance Test and Quality Assurance

6. National Voluntary Laboratory Accreditation (NVLAP) Accreditation 

Certificates for Applicable Radiation Categories

EVALUATION

CRITERIA: Upon completion of the tasks in this OJT, the inspector should be able to:

1. Discuss licensee documentation that is used to identify, report and 

track problems involving personnel contaminations, unexpected 

exposures, radiological incidents and events involving uncontrolled 

or unmonitored internal or external exposure of workers.

2. If the plants have individuals with internal exposures greater than 10 

mrem Committed Effective Dose Equivalent (CEDE), describe the 

process used to properly assess the exposure. Otherwise, describe 

the procedural process that would be used if such an exposure were 

to be unexpectedly discovered.
Issue Date: 02/20/17 91 IMC 1245, Appendix C3

3. Describe the licensee’s methodology for monitoring external dose 

situations in which non-uniform fields are expected.

4. Describe the licensee’s methodologies for accounting for neutron 

doses.

5. Describe the licensee’s methodology for assessing skin doses 

resulting from skin contamination.

6. Describe the process used to determine and assign fetal dose. 

Describe a woman’s rights with respect to declaring her pregnancy.

TASKS: 1. Review licensee self-assessments, audits, and Licensee Event 

Reports and focus on radiological incidents that involved personnel 

contamination monitor alarms due to personnel internal exposures

and dose alarms from electronic dosimeters.

2. For any internal exposure > 100 mrem CEDE, determine whether 

the intake was properly evaluated using calibrated equipment and 

the internal exposure was properly assessed in accordance with 

licensee procedure. Determine how the licensee includes hard-todetect radionuclides (e.g., scaling factors) into the internal dose 

assessments.

3. Verify that the licensee’s personnel dosimeters that require 

processing are processed by a NVLAP accredited processor that is 

certified in the applicable Radiation Categories.

4. Review Declared Pregnant Worker (or Woman) procedures. 

Review the exposure results and monitoring controls employed by

the licensee with respect to the requirements. (Program should 

have provisions for determining fetal dose from internal deposition 

as well as external radiation.)

5. Review the licensee’s procedure for monitoring Effective Dose 

Equivalent for workers in radiation fields with large dose gradients. 

Determine if during the most recent outage, the multi-badging 

program and practices meet the guidance in RG 8.40 and HPS

N13.41-1997.

6. Meet with your supervisor or a qualified health physics inspector to 

discuss any questions you may have as a result of this activity and 

demonstrate that you can meet the evaluation criteria for this 

activity.

DOCUMENTATION: Document completion on the Health Physics Inspector Technical 

Proficiency Level Signature Card and Certification Item OJT-HP-4.
Issue Date: 02/20/17 92 IMC 1245, Appendix C3

Health Physics Inspector On-the-Job Activity

TOPIC: (OJT-HP-5) Radiation Monitoring Instrumentation

PURPOSE: The purpose of this activity is to familiarize you with the inspection 

activities in inspection procedure 71124.05, “Radiation Monitoring 

Instrumentation.” The objective of this procedure is to verify that the 

licensee is ensuring the accuracy and operability of radiation monitoring 

instruments that are used to monitor areas, materials, and workers to 

ensure a radiologically safe work environment. The instrumentation 

subject to this review includes equipment used to monitor radiological 

conditions incident to normal plant operations, including anticipated 

operational occurrences, and conditions resulting from postulated 

accidents.

Upon completion of this on-the-job activity, you will be able to perform 

assessments of reactor licensee’s radiation monitoring instrumentation

activities.

COMPETENCY

AREA: TECHNICAL AREA EXPERTISE

ASSESSMENT AND ENFORCEMENT

COMMUNICATION

INSPECTION

LEVEL 

OF EFFORT: 40 hours

REFERENCES: 1. Licensee fixed and portable instrumentation procedures

2. Licensee air sampling equipment procedures

3. Licensee post-accident monitoring equipment procedures

4. Licensee counting room instrumentation procedures

5. Licensee personnel contamination monitor and small article monitor 

procedures.

6. Licensee whole body equipment procedure

EVALUATION

CRITERIA: Upon completion of the tasks in this OJT, the inspector should be able to:

1. Describe the extent that emergency planning and maintenance rule 

inspections (in your region) address radiological instruments well 

enough to preclude redundant inspection activities while ensuring 

adequate inspection coverage.
Issue Date: 02/20/17 93 IMC 1245, Appendix C3

2. Describe the plant activities and areas that have permanently 

installed area radiation monitors based on examination of the plant’s 

Final Safety Analysis Report (FSAR).

