$$Appendix C-3

Health Physics Inspector

Technical Proficiency

Training and Qualification Journal

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Table of Contents

Introduction 5

Required Health Physics Inspector Training Courses 5

Additional Required Health Physics Inspector Training Courses: 5

Required Post-Qualification Training Courses: 5

**Required Refresher Training:** 5

Health Physics Inspector Individual Study Activities 7

(ISA-HP-1) Code of Federal Regulations (CFR=s) 9

(ISA-HP-2) Licensee Documents for Health Physics Inspectors 13

(ISA-HP-3) Access Controls to Radiologically Significant Areas 17

(ISA-HP-4) ALARA Planning and Controls 25

(ISA-HP-5) Radiation Monitoring Instrumentation 33

(ISA-HP-6) Radioactive Material Processing and Transportation 42

(ISA-HP-7) Radiological Environmental Monitoring Program (REMP) and
Radioactive Material Control Program 48

(ISA-HP-8) Radioactive Gaseous and Liquid Effluent Treatment and
Monitoring Systems 54

(ISA-HP-9) Significance Determination Process - Occupational Radiation

> Safety 60

(ISA-HP-10) Significance Determination Process - Public Radiation
Safety: Radioactive Material Control 64

(ISA-HP-11) Significance Determination Process - Public Radiation
Safety: Effluent Release Program 68

(ISA-HP-12) Significance Determination Process - Public Radiation
Safety: Environmental Monitoring Program 72

(ISA-HP-13) Significance Determination Process - Public Radiation
Safety: Transportation Branch 76

(ISA-HP-14) Performance Indicator - Occupational Radiation Safety 82

HEALTH PHYSICS INSPECTOR

ON-THE-JOB ACTIVITIES 86

(OJT-HP-1) Inspecting Access Controls to Radiologically Significant
Areas 88

(OJT-HP-2) Inspecting ALARA Planning and Controls 94

(OJT-HP-3) Inspecting Radiation Monitoring Instrumentation 98

(OJT-HP-4) Radioactive Material Processing and Transportation 104

(OJT-HP-5) Radiological Environmental Monitoring Program (REMP) and
Radioactive Material Control Program 110

(OJT-HP-6) Radioactive Gaseous and Liquid Effluent Treatment and
Monitoring Systems 114

Health Physics Inspector Technical Proficiency Level

Signature Card and Certification 120

**Introduction**

With the exception of the courses identified below, do not begin the
activities or complete the courses in this qualification journal until
you have completed the Basic Inspector Certification Journal. You may
complete the General Proficiency requirements contained in Appendix B
together with the Technical Proficiency requirements outlined in this
journal.

Several of the topics have both an individual study activity and an
on-the-job activity. You must complete the individual study activity
before beginning the corresponding on-the-job activity.

**Required Health Physics Inspector Training Courses:**

(These courses have the completion of Appendix A as a prerequisite)

(R-104B) - GE Technology

(R-104P) - Westinghouse Technology

(H-201) - Health Physics Technology

(H-202) - Radwaste Management

**Additional Required Health Physics Inspector Training Courses:**

(These courses DO NOT require the completion of Appendix A but you must
meet 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

**Required Post-Qualification Training Courses:**

> (To be completed within three years of initial qualification)
>
> Whole Body Counting/Internal Dosimetry (H-312)

**Required Refresher Training:**

Health Physics Topical Review Course (H-401) - every three years

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Health Physics Inspector

Individual Study Activities

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Health Physics Inspector Individual Study Activity

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

**PURPOSE:** The Code of Federal Regulations (CFR) provides 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 CFRs provide the content and scope that licensees must
comply with or receive NRC approval to deviate from the requirements.
For this reason, it is mandatory that all radiation protection
inspectors gain a general and comprehensive knowledge of the contents of
relevant radiation protection requirements in the CFR. This activity
will provide the inspector with detailed knowledge of the contents of
the requirements and how to apply the appropriate radiation protection
regulation requirements.

**COMPETENCY**

**AREA:** REGULATORY FRAMEWORK

**LEVEL OF**

**EFFORT**: 40 hours

**REFERENCES:** **40 CFR Part 190**, Environmental Radiation Protection
for Nuclear Power Operations

> 29 CFR 1910, Occupational Safety & Health Standards

**49 CFR Parts 170-189**, Transportation

> **10 CFR Part 19,** Notices, Instructions, and Reports to Workers
>
> **10 CFR Part 20**, Standards for Protection against Radiation
>
> 10 CFR Part 21, Reporting of Defects and Noncompliance
>
> 10 CFR Part 30, Rules of General Applicability to Domestic

Licensing of Byproduct Material

> 10 CFR Part 31, General Domestic Licenses for Byproduct Material
>
> 10 CFR Part 34, Licenses for Industrial Radiography and Radiation
> Safety Requirements for Industrial Radiographic Operations
>
> **10 CFR Part 50**, Domestic Licensing of Production and Utilization
> Facilities
>
> 10 CFR Part 51, Environmental Protection Regulations for Domestic
> Licensing and Related Regulatory Functions
>
> **10 CFR Part 61**, Licensing Requirements for Land Disposal of
> Radioactive Waste
>
> **10 CFR Part 71**, Packaging and Transportation of Radioactive
> Material
>
> 10CFR Part 74, Material Control and Accounting of Special Nuclear
> Material
>
> 10CFR Part 100, Reactor Site Criteria