3. Discuss data analysis techniques, including MDA and LLD. Verify 

licensee MDA and LLD calculations for each release pathway.

4. Describe the instrumentation used for high risk jobs including survey 

meter types and designations, remote readout area monitors and 

continuous air monitors.

5. Describe the fixed and portable instrumentation used for personnel 

release from radiologically controlled area and how it is used.

6. Discuss the calibration technique for radiation instruments used to 

screen and clear potentially contaminated materials/articles (e.g., a 

small article monitor (SAM) and a bag monitor) based on IE Circular 

81-07 and Information Notice 85-92.

7. Describe the methodology used to calibrate instruments, verify their 

operability, determine if appropriate calibration sources are used,

and determine, when applicable, if alarm set points have been 

properly set. Describe the channel operability testing frequency.

8. Describe the process used to resolve situations where instruments 

are found to be significantly out of calibrations. State the impact on 

surveys done with the instrument?

9. Describe the likely consequences of having a substantially over or 

under responding instrument used to cover work and perform 

surveys. State the effect of significantly extending the source check 

frequencies of survey instruments.

TASKS: 1. Determine what radiological instrumentation is inspected in your 

region under NRC emergency planning or maintenance rule 

inspections if any. (This will require discussion with Maintenance 

Rule and Emergency Planning Inspectors normally prior to leaving 

the licensee site.)

2. Review the plant FSAR to identify applicable radiation monitors 

associated with transient high and very high radiation areas 

including those used in remote emergency assessment.

3. Identify the types of portable radiation detection instrumentation 

used for job coverage of high radiation area work, other temporary 

area radiation monitors currently used in the plant, and continuous 

air monitors.

4. Identify types of radiation detection instruments utilized for 

unconditional release of personnel and equipment from the 

radiologically controlled area.
Issue Date: 02/20/17 94 IMC 1245, Appendix C3

5. Verify calibration, operability, frequency and alarm set point (if 

applicable) of several types of instruments and equipment.

6. Determine what actions are taken when, during calibration or source 

checks, an instrument is found significantly out of calibration 

(>50%).

7. Determine possible consequences of instrument use since last 

successful calibration or source check for those out of calibration or 

that have failed a source check.

8. Review and understand the calibration techniques and results of the 

measurements made with counting room instruments.

9. Review and understand the measurement laboratory’s analytical 

techniques (e.g., gamma, Sr-89/90, H-3, and gross alpha/beta).

10. Review and evaluate the calibration of the Whole Body Counter.

11. Review and evaluate the calibration of Post-Accident Monitoring 

Instrumentation.

12. Meet with your supervisor or a qualified health physics inspector to 

discuss any questions you may have as a result of this activity and 

demonstrate that you can meet the evaluation criteria for this 

activity.

DOCUMENTATION: Document completion on the Health Physics Inspector Technical 

Proficiency Level Signature Card and Certification Item OJT-HP-5.
Issue Date: 02/20/17 95 IMC 1245, Appendix C3

Health Physics Inspector On-the-Job Activity

TOPIC: (OJT-HP-6) Radioactive Gaseous and Liquid Effluent Treatment 

PURPOSE: The purpose of this on-the-job activity is to provide tasks and information 

that will familiarize you with the inspection requirements contained in 

procedure IP 71124.06, “Radioactive Gaseous and Liquid Effluent 

Treatment,” to allow you to independently conduct inspections in this area. 

The objectives of this procedure are:

1. To ensure that the gaseous and liquid effluent processing systems 

are maintained so that radiological discharges are properly 

mitigated, monitored, and evaluated with regard to public exposure. 

The principal design Criterion for these systems are found in 

General Design Criteria 60 and 64 of Appendix A to 10 CFR Part 

50, Radiological Effluent Technical Specifications (RETS). Generic 

Letter 89-01, NUREG-1301, NUREG-1302, Standard Radiological 

Effluent Controls (SREC), and the Offsite Dose Calculation Manual 

(ODCM) provide acceptable methods of implementation to meet 

these design criteria.

2. To ensure that abnormal radioactive gaseous or liquid discharges 

and conditions, when effluent radiation monitors are out-of-service, 

are controlled in accordance with applicable regulatory requirements 

and licensee procedures.

3. To verify that the licensees’ quality control program ensures that the 

radioactive effluent sampling and analysis requirements are 

satisfied so that discharges of radioactive materials are adequately 

quantified and evaluated.

4. To verify the adequacy of public dose calculations and projections 

resulting from radioactive effluent discharges.