**Note: Bolded CFRs are most important**

**EVALUATION**

**CRITERIA:** At the completion of this activity, and as determined by
the supervisor the inspector should be able to:

> 1\. Identify, recognize and locate specific radiation protection
> topics presented in the CFRs.
>
> 2\. Discuss and interpret the content of radiation protection
> requirements identified in the CFRs.
>
> 3\. Discuss and interpret the definitions of radiation protection
> terms identified in the CFRs.
>
> 4\. Recognize and discuss the regulatory bases for the Safety Analysis
> Report and the Technical Specifications as found in 10 CFR 50. Relate
> radiation protection requirements to those and other similar documents
> such as the Off-Site Dose Calculation Manual. 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 CFRs related to health physics to
> the (ROP). Discuss how the enforcement process as described in the
> CFRs is reflected in the implementation of the health physics aspects
> of the ROP. (See NUREG-1600, General Statement of Policy and
> Procedures for NRC Enforcement Actions, On NRC Office of Enforcement
> Web site)
>
> 6\. Relate the occupational and public radiation safety Performance
> Indicators to the requirements in the CFRs.
>
> 7\. Relate the requirements of the CFRs to the inspection objectives
> of the occupational radiation safety inspection attachments
> (71121-01,02,03). Cross reference CFR requirements inspection
> findings.
>
> 8\. Relate the requirements of the CRFs to the inspection objectives
> of the public radiation safety inspection attachments
> (71122-01,02,03). Cross reference CFR requirements to inspection
> findings.

**TASKS**: 1. Locate general and specific radiation protection
activities described in the CFRs.

> 2\. Compare and contrast the requirement contained in the health
> physics inspection procedure attachments to the requirements of the
> CFRs.
>
> 3\. Review the CFR to identify the regulatory bases for the radiation
> protection programs at nuclear power plants.
>
> 4\. Review the CFR to identify the regulatory significance of
> Performance Indicators.
>
> 5\. Review how the health physics significance determination processes
> are related to the enforcement policy of the CFRs.
>
> 6\. Meet with your supervisor, mentor, 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:** Health Physics Proficiency Level Qualification
Signature Card Item ISA-HP-1

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Health Physics Inspector Individual Study Activity

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

**PURPOSE:** The NRC requires that licensees maintain an Updated Final
Safety Analysis Report (UFSAR), which describes the design features,
systems, programs, and design basis operation of the facility. For this
reason, it is vital that health physics inspectors gain a detailed
knowledge of certain sections of the UFSAR. Additionally, the NRC
requires that licensees operate their facilities in compliance with the
Technical Specifications (TS), which are approved by the NRC; Technical
Requirements Manual (TRM); Process Control Program (PCP); and the
Offsite Dose Calculation Manual (ODCM). It is vital that all health
physics inspectors gain a detailed knowledge of the ODCM, TRM, PCP and
applicable sections of the TS. This activity will provide health physics
inspectors with the detailed knowledge of the contents of the UFSAR, TS,
TRM, and ODCM, where the applicable information and requirements for
specific topics, and how to apply the requirements.

**COMPETENCY**

**AREA:** TECHNICAL AREA EXPERTISE

REGULATORY FRAMEWORK

**LEVEL OF**

**EFFORT:** 40 Hours

**REFERENCES:** UFSAR, TS, TRM, PCP and ODCM for a facility designated
by your supervisor

**EVALUATION**

**CRITERIA:** At the completion of this activity, you should be able to:

15988. 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.

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1.  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.

2.  Discuss the content and basis for the requirements of the designated
    facility ODCM.

3.  Discuss definitions and terms found in the designated facility
    UFSAR, TS, TRM, PCP and ODCM.

4.  Discuss the legal basis, purpose, license conditions, and how these
    documents (UFSAR, TS, TRM, PCP and ODCM) can be changed.

5.  Discuss the delineated programs, processes, equipment, and limits,
    and the reasons they are required.

6.  Discuss the requirements for surveillances, action statements, and
    reporting.

7.  Discuss the administrative controls section of the TS and the types
    of information located in this section. Focus on High Radiation
    alternative controls and radiative materials effluent section

8.  Discuss the TRM purpose, legal basis, and how it can be changed.

**TASKS:** 1. Locate a copy of the UFSAR for the facility designated by
your supervisor and review the various sections as listed in the
Evaluation Criteria section.

2.  Locate a copy of the TS for the facility designated by your
    supervisor and review the various sections as listed in the
    Evaluation Criteria section.

3.  Locate a copy of the TRM for the facility designated by your
    supervisor and review the various sections as listed in the
    Evaluation Criteria section.

4.  Locate a copy of the ODCM for the facility designated by your
    supervisor and review the content as listed in the Evaluation
    Criteria section.

5.  Locate a copy of the PCP for the facility designated by your
    supervisor and review the content as listed in the Evaluation
    Criteria section

6.  Meet with your supervisor or a senior health physics 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:** Health Physics Inspector Proficiency Level
Qualification Signature Card, Item ISA-HP-2

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Health Physics Inspector Individual Study Activity

**TOPIC:** (ISA-HP-3) Access Controls to Radiologically Significant
Areas

**PURPOSE:** The purpose of this activity is familiarize you with the
regulatory basis and historical agency position related to access to
radiologically significant areas. The information is addressed in
generic correspondence, regulatory guides and regulations that pertain
to access controls to radiologically significant areas. The related
inspection procedure is IP 71121.01, AAccess Controls to Radiologically
Significant Areas@

The inspection procedure has three main objectives:

-   To review and assess licensee\'s performance in implementing
    physical and administrative controls for airborne radioactivity
    areas, radiation areas, high radiation areas (HRAs), and worker
    adherence to these controls.

-   To observe access controls to radiation and high radiation areas
    \<1000 mrem/hr and areas that are subject to transient dose rates.
    Review the controls that have been established and confirm that
    workers follow established rules. Review the high radiation area
    incidents found in the performance indicators (PIs) and in the
    licensee\'s corrective action program during the current assessment
    period.

-   To verify that the licensee is maintaining adequate controls over
    HRAs (with dose rates greater than 25 rem/h) and all very high
    radiation areas (VHRA).

**COMPETENCY**

**AREA:** REGULATORY FRAMEWORK

TECHNICAL AREA EXPERTISE

**LEVEL OF**

**EFFORT:** 40 hours

**REFERENCES:** 1. 10CFR19.12

2.  NUREG-1736, AConsolidated Guidance: 10 CFR Part 20- Standards for
    Protection Against Radiation@ [read the sections that address the
    following]{.underline}:

> 10CFR20.1003 (Definitions of Asurvey@, Aairborne radioactivity area@
> and ADAC@
>
> 10CFR20.1101
>
> 10CFR20.1203 through 1204
>
> 10CFR20.1501 through 1502
>
> 10CFR20.1601 through 1602
>
> 10CFR20.1701 through 1704
>
> 10CFR20.1902 through 1906
>
> 3\. Regulatory Guide 8.8, AInformation Relevant to Ensuring that
> Occupational Radiation Exposures at Nuclear Power Stations Will Be As
> Low As Is Reasonably Achievable@ \[ML003739549\]

4.  Regulatory Guide 8.25, AAir Sampling in the Workplace@
    \[ML003739616\]

5.  Regulatory Guide 8.38, AControl of Access to High and Very High
    Radiation Areas of Nuclear Plants@ \[ML003739558\]

6.  IN 82-31, \"Overexposure of Diver During Work in Fuel Storage Pool@

7.  IN 84-61, \" Overexposure of Diver in Pressurized Water Reactor(PWR)
    Refueling Cavity@

8.  IN 88-79, AMisuse of Flashing Lights for High Radiation Area
    Controls@

9.  IN 90-33, ASources of Unexpected Occupational Radiation Exposures at
    Spent Fuel Storage Pools@

> 10\. IN 90-47, AUnplanned Radiation Exposures to Personnel Extremities
> Due to Improper Handling of Potentially Highly Radioactive Materials@

11. IN 92-75, AUnplanned Intakes of Airborne Radioactive Materials By
    Individuals At Nuclear Power Plants@

12. IN 97-36, AUnplanned Intakes By Workers of Transuranic Airborne
    Radioactive Materials and External Exposure Due to Inadequate
    Control of Workers@

13. IN 97-68, ALoss of Control of Diver in a Spent Fuel Storage Pool@

**EVALUATION**

**CRITERIA:** Upon completion of this independent study activity you
should be able to:

18732. Discuss the regulatory requirements associated with airborne
       radioactivity areas, radiation areas, high radiation areas, and
       very high radiation areas.

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19080. Discuss the regulatory requirements associated with surveys and
       be able to describe what may constitute a survey other than
       documented direct meter measurements.

> 3\. Discuss the answers to the questions associated with each scenario
> given in the context of regulatory requirements without using licensee
> procedures.

**TASKS:** 1. Read each reference and discuss its application with
Senior Health Physics Inspector.

2.  Read the scenarios and answer the associated questions.

3.  Meet with your supervisor or a qualified health physics inspector to
    discuss any questions you may have as a result of this training
    activity. 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 spent
> coolant volume control system (CVCS lets down primary coolant and
> filters and returns the coolant to the operating loops) filters into
> 55 gallon drums that have accumulated and been stored in the filter
> room for decay for the last several years.
>
> A radiological pre-job survey was done earlier this week. The RP
> (Radiation Protection) tech providing job coverage will confirm survey
> results on entry.
>
> There are approximately 110 old wound fiber filters that originally
> had dose rates of up to 40 R/hr contact and 3 of the newer type high
> capacity submicron cartridge filters that have dose rates up to 600
> R/hr contact. The general area around where the filters are piled is
> about 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 ft 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 cm^2^ are common. Airborne activity in the room is typically about
> 0.2 DAC with no activity going on and gets up to about 0.4 DAC when
> someone walks in there to dose rate 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 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:]{.underline}
>
> What are the legally required postings for the room?
>
> What are the significant radiological risks associated with the job?
>
> What surveys are appropriate given the above scenario?
>
> Where should air sampling be done? Why?

**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 Aelephant
trunk@ (flexible 6 inch ventilation hose) run in through the open door
and the discharge was routed using an Aelephant 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) was chosen to provide an appropriate
protection factor, provide cooling and reduce worker fatigue.

[Questions:]{.underline}

> Does the door blocked open with the Aelephant trunk@ pose a potential
> regulatory compliance problem?
>
> What are the licensee options with regard to the Aelephant trunk@?
>
> Can flashing lights be used under these circumstances?
>
> What is the protection factor allowed for a PAPR when the blower is
> running?
>
> What is the protection factor allowed for a PAPR when the blower is
> not running?
>
> 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 airborne radioactivity
> area required?

**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.

[Question:]{.underline}

> Calculate the estimated uptake by the individual.

**Assume:** 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:]{.underline}

> Does the room have to be posted airborne?
>
> Can credit be taken for the PAPR respirators to avoid posting the room
> airborne radioactivity area?
>
> Can credit be taken for the PAPR respirator when assigning DAC-Hrs 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 uptake 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:]{.underline}

> What are the potential violations that could be identified?
>
> 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 10 year in-service inspection
of 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 rate. 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. 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.(floating plexiglass
> windows that eliminate the visual distortion caused by ripples on the
> water.) All communications with diver will be through the Dive
> Supervisor. The Dive Supervisor has wired 2 way communication 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 as far from the edge of the platform as the diver
> can reach without lifting his feet from the platform in all
> directions.
>
> Chemistry will sample the water once a shift to determine nuclide
> content and activity. Tritium will be analyzed for 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:

> What are the radiological hazards associated with this job?
>
> What is the purpose of the dive platforms?
>
> Does the RP organization have to post signs under water?
>
> Does the RP organization have to post the entire cavity as a Very High
> Radiation Area while diving operations are in progress?
>
> Why is the alarming dosimeter in the diver=s helmet instead of on his
> chest?
>
> Why is the water analyzed periodically? Is it a regulatory
> requirement?
>
> Why is extremity dosimetry required?
>
> 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.
>
> Does the RP technician performing continuous coverage from the cavity
> walkway meet requirements?