This inspection activity verifies aspects of the Public Radiation Safety 

cornerstone not measured by performance indicators. In Public Radiation 

Safety, the effluent release occurrence performance indicator measures 

radioactive gaseous and liquid releases that were above Technical 

Specification and/or Offsite Dose Calculation Manual limits. Radiation 

exposure to the public is to be below the 10 CFR Part 20 and 40 CFR Part 

190 limits. Doses below the design objectives of Appendix I to 10 CFR 

Part 50 and 40 CFR Part 190 dose values are considered ALARA. 

Radioactive effluent treatment systems and monitors are required by 

Criteria 60 and 64 of Appendix A to 10 CFR Part 50. Proper operation of 

the system and monitors, as described in the licensee’s Radioactive 

Effluent Control Program, will ensure an adequate defense-in-depth 

against an unmonitored, unanticipated, and unplanned release of 

radioactive material to the environment.
Issue Date: 02/20/17 96 IMC 1245, Appendix C3

Upon completion of this on-the-job activity, you will be able to conduct 

assessments of a reactor licensee’s radioactive gaseous and liquid effluent 

treatment and monitoring systems.

COMPETENCY

AREA: INSPECTION

LEVEL 

OF EFFORT: 90 HOURS

REFERENCES: 1. Plant Updated Final Safety Analysis Report

2. Licensee Procedures for Effluent Monitor Calibration and Effluent 

Sampling

3. Licensee Offsite Dose Calculation Manual

4. Licensee Radiological Effluent Tech Specs (RETS)

5. Most recent copy of the Licensee’s Annual Radioactive Effluent 

Release Report

6. Licensee PIs for the Public Radiation Safety Cornerstone

7. Licensee Radiation Monitor System Health Report and Maintenance 

Rule 10 CFR 50.65 (a)1, Action Plans for Effluent and Accident 

Monitors

EVALUATION

CRITERIA: Upon completion of the tasks, the inspector should be able to:

1. Discuss the effluent sampling techniques (water, air iodine, and air 

particulate) that you inspected. Compare and contrast the 

capabilities and limitations of each method. 

2. Discuss the Laboratory’s QA Policy and QC implementation, 

including inter-laboratory and intra-laboratory comparison. 

Compare it with NRC expectations. 

3. Discuss the radioactive liquid treatment systems and the effluent 

release pathways. Compare these to those presented in the 

UFSAR and ODCM.

4. Discuss the radioactive gas treatment systems and the gaseous 

effluent release pathways. Compare these to those presented in 

the UFSAR and ODCM.

5. Discuss the radioactive liquid and gaseous effluent radiation 

monitoring systems listed in the ODCM. Discuss the capabilities of 

the monitors and their conformance with UFSAR and ODCM 

descriptions.
Issue Date: 02/20/17 97 IMC 1245, Appendix C3

6. Discuss the calibration techniques for the effluent radiation 

monitoring systems, including energy responses to various 

detectors, primary and secondary calibrations, and establishing the 

operating high voltage.

7. Discuss the air cleaning systems, their functions and required 

periodic testing (e.g., Augmented Off-Gas, Reactor Building Air 

Cleaning System.).

8. Discuss the calibration techniques for the effluent flow rate 

measurement devices listed in the ODCM. Identify what type of flow 

the sampler is seeing and if there is isokinetic sampling and how to 

determine whether or not there is isokinetic sampling.

TASKS: 1. Review and become familiar with the documents listed in the 

reference list.

2. Locate a copy of the air cleaning system section from the UFSAR 

and Technical Specification. Conduct walk downs of as much of the 

system as possible. Compare your observations with descriptions 

in the UFSAR.

3. Identify the various sources of liquid and gaseous radioactive waste, 

waste streams, and technologies associated with liquid and 

gaseous radioactive waste processing for the facility. Compare your 

observations with descriptions in the UFSAR.

4. Review the licensee’s ODCM and identify the radioactive gaseous 

and liquid effluent pathways. Conduct walk downs of gaseous and 

liquid monitoring systems and look for deficiencies similar to those 

discussed in IN 13-13 and IN 13-12. 

5. Review and evaluate the projected public dose calculation 

methodologies (all pathways) listed in Regulatory Guide 1.109. 

6. Review and evaluate the air cleaning system surveillance test 

results. 

7. Review and evaluate the effluent radiation monitoring system 

channel calibration, functional test, and source check results.

8. Review and evaluate the implementation of the QA/QC in the 

measurement laboratory, including inter-laboratory and intralaboratory comparisons.

9. Review and understand the radioactive liquid and gaseous release 

permits.