**DOCUMENTATION**: Health Physics Inspector Proficiency Level
Qualification Signature Card Items ISA-HP-3.

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Health Physics Inspector Individual Study Activity

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

**PURPOSE:** The purpose of this activity is to familiarize you with
concepts that will be important to conducting the related inspection
requirements contained in IP 71121.02 AALARA Planning and Controls@

> The purpose of that procedure 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.
>
> ALARA is the central concept of occupational radiation protection. In
> the United States the linear non-threshold 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 increase 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.
>
> National Committee for Radiation Protection (NCRP) publication 116
> ALARA Guidance can be summarized as follows:
>
> Justification-The need to justify radiation dose on the basis of
> benefit.
>
> Optimization-The need to ensure the benefits are maximized.
>
> Limitation-The need to apply dose limits.

**COMPETENCY**

**AREA:** TECHNICAL AREA EXPERTISE

REGULATORY FRAMEWORK

**LEVEL**

**OF EFFORT**: 40 hours

**REFERENCES:** 1. 10 CFR 20.1101

4\. NUREG 1736,@ Consolidated Guidance:10 CFR 20 - Standards for
Protection Against Radiation,@ Section 3.20.1101

5.  Regulatory Guide 8.8, A Information Relevant to Ensuring that
    Occupational Radiation Exposures at Nuclear Power Plants Will Be as
    Low As Is Reasonably Achievable@

6.  Regulatory Guide 8.10, AOperating Philosophy for Maintaining
    Occupational Radiation Exposures as Low As Is Reasonably Achievable@

7.  Radiation Field Control Manual - 1997 Revision, Final Report,
    October, 1997, EPRI TR-107991

8.  HPPOS-018, 020, 021, 022, 217, AHealth Physics Positions@, NUREG/CR-
    5569, Rev 1.

9.  Regulatory Guide 8.13, AInstructions Concerning Prenatal Radiation
    Exposure@

10. Regulatory Guide 8.36, ARadiation Dose to the Embryo/fetus@

11. HPPOS-091 (Lead Shielding Safety Concerns)

12. NUREG/CR-5631 and NCRP 128, Radionuclide Exposure of Embryo/Fetus

**EVALUATION**

**CRITERIA:** Upon completion of the tasks in this independent study
activity you should be able to:

20796. Explain ALARA using a working definition that would be
       understandable to someone not affiliated with nuclear power or
       its regulation.

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1.  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.)

2.  Name the activities which account for the greatest percentage of the
    collective exposure received in a nuclear power plant.

3.  Explain where in the life cycle of a nuclear plant the most
    significant ALARA impacts can be achieved.

4.  Explain why corrosion control is an ALARA consideration and how it
    can affect the overall radiation exposure of the plant staff.

5.  Explain the key elements to an ALARA program as described in
    Regulatory Guide 8.8.

6.  Explain the NRC philosophy with regard to assigning a set dollar
    value for ALARA planning purposes.

7.  Explain the interrelationship between economics and other factors
    such as other risk.

8.  Explain the concept of ATEDE ALARA@.

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

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

11. 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.

12. Explain the modes by which radioactive material can be deposited in
    the body.

13. 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.

14. 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.

15. Describe how the ALARA work planning is a vital, key component in
    the identification hazards, benefits of early HP and working craft
    involvement and coordination and how this affects station collective
    exposures.

16. Describe and discuss the benefits of using employee ALARA
    suggestions, feedback and lessons learned in the planning and
    evaluation process.

17. 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.

18. Identify the information that should be available during an
    inspection of an ALARA package.

19. Discuss the qualifications and requirements for HP technician shift
    coverage and its purpose.

20. Explain how ALARA planning depends on extrapolating the future from
    past experience where successes and failures can be identified to
    aid in future planning.

21. State the regulatory basis of ALARA and identify what condition must
    exist in order to cite a violation on basis of ALARA.

22. Describe the responsibility of a licensee for limiting, monitoring
    and tracking the exposure of a declared pregnant worker.

    a.  Discuss the timing of licensee actions.

    b.  Discuss the embryo/fetus exposure limits

    c.  Discuss the inspection considerations detailed in IP 71121.02
        section 02.07.

**TASK:** 1. Read Regulatory Guide 8.8 and complete the following:

v.  Define ALARA.

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d.  Identify the factors that must be taken into account when making
    ALARA recommendations.

e.  Identify the types of activities typically accounting for the
    majority of the exposures.

f.  Identify where in the life cycle of a power plant ALARA
    considerations should be taken.

g.  Identify how initial material selection and subsequent chemistry
    control regimens to reduce reactor, feed and steam generator systems
    corrosion contribute to the ALARA concept.

h.  Identify the key elements of a radiation control program

i.  Determine the NRC defined dollar per rem amount to be used in these
    analyses for occupational exposure(if any)?

j.  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]{.underline} be performed?

k.  Determine if ALARA applies equally to internal and external
    exposures.

l.  List the 4 factors that determine the intensity of a radiation field
    at a given point.

m.  List the 3 principles/techniques that are commonly used to reduce
    exposure to radioactive material.

n.  List the 3 modes of entry for radioactive materials into the body.

o.  Determine how the risk of worker skin contamination fits into ALARA
    decision making.

p.  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?

o.  Identify the purpose of the ALARA briefings and how they contribute
    to the ALARA program.

p.  List the typical contents of a radiation work permit as described in
    Regulatory Guide 8.8.

q.  List the typical contents and considerations of an AALARA Package.@

r.  Determine the purpose of having a health physics (Radiation Safety
    or Radiation Protection) technician assigned to each operating
    shift.

s.  Determine the purpose of a post-operational debriefing with regard
    to ALARA.