10. Review and understand the Annual Radioactive Effluent Release 

Report.
Issue Date: 02/20/17 98 IMC 1245, Appendix C3

11. Review the Tech Specs and ODCM requirements for the ISFSI and 

High Level Waste Facility if applicable. Perform a walk down of 

these facilities and observe how these requirements are being met.

12. Review the licensee’s ODCM and determine for each of the 

Gaseous and Liquid Effluent Monitors, what the Limiting Conditions 

for Operations, Surveillance Requirements and Actions are. Review 

the licensee’s procedures for sampling the various effluent release 

points.

13. Review corrective action reports in the area of radioactive gaseous 

and liquid effluent and monitoring systems to identify problems in 

this area and to understand the licensee’s corrective action process.

14. Meet with your supervisor or a qualified Health Physics inspector to 

discuss any questions you may have as a result of this activity and 

to demonstrate that you can meet the evaluation criteria for this 

activity.

DOCUMENTATION: Document completion on the Health Physics Inspector Technical 

Proficiency Level Signature Card and Certification Item OJT-HP-6.
Issue Date: 02/20/17 99 IMC 1245, Appendix C3

Health Physics Inspector On-the-Job Activity

TOPIC: (OJT-HP-7) Radiological Environmental Monitoring Program 

PURPOSE: The purpose of this on-the job-activity is to provide you with tasks and 

information that will familiarize you with the inspection requirements 

contained in procedure IP71124.07, “Radiological Environmental 

Monitoring Program,” to allow you to independently conduct inspections in 

these areas. The objectives of the procedure are:

1. To verify that the REMP quantifies the impact of radioactive effluent 

releases to the environment and sufficiently validates the integrity of 

the radioactive gaseous and liquid effluent release monitoring 

program.

2. To verify that the REMP is implemented consistently with the 

licensee’s TS and/or ODCM and to validate that the radioactive 

effluent release program meets the design objective in Appendix I to 

10 CFR Part 50.

3. To ensure that the REMP (1) monitors non effluent exposure 

pathways (e.g., onsite spills or leaks, exposures from direct and 

scattered (sky shine) radiation from plant facilities and components), 

(2) is based on sound principles and assumptions, and (3) validates 

that doses to members of the public are within the dose limits of 10 

CFR 20, “Standards for Protection against Radiation,” and 40 CFR 

Part 190, “Environmental Radiation Protection Standards for 

Nuclear Power Operations,” as applicable.

4. To verify that the licensee is continuing to implement the voluntary 

NEI/Industry Groundwater Protection Initiative (GPI).

The NRC requires that licensees ensure adequate protection of public 

health and safety from exposure to radioactive material released to the 

public domain as a result of routine operations. The REMP is required by 

Criterion 64 of Appendix A to 10 CFR Part 50 and supplements the effluent 

monitoring program by verifying that the measurable concentrations of 

radioactive materials and levels of radiation in the environment are in 

agreement with the values predicted by the radioactive effluent monitoring 

program. The licensee is required to implement the REMP in accordance 

with its Technical Specifications and/or Offsite Dose Calculation Manual, 

which are based on the design objectives contained in Appendix I of 10 

CFR Part 50, as required by 10 FR 50.34a. 

Upon completion of this on-the-job activity, you will be able to perform 

assessments of reactor licensee’s radiological environmental monitoring 

activities.

COMPETENCY

AREA: INSPECTION
Issue Date: 02/20/17 100 IMC 1245, Appendix C3

LEVEL 

OF EFFORT: 37 HOURS

REFERENCES: 1. Plant Updated Final Safety Analysis Report

2. Licensee’s Land Use Census

3. Licensee or vendor procedures for sampling and analysis of 

environmental samples

4. Licensee Meteorological Tower procedures

5. Most recent copy of licensee’s Annual Radiological Environmental 

Operating Report

6. Licensee’s Offsite Dose Calculation Manual

EVALUATION

CRITERIA: Upon completion of the tasks, the inspector should be able to:

1. Discuss the environmental sampling techniques (water, milk, air 

iodine, air particulate, vegetation, fish, and soil/sediment) that you 

inspected.

2. Discuss data analysis techniques used by the licensee, including 

MDA and LLD.

3. Discuss the Laboratory’s QA Policy and QC implementation, 

including inter-laboratory and intra-laboratory comparisons.

4. Discuss the Land Use Census. Identify principle land water uses by 

members of the public including any onsite activities. 

5. Discuss the Meteorological Monitoring Program, including 

calibration methodology for wind direction, wind speed, and delta 

temperature.

6. Discuss the licensee’s determination of χ/Q and D/Q, and annual 

average data.

7. Review the REMP Annual Report and discuss the soundness of the 

report.

8. Discuss the licensee’s continued implementation of the Ground 

Water Protection Initiative.

TASKS: 1. Review and become familiar with the documents listed in the 

reference list.
Issue Date: 02/20/17 101 IMC 1245, Appendix C3

2. Locate a copy of the REMP section from the UFSAR and ODCM. 

Identify the location and sampling frequency for each of the required 

environmental sampling media and environmental TLDs.

3. Review the requirements of the REMP, including review of the 

REMP Annual Report.

4. Review the contents of the licensee’s ODCM and compare the 

contents against NUREG 1301/1302 (or Branch Technical Position, 

November 1979).

5. Review the radiological measurement instrument data (proportional 

counter, gamma spectroscopy, liquid scintillation counter, and TLD 

reader).

6. Review the counting statistics used by the licensee, including MDA 

and LLD.

7. Review and evaluate calibration records and maintenance records 

for ambient radiation measurements (using TLDs).

8. Review and evaluate the meteorological monitoring requirements 

and evaluate the adequacy of the calibration results for wind 

direction, wind speed, and delta temperature. Conduct a walk down 

of the primary meteorological tower. Record results of wind 

direction, wind speed and delta temperature. Compare results to

the corresponding control room readings. 

9. Conduct walk downs of as much of the REMP sampling stations as 

possible, including the meteorological monitoring tower and the 

control room as well as on-site monitoring wells and liquid discharge 

composite samplers. Compare the findings with what is described 

in the applicable documents. 

10. Review the sequential relationship between Radiological Effluent 

Controls and the licensee’s REMP.

11. Review the man-made and natural radiation exposure pathways 

(fission/activated products and the source of the natural background 

radiation). Evaluate how the licensee takes these into consideration 

when analyzing its samples.

12. Select three to five structure systems, or components (SSCs) that 

involve or could reasonably involve licensed material for which there 

is a credible mechanism for license material to reach ground water, 

and verify that the licensee has implemented a sampling and 

monitoring program sufficient to detect leakage of these SSCs to 

ground water. 

13. Review the Fundamentals of Laboratory QA/QC Policy and its 

implementation at the licensee’s facility.
Issue Date: 02/20/17 102 IMC 1245, Appendix C3

14. Review corrective action reports in the area of radiological 

environmental monitoring and radioactive material controls to 

identify problems in this area and understand the licensee’s 

corrective action process.

15. Review reported ground water monitoring results, and changes to 

the licensee’s written program for identifying and controlling 

contaminated spills/leaks to ground water.

16. Review identified leakage or spill events and entries made into the 

10 CFR 50.75 (g) records. Review evaluations of leaks or spills, 

and review any remediation actions taken for effectiveness. Review 

onsite contamination events involving contamination of ground 

water. Assess whether the source of the leak or spill was identified 

and mitigated, and verify that applicable notification to offsite 

agencies was made and that a source/evaluation was performed to 

include consideration of hard-to-detect radionuclides.

17. Meet with your supervisor or a qualified Health Physics inspector to 

discuss any questions you may have as a result of this activity and 

to demonstrate that you can meet the evaluation criteria for this 

activity.

DOCUMENTATION: Document completion on the Health Physics Inspector Technical 

Proficiency Level Signature Card and Certification Item OJT-HP-7.
Issue Date: 02/20/17 103 IMC 1245, Appendix C3

Health Physics Inspector On-the-Job Activity

TOPIC: (OJT-HP-8) Radioactive Solid Waste Processing and Radioactive Material 

Handling, Storage, and Transportation

PURPOSE: The purpose of this on-the job activity is to provide you tasks and 

information that will familiarize you with the inspection requirements 

contained in procedure IP 71124.08, “Radioactive Solid Waste Processing 

and Radioactive Material Handling, Storage, and Transportation,” to allow 

you to independently conduct inspections in this area. The objective of 

this procedure is to verify the effectiveness of the licensee’s programs for 

processing, handling, storage, and transportation of radioactive material.

This inspection activity verifies aspects of the Public Radiation Safety 

cornerstone for which there are no performance indicators for unplanned 

public exposure during transportation of radioactive material. The 

licensee’s radioactive material processing and shipping programs are 

required by Criterion 60 of Appendix A to 10 CFR 50, Parts 20, 61, and 71, 

and Department of Transportation regulations contained in 49 CFR Parts 

100-189. 

Upon completion of this on-the job activity, you will be able to perform

assessments of a reactor licensee’s radioactive materials processing, 

storage, handling, and transportation activities.