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2.  Review the regulatory basis, including the statements of
    consideration, for the application of the ALARA regulations.

3.  Review the listed references related to a declared pregnant worker
    (DPW) and discuss the technical basis for the DPW dose limit.

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

5.  Meet with your supervisor or a qualified health physics 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.

**DOCUMENTATION:** Health Physics Inspector Proficiency Level
Qualification Signature Card Item ISA-HP-4.

**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 | Actual     |
|                                            | Dose      | Dose       |
|                                            | (Rems)    |            |
|                                            |           | (Rems)     |
+--------------------------------------------+-----------+------------+
| Scaffolding in the reactor building        | 13.000    | 26.125     |
+--------------------------------------------+-----------+------------+
| Remove and install steam generator manway  | 4.234     | 7.675      |
| covers and inserts                         |           |            |
+--------------------------------------------+-----------+------------+
| Steam generator Eddy current testing/tube  | 18.985    | 42.295     |
| plugging                                   |           |            |
+--------------------------------------------+-----------+------------+
| Health physics support for Refueling       | 3.450     | 6.279      |
| Outage 15                                  |           |            |
+--------------------------------------------+-----------+------------+
| Spent fuel pool cleanup/diving operations  | 3.198     | 6.723      |
+--------------------------------------------+-----------+------------+
| Reactor coolant pump seal removal and      | 3.745     | 7.176      |
| replacement                                |           |            |
+--------------------------------------------+-----------+------------+

> 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 pump (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, Aan 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.

6.  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 Athe crews were inexperienced@ and Aused 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 Apoor.@ Radiation protection personnel Aseldom@ knew
    job status or the schedule for the upcoming shift work. Therefore,
    they could not plan their activities to reduce dose.

    There was a Alack 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.

  -------------------------- ----------- ----------- ---------- -----------
                             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  142.6       149.2       159.6      191.2
  Dose                                                          
  -------------------------- ----------- ----------- ---------- -----------

[Questions:]{.underline}

> What are acceptable reasons for giving credit for exceeding dose
> projections?

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

What constitutes an ALARA finding?

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

When would an ALARA violation be warranted?

Health Physics Inspector Individual Study Activity

**TOPIC:** (ISA-HP-5) Radiation Monitoring Instrumentation.

**PURPOSE:** The purpose of this study activity is to familiarize you
with various regulatory requirements that support Inspection Procedure
71121.03, Radiation Monitoring Instrumentation. The inspection procedure
has two basic objectives which are (1) To determine the accuracy and
operability of radiation monitoring instruments that are used for the
protection of occupational workers and (2) To determine the adequacy of
program to provide self-contained breathing apparatus (SCBA) for
entering and working in areas of unknown radiological and/or potential
immediately dangerous to life and health (IDLH) areas.

**COMPETENCY**

**AREA:** TECHNICAL AREA EXPERTISE

REGULATORY FRAMEWORK

**LEVEL**

**OF EFFORT:** 32 hours

**REFERENCES:** 1. 10 CFR 20.1703(a)

2.  10 CFR 20.1501(b)

3.  HPPOS-001 AProposed Guidance for Calibration and Surveillance
    Requirements to meet Item II.F.1 of NUREG 0737

4.  HPPOS-088 ACorrections for Sample Conditions for Air and Gas
    Monitoring@

5.  HPPOS-279 A Technical Assistance Request Regarding Electronic
    Calibration of Survey Instruments@

6.  HPPOS-328 AProper Operation and Use of Alarm Dosimeters at Nuclear
    Power Plants@

7.  Regulatory Guide 8.15, Rev.1 AAcceptable Programs For Respiratory
    Protection@

8.  Information Notice 85-87 AHazards of Inerting Atmospheres@

9.  Information Notice 98-20 AProblems with Emergency Preparedness
    Respiratory Protection Programs@

10. HPPOS-147 ARespirator User=s Notice -Use of Unapproved Subassemblies
    (NIOSH warns against the use of unapproved subassemblies (parts and
    components) and unauthorized modifications for / of approved
    respirators)@

11. 29 CFR 1910 Subpart I, APersonal Protective Equipment@

12. NUREG 1736,@ Consolidated Guidance:10 CFR 20 - Standards for
    Protection Against Radiation,@ Sections 3.20.1703 and 1501

13. Federal Register Notice 64 FR 54543, Part 20, Subpart H revision in
    1999, Final Rule Summary, Statements of Consideration and
    Supplemental Information.

**EVALUATION**

**CRITERIA:** Upon completion of the tasks listed below you should be
able to :

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

2.  Describe calibration techniques including where electronic source
    usage is appropriate.

3.  Describe the appropriate uses for alarming dosimeters and their
    limitations.

4.  Explain the NRC requirements associated with calibration interval.

5.  Describe the guidance in place to assess the implementation of 10
    CFR 20.1501(b).

6.  Explain the application of the relevent portions of 10 CFR 20 which
    pertain to respiratory protective equipment.

7.  Describe the essential elements of basic respiratory protection
    program that would be acceptable to the NRC.

8.  Describe the characteristics of Grade D air quality, and where,
    when, and how air quality is tested.

9.  Explain the potential consequences of not using SCBA in IDLH
    environments, and describe the testing, maintenance, and replacement
    cycles for the various SCBA cylinders.

10. Describe deficiencies that are commonly found in respiratory
    protection training programs for the use of SCBA and how they might
    impact emergency preparedness.

11. 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 HPPOS-088 which pertains to calibration of air
sampling equipment. (The following equation is provided to help with
understanding the issue.)

> $V_{s} = V_{c}(\frac{P_{c}}{P_{s}})(\frac{T_{s}}{T_{c}})$ (See NUREG
> 1400 for further details)

Gas Law Correction

> Where:
>
> V~s~= volume under field conditions (appropriate volume unit)
>
> V~c~=volume under calibration conditons (appropriate volume unit)
>
> P~s~=absolute pressure during sampling (mm Hg)
>
> P~c~=absolute pressure during calibration (mm Hg)
>
> T~s~=absolute temperature during sampling (Kelvin)
>
> T~c~=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 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(b)

6.  Read 10 CFR 20, Subpart H portions applicable to SCBA.

7.  Read Regulatory Guide 8.15 (Rev. 1), AAcceptable Programs For
    Respiratory Protection@

8.  Read Information Notice 85-87, AHazards of Inerting Atmospheres@

9.  Read Information Notice 98-20, AProblems with Emergency Preparedness
    Respiratory Protection Programs@ (This one is particularly important
    from an inspection standpoint.)

10. Read HPPOS-147, ARespirator User=s Notice -Use of Unapproved
    Subassemblies@ (This is a recurrent problem.)

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 health physics 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.

**DOCUMENTATION:** Health Physics Inspector Proficiency Level
Qualification Signature Card Item ISA-HP-5.

**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 and he so 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 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.

**Question:**

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

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

Could this a violation of TSs?

**Scenario B: Calibration Instrumentation**

During a tour of the radiologically controlled area on July 18, 2001, an
inspector identified a continuous air monitor in the radwaste 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:**

Is there a technical specification that requires written procedures?

Why regulatory guide recommends procedures for area, portable, and
airborne radiation monitor calibrations?

Would this be a violation? Why?

**Scenario C Respiratory Protection Equipment**

During a review of self-contained breathing apparatus maintenance and
surveillance records, an inspector identified that 36 self-contained
breathing apparatus air bottles were past the 3-year hydrostatic test
dates. Hydrostatic testing had expired in April 2001 for 31 of the
self-contained breathing apparatus air bottles that were in service.
According to the NIOSH, self-contained breathing apparatus units with
expired hydrostatic testing are no longer certified.

**Questions**

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

Could this be a violation? Why or why not.

Other than Radiation Protection, what NRC cornerstone would also need to
be evaluated?

**Additional References Related to ISA-HP-5, Radiation Monitoring
Instrumentation**.

**Instrument**

10 CFR 20.1501(b)

Regulatory Guide 8.6 AStandard Test Procedure for Geiger-Muller Counters

> ANSI N42.17A-1989, APerformance Specifications for Health Physics
> Instrumentation-Portable Instrumentation for Use in Normal
> Environmental Conditions@
>
> ANSI N42.17B-1989,@Performance Specifications for Health Physics
> Instrumentation-Occupational Airborne Radioactivity Monitoring
> Instrumentation@
>
> ANSI N42.17C-1989,@Performance Specifications for Health Physics
> Instrumentation-Portable Instrumentation for Use in Extreme
> Environmental Conditions@
>
> ANSI N323A-1997, ARadiation Protection Instrumentation and
> Calibration, Portable Survey Instruments@

Q&A 147 Calibration Frequency

Q&A 209 Calibration Frequency

**SCBA**

> 10 CFR 20.1700 Subpart H- ARespiratory Protection and Controls to
> Restrict Internal Exposure in Restricted Areas@
>
> NUREG/CR-0041, Rev. 1 AManual of Respiratory Protection Against
> Airborne Radioactive Materials@
>
> Information Notice 83-68 ARespirator User Warning: Defective Self
> Contained Breathing Apparatus Air Cylinders@
>
> Information Notice 85-48 ARespirator User Warning: Defective Self
> Contained Breathing Apparatus Air Cylinders@

Information Notice 86-103 ARespirator Coupling Nut Assembly Failures@

> Information Notice 89-47 A Potential Problems with Worn or Distorted
> Hose Clamps on Self Contained Breathing Apparatus@
>
> Information Notice 94-35 ANIOSH Respirator User Notices,@ Inadvertent
> Separation of the Mask Mounted Regulator (MMR) from the Facepiece of
> the Mine Safety Appliances (MSA) Company Self Contained Breathing
> Apparatus (SCBA) and Status Update@
>
> Information Notice 95-01 ADOT Safety Advisory: High Pressure Aluminum
> Seamless and Aluminum Composite Hoop Wrapped Cylinders@
>
> Q&A 91 AClarifies the need to comply with programmatic requirements
> when using respirators@
>
> Q&A 124 ANotes that this section=s requirements apply to respirators
> used during emergencies@
>
> Q&A 418 AExplains that licensees need a formal program whenever a
> respirator is used to limit intake@
>
> HPPOS 094, AGuidance Concerning Bearded Users of SCBA - Beards no
> longer

allowed, but background information still pertinent and useful.

> HPPOS 103, AGuidance for Medical Evaluations for Respiratory Users

Health Physics Inspector Individual Study Activity

**TOPIC:** (ISA-HP-6) Radioactive Material Processing and Transportation

**PURPOSE:** The purpose of this individual study activity (ISA) is to
familiarize you with the regulatory bases and historical agency
positions in the area of radioactive material processing and
transportation. Specifically, the information provides you the necessary
technical knowledge to conduct inspections using procedure 71122.02,
Radioactive Material Processing and Transportation. 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. This
activity will provide you with detailed knowledge of the requirements
contained in regulations and position documents.