COMPETENCY 

AREA: INSPECTION

TECHNICAL AREA EXPERTISE

LEVEL

OF EFFORT: 40 HOURS

REFERENCES: 1. Plant Updated Final Safety Analysis Report

2. Plant Annual Radioactive Effluent Release Report

3. Licensee procedures covering all aspects of the radioactive waste 

program

4. Licensee Process Control Program (PCP)

EVALUATION

CRITERIA: Upon completion of the tasks, the inspector should be able to:

1. Discuss the station’s recent radiological effluent release report and 

the types and amounts of radioactive waste that the licensee has 

disposed in the past year. Specifically, you should be able to 

identify the types of waste, quantities, and the principal 

radionuclides contained in the various types of waste. You should 

also identify those types of waste that would require more stringent 
Issue Date: 02/20/17 104 IMC 1245, Appendix C3

packaging and identify where in the ODCM such reports are 

required.

2. Discuss the conformance of the station’s solid and liquid radioactive 

waste processing systems, and their operation, with that described 

in the station’s process control program (PCP) and the updated final

safety analysis report (UFSAR). You should be able to compare 

and contrast the licensee’s processing of its waste with NRC 

positions outlined in branch technical positions and 10 CFR 61.

3. Discuss the licensee’s administrative and physical controls 

established to ensure that non-operational or abandoned radioactive 

waste processing equipment will not contribute to an unmonitored 

release path and/or affect operating systems or be a source of 

unnecessary personnel exposure. You should compare and 

contrast the controls with that identified in NRC guidance in Bulletin 

80-10, and NUREG/CR-5569.

4. Discuss your review of recent changes to radioactive waste 

processing systems and if the changes were in accordance with 

10 CFR 50.59. Discuss the licensee’s estimates of doses to 

members of the public for these changes.

5. Identify and discuss the licensee’s 10 CFR Part 61 waste stream 

analysis program and the various waste streams present and 

compare those identified with that reported by the licensee in its 

annual report. 

6. Discuss how the licensee ensures collection of representative 

samples of waste for the 10 CFR Part 61 program and how it 

monitors for changes in its waste streams for purposes of waste 

characterization and classification. Identify how the licensee 

ensures meeting branch technical positions and is aware of 

guidance in Information Notice 86-20.

7. Show and discuss how the licensee uses the 10 CFR Part 61 waste 

stream analysis data, and any scaling factors and calculations, 1) to 

account for difficult-to-measure radionuclides, and 2) to determine 

curie content for waste to be shipped including concentration 

averaging. Specifically, identify how the licensee quantifies Table 1 

and Table 2 radionuclides in 10 CFR 61.55. 

8. Discuss the packaging, labeling, and marking requirements for 

various types of radioactive materials packages expected to be 

shipped from the facility. Identify in NUREG-1660 and Title 49 the 

specific labeling and placarding requirements. 

9. Discuss the manifesting and placarding requirements for various 

types of radioactive materials packages you inspected. Show in 

NUREG 1660 the specific requirements. Discuss conformance with 

NUREG/BR-0204.
Issue Date: 02/20/17 105 IMC 1245, Appendix C3

10. For the same shipments identified in item 9, identify the allowable 

radiation and contamination dose limits specified in regulatory 

documents 49 CFR100-189, and compare and contrast them with 

the values the licensee identified. 

11. Discuss the Hazmat training and emergency response program 

used by the licensee to meet regulatory requirements. Discuss this 

information relative to guidance in Information Notices 92-62, 92-72, 

95-09, and 49 CFR.

12. Discuss the characterization and classification of non-exempt 

low-level radioactive waste shipments reviewed in accordance with 

10 CFR Part 61. Specifically, identify how the shipments met 

10 CFR 61 and branch technical positions.

13. Discuss how the licensee maintains its regulatory documents, for its 

radioactive material processing and transportation program, up-todate. 

14. Discuss the licensee’s corrective action program in the area of 

radioactive material processing, handling, storage, and shipping. 

Discuss the applicability of 10 CFR 50, Appendix B.

TASKS: 1. Review and familiarize yourself with the documents listed in the 

reference list. Specifically identify the purpose of each document 

and what guidance the document provides. 

2. Locate a copy of the radioactive waste section from the UFSAR. 

Conduct walk downs (considering ALARA and safety constraints) of 

as much of the station’s liquid and solid radioactive waste 

processing systems during your OJT activity. Assess the licensee’s 

conformance of the facility with its design documents including any 

10 CFR 50.59 changes to the systems. 

3. Identify the various sources of liquid and solid radioactive waste, 

waste streams, and technologies associated with liquid radioactive 

waste processing for the facility. Compare them with that reported 

by the licensee in its Part 61 analysis. 