**COMPETENCY**

**AREA:** TECHNICAL AREA EXPERTISE

**LEVEL OF**

**EFFORT:** 40 HOURS

**REFERENCES:** Title 10, Code of Federal Regulations, Parts 20, 61, and
71

> Title 10, Code of Federal Regulation, Part 30.41
>
> Title 49, Code of Federal Regulations, Parts 100-189
>
> Radwaste system and facility description from the UFSAR for a facility
> designated by your supervisor

Plant Annual Effluent Release Report

Offsite Dose Calculation Manual

> Inspection Procedure 71122, APublic Radiation Safety@

NRC Branch Technical Position, Waste Form Technical Position

> NRC Branch Technical Position on Concentration Averaging and
> Encapsulation
>
> NUREG - 1608, Categorizing and Transporting Low Specific Activity
> Materials and Surface Contaminated Objects
>
> NUREG-1660, U.S. - Specific Schedules of Requirements for Transport of
> Specified Types of Radioactive Material Consignments
>
> NUREG/BR-0204, Instructions for Completing NRC=s Uniform Low-level
> Radioactive Waste Manifest

**EVALUATION**

**CRITERIA:** At the completion of this activity, you 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).

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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.

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 rate 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
    radwaste as outlined in 10 CFR Part 61.

11. Discuss Quality Assurance Requirement for a radioactive materials
    shipping program, as described in 10 CFR 71.

**TASKS:** 1. Review and familiarize yourself with the documents listed
in the reference list and in the list of other important references.
Specifically, identify the purpose of each document and what guidance
the document provides.

2.  Locate a copy of the radwaste section from the UFSAR of your
    designated facility. Review the design and operation of the radwaste
    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 characteristics requirements for disposal of solid
    radioactive waste.

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.  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.

**DOCUMENTATION:** Health Physics Proficiency Level Signature Card Item
ISA-HP-6

**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 AGreen is Clean@
potentially contaminated dry waste and the other dry active waste (DAW).
The sea-land containers were loaded back-to-back on an open, flat bed
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) material, 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 coin-sized (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 Ahot@ spot identified at waste
processing contractor.

**Questions**

What two Code of Federal Regulations apply to this condition? (NRC/DOT)

What is the DOT limit on external surface of containers?

What are the possible reasons for the licensee Amissing@ the hot spot on
the container?

**Scenario B**

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, Co-58, 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 A\...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.

**Question:**

What Code of Federal Regulations applies?

**Other Important References Related to this Topic**

Information Notice 86-20, Technical Position on Concentration Averaging

Information Notice 90-50, Minimization of Methane Gas in Plant Systems
and Radwaste Shipping Containers

Information Notice 92-62, Emergency Response Information Requirements
for Radioactive Material Shipments

Information Notice 92-72, Employee Training and Shipper Registration
Requirements for Transporting Radioactive Materials

Information Notice 95-09, Monitoring and Training of Shippers and
carriers of Radioactive Materials

NRC Generic Letter 81-38, Storage of Low-Level radioactive Wastes at
Power Reactor Sites

Note: The information in this Generic Letter is staff guidance (and not
requirements).

NRC Bulletin 80-10, Contamination of Non-Radioactive System and
Resulting Potential for Unmonitored, Uncontrolled Release to Environment

ANSI/ANS-40.37-1993, Mobile Radioactive Waste Processing Systems

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

NUREG/CR-6204, Questions and Answers Based on Revised 10 CFR Part 20

NUREG 1736, Consolidated Guidance: 10 CFR Part 20- Standards for
Protection Against Radiation

Health Physics Inspector Individual Study Guide

**TOPIC:** (ISA-HP-7) Radiological Environmental Monitoring Program
(REMP) and Radioactive Material Control Program

**PURPOSE:** The purpose of this individual study activity (ISA) is to
familiarize you with the regulatory bases and historical agency
positions in the areas of radiological environmental monitoring and
radioactive material control. Specifically, the information provides you
the necessary technical knowledge to conduct inspections using procedure
71122.03, Radiological Environmental Monitoring Program (REMP) and
Radioactive material Control. 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 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 CFR 50.34a. The radioactive material control
program verifies that the licensee maintains a program to ensure that
licensed radioactive material is controlled in accordance with the
requirements of 10 CFR Part 20.

> This activity will provide you with detailed knowledge of the
> requirements contained in regulations and position documents in the
> area of radiological environmental monitoring and radioactive material
> controls.

**COMPETENCY**

**AREA:** TECHNICAL AREA EXPERTISE

**LEVEL OF**

**EFFORT:** 40 HOURS

**REFERENCES:** 1. Title 10, Code of Federal Regulations, Parts 20 and
50

2.  Inspection Procedure 71122, APublic Radiation Safety@

3.  Plant Updated Final Safety Analysis Report

4.  IE Circular 81-07, Control of Radioactively Contaminated Material

5.  IE Information Notice 85-92, Surveys of Wastes Before Disposal From
    Nuclear Reactor Facilities

6.  Regulatory Guide 1.23, \@On-site Meteorological Program@

7.  Regulatory Guide 1.111, AMethods of Estimating Atmospheric Transport
    and Dispersion of Gaseous Effluents in Routine Release from
    Light-Water-Cooled Reactors@

8.  Regulatory Guide 4.1, APrograms for Monitoring Radioactivity in the
    Environs of Nuclear Power Plants@

9.  Regulatory Guide 4.13, APerformance, Testing, and Procedural
    Specifications for Thermoluminescence Dosimetry: Environmental
    Application@

10. Regulatory Guide 4.15, AQuality Assurance for Radiological
    Monitoring Programs (Normal Operations) BEffluent Streams and the
    Environment@

11. NUREG/CR-5569, Rev. 1, Health Physics Positions Data Base

12. NUREG/CR-6204, Questions and Answers Based on Revised 10 CFR Part 20

13. NUREG 1736, Consolidated Guidance: 10 CFR Part 20- Standards for
    Protection Against Radiation

**EVALUATION**

**CRITERIA:** At the completion of this activity, you should be able to:

26988. Discuss the specific regulatory requirements that require a
       licensee to have a REMP

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1.  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.