4. Review the contents of the licensee’s process control program 

(PCP), use of scaling factors for hard to detect nuclides, and the 

waste form and characteristics requirements for disposal of solid 

radioactive waste. Identify how the licensee quantifies those 

radionuclides listed in 10 CFR 61 Tables 1 and 2.

5. Review the requirements for the transfer of radioactive material 

contained in licensee procedures and compare that with 

requirements contained in 10 CFR 20 and 10 CFR 71.
Issue Date: 02/20/17 106 IMC 1245, Appendix C3

6. Review the burial license requirements for a low-level radioactive 

waste disposal facility used by the licensee and discuss the specific 

requirements, in the area of waste characteristics, for burial of the 

waste.

7. Review the requirements for receipt of radioactive material by the 

licensee relative to criteria contained in 10 CFR 20 including 

reporting requirements for problems identified.

8. Review the licensee’s implementation of the regulations in the area 

of training and emergency response as specified in 49 CFR 172 

Subparts G and H. 

9. Select at least five non-exempt radioactive materials shipments and 

review for compliance with all appropriate regulatory requirements 

including conformance with the cask certificate of compliance for the 

shipping casks used. 

10. Observe the packaging, surveying, labeling, marking, vehicle 

checks, emergency instruction, disposal manifests, shipping papers, 

and loading of a radioactive waste shipment (a non-exempt 

shipment preferable). Inter-compare your findings with applicable 

regulatory requirements. 

11. Review corrective action reports in the area of radioactive waste 

processing, handling, storage, and transportation to identify 

problems in this area and to understand the licensee’s corrective 

action process.

12. Meet with your supervisor or a qualified Health Physics inspector to 

discuss any questions you may have as a result of this activity and 

to demonstrate that you can meet the evaluation criteria for this 

activity.

DOCUMENTATION: Document completion on the Health Physics Inspector Technical 

Proficiency Level Signature Card and Certification Item OJT-HP-8.
Issue Date: 02/20/17 107 IMC 1245, Appendix C3

Health Physics Inspector Technical Proficiency Level Signature Card and Certification

Inspector Name: ________________________________

Employee

Initials / Date

Supervisor’s 

Signature / 

Date

A. Training Courses

Power Plant Engineering (self-study of selected chapters)

GE BWR Systems Overview (R-104B)

Westinghouse Systems Overview (R-104P)

Advanced Health Physics (H-201)

Respiratory Protection (H-311)

Environmental Monitoring for Radioactivity (H-111)

Radioactive Waste Management (H-202S Self-Study)

Transportation of Radioactive Materials (H-308)

B. Individual Study Activities

(ISA-HP-1) Code of Federal Regulations (CFR)

(ISA-HP-2) Licensee Documents for Health Physics 

Inspectors

(ISA-HP-3) Radiological Hazard Assessment and 

Exposure Controls

(ISA-HP-4) Occupational ALARA Planning and Controls

(ISA-HP-5) In-Plant Airborne Radioactivity Control and 

Mitigation

(ISA-HP-6) Occupational Dose Assessment

(ISA-HP-7) Radiation Monitoring Instrumentation

(ISA-HP-8) Radioactive Gaseous and Liquid Effluent 

Treatment

(ISA-HP-9) Radiological Environmental Monitoring 

Program

(ISA-HP-10) Radioactive Solid Waste Processing and 

Radioactive Material Handling, Storage and 

Transportation

(ISA-HP-11) Significance Determination Process -

Occupational Radiation Safety

(ISA-HP-12) Significance Determination Process - Public 

Radiation Safety: Radioactive Material Control

(ISA-HP-13) Significance Determination Process - Public 

Radiation Safety: Radioactive Effluent Release Program

(ISA-HP-14) Significance Determination Process - Public 

Radiation Safety: Environmental Monitoring Program
Issue Date: 02/20/17 108 IMC 1245, Appendix C3

(ISA-HP-15) Significance Determination Process – Public 

Radiation Safety: Transportation Branch

(ISA-HP-16) Performance Indicator - Occupational 

Radiation Safety

On-the-Job Activities

(OJT-HP-1) Radiological Hazard Assessment and 

Exposure Controls

(OJT-HP-2) Occupational ALARA Planning and Controls

(OJT-HP-3) In-Plant Airborne Radioactivity Control and 

Mitigation

(OJT-HP-4) Occupational Dose Assessment

(OJT-HP-5) Radiation Monitoring Instrumentation

(OJT-HP-6) Radioactive Gaseous and Liquid Effluent 

Treatment

(OJT-HP-7) Radiological Environmental Monitoring 

Program

(OJT-HP-8) Radioactive Solid Waste Processing and 

Radioactive Material Handling, Storage, and 

Transportation

Supervisor’s signature indicates successful completion of all required courses and activities 

listed in this journal and readiness to appear before the Oral Board.

Supervisor’s Signature: ________________________________ Date: _____________

This signature card and certification must be accompanied by Form 1, Health Physics Inspector 

Technical Proficiency Level Equivalency Justification, if applicable.

Copies: Inspector

HR Office

Supervisor
Issue Date: 02/20/17 109 IMC 1245, Appendix C3

Form 1: Health Physics Inspector Technical Proficiency Level Equivalency Justification

Inspector Name: 

_____________________________

Identify equivalent training and experience 

for which the inspector is to be given 

credit.

A. Training Courses

Power Plant Engineering (self-study of selected 

chapters)

GE BWR Systems Overview (R-104B)

Westinghouse Systems Overview (R-104P)

Advanced Health Physics (H-201)

Respiratory Protection (H-311)

Environmental Monitoring for Radioactivity (H111)

Radioactive Waste Management (H-202S selfstudy)

Transportation of Radioactive Materials (H-308)

B. Individual Study Activities

(ISA-HP-1) Code of Federal Regulations (CFR)

(ISA-HP-2) Licensee Documents for Health 

Physics Inspectors

(ISA-HP-3) Radiological Hazard Assessment 

and Exposure Controls
Issue Date: 02/20/17 110 IMC 1245, Appendix C3

(ISA-HP-4) Occupational ALARA Planning and 

Controls

(ISA-HP-5) In-Plant Airborne Radioactivity 

Control and Mitigation

(ISA-HP-6) Occupational Dose Assessment

(ISA-HP-7) Radiation Monitoring and 

Instrumentation

(ISA-HP-8) Radioactive Gaseous and Liquid 

Effluent Treatment

(ISA-HP-9) Radiological Environmental 

Monitoring Program

(ISA-HP-10) Radioactive Solid Waste 

Processing and Radioactive Material Handling, 

Storage and Transportation

(ISA-HP-11) Significance Determination Process

- Occupational Radiation Safety

(ISA-HP-12) Significance Determination Process

- Public Radiation Safety: Radioactive Material 

Control

(ISA-HP-13) Significance Determination Process

- Public Radiation Safety: Radioactive Effluent 

Release Program

(ISA-HP-14) Significance Determination Process

- Public Radiation Safety: Environmental 

Monitoring Program

(ISA-HP-15) Significance Determination Process

- Public Radiation Safety: Transportation Branch

(ISA-HP-16) Performance Indicator -

Occupational Radiation Safety
Issue Date: 02/20/17 111 IMC 1245, Appendix C3

C. On-the-Job Activities

(OJT-HP-1) Radiological Assessment and 

Exposure Controls

(OJT-HP-2) Occupational ALARA Planning and 

Controls

(OJT-HP-3) In-Plant airborne Radioactivity 

Control and Mitigation

(OJT-HP-4) Occupational Dose Assessment

(OJT-HP-5) Radiation Monitoring

Instrumentation

(OJT-HP-6) Radioactive Gaseous and Liquid 

Effluent Treatment

(OJT-HP-7) Radiological Environmental 

Monitoring Program

(OJT-HP-8) Radioactive Solid Waste Processing 

and Material Handling, Storage, and 

Transportation

Supervisor’s Recommendation: Signature / Date: ______________________________

Division Director’s Approval: Signature / Date: ______________________________

Copies to: Inspector 

HR Office

Supervisor
Issue Date: 02/20/17 Att1-1 1245, Appendix C3

Revision History Table

Commitment 

Tracking 

Number

Accession 

Number

Issue Date

Change Notice

Description of Change Description of 

Training Required 

and Completion Date

Comment Resolution and 

Closed Feedback Forms 

Accession Number (PreDecisional, Non-Public)

N/A 10/16/06

CN 06-027

This IMC has been revised to incorporate 

comments from the Commission in which 

the term public confidence has been 

change to openness

N/A N/A

N/A

07/08/09

CN-09-017

This revision moves post-qualification and 

refresher training requirements out of the 

appendix and into Appendix D-1.

N/A ML091590710

N/A ML15323A464

02/20/17

CN 17-004

This revision incorporates changes to 

reflect the transition from 6 baseline HP 

inspections to 8. 

N/A ML15324A264

ML16265A576