2.  Discuss the different measuring techniques (proportional counter,
    gamma spectroscopy, liquid scintillation counter, and TLD reader)
    that may be used by the licensee and the capabilities and
    limitations of each.

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
    intra-laboratory 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 χ/Q and D/Q, and annual average data.

8.  Discuss the principle parts of a REMP Annual Report and identify the
    maximum population doses.

9.  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 with guidance in NRC Circular
    81-07 and IE Information Notice 85-92.

10. Discuss the contents of the licensee=s ODCM and explain and compare
    the contents to NUREG 1301/1302 (or Branch Technical Position,
    November 1979).

11. 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.

12. Explain the Fundamentals of Laboratory QA/QC Policy and its
    implementation.

13. Identify the specific radiological dose limits for your assigned
    facility as described in the REMP and ODCM.

14. Identify the specific environmental sample requirements for your
    assigned facility as described in the REMP and ODCM.

15. Discuss radiological analytical methodology for tritium, gamma,
    ambient radiation (using TLD), and gross alpha and beta.

16. Discuss and identify free release criteria for contaminated
    materials from RCA to the public domain.

17. Discuss the sequential relationship between Radiological Effluent
    Controls and the REMP.

**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, SAM, and
> TLD reader).

5.  Review counting statistics and data reduction, including minimum
    detectable activity (MDA) and lower limits of detection (LLDs).

6.  Review radiological analytical methodology for tritium, gamma,
    ambient radiation (using TLD), and gross alpha and beta.

7.  Review free release criteria of contaminated materials from RCA to
    the public domain.

8.  Review meteorological monitoring requirements and calibration
    results for wind direction, wind speed, and delta temperature.

9.  Review the sequential relationship between Radiological Effluent
    Controls and the REMP.

10. Review man-made and natural radiation exposure pathways
    (fission/activated products and the source of the natural background
    radiation).

11. Review the Fundamental of Laboratory QA/QC Policy and its
    implementation.

**DOCUMENTATION:** Health Physics Proficiency Level Signature Card Item
ISA-HP-7

**Other Important References Related to ISA-HP-7**

Environmental Measurement Laboratory, HASL-300 Procedures Manual U.S.
Department of Energy, New York, NY

NCRP Report No. 45, ANatural Background Radiation in the US@

NCRP Report No. 47, ATritium measurement Techniques@

NCRP Report No. 50, AEnvironmental Radiation Measurements@

NCRP Report No. 58, AA Handbook of Radioactivity Measurements
Procedures@

ANSI N545-1975, APerformance, Testing, and Procedural Specification for
TLD, Environmental Application@

ANSI N13.1-1969, AGuide to Sampling Airborne Radioactive materials in
Nuclear Facilities@

ANSI N13.4-1971, AAmerican National Standard for the Specification of
Portable X- or Gamma-Radiation Survey Instruments@

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Health Physics Inspector Individual Study Guide

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

**PURPOSE:** The purpose of this study activity is to familiarize you
with the regulatory bases and historical agency positions in the area of
radioactive gaseous and liquid effluent treatment and monitoring
systems. The NRC requires that licensee=s 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 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. This activity will provide you with
detailed knowledge of the requirements contained in these regulations.

**COMPETENCY**

**AREA:** TECHNICAL AREA EXPERTISE

**LEVEL OF**

**EFFORT:** 45 HOURS

**REFERENCES:** 1. Title 10, Code of Federal Regulations, Parts 20 and
50

2.  Title 40, Code of Federal Regulations, Part 190

3.  Plant Updated Final Safety Analysis Report

4.  Inspection Procedure 71122, APublic Radiation Safety@

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

6.  NUREG/CR-6204, Questions and Answers Based on Revised 10 CFR Part 20

7.  NUREG 1736, Consolidated Guidance: 10 CFR Part 20- Standards for
    Protection Against Radiation

**EVALUATION**

**CRITERIA:** 1. Discuss and identify effluent sampling techniques that
may be used by the licensee for sampling effluents such as radioactive
waste water, air iodine, and air particulate.

2.  Discuss different measuring techniques, such as proportional
    counter, gamma spectroscopy, and liquid scintillation counter that a
    licensee may use to analyze its samples. Compare and contrast the
    techniques..

3.  Discuss calibration techniques for a laboratory=s gamma
    spectroscopy, proportional counter, and liquid scintillation
    counter. Identify the types of source to be used an expected MDA or
    LLDs

4.  Discuss laboratory analytical techniques for radionuclides and
    radiations such as gamma, Sr-89/90, H-3, and gross alpha/beta.
    Identify best methods.

5.  Discuss laboratory QA/QC Policy and its implementation, including
    inter-laboratory and intra-laboratory comparisons. Discus NRC
    expectations in this area.

6.  Discuss and calculate minimum detectable activity (MDA) and lower
    limits of detection (LLDs) for hypothetical samples.

7.  Discuss calculation of projected public dose calculation
    methodologies (all pathways) listed in Regulatory Guide 1.109.

8.  Discuss the characteristics of typical radiation monitoring systems
    (RMS), such as gamma-scintillation detector, beta-scintillation
    detector, GM, and ion chamber used in a power plant environment and
    compare and contrast their capabilities

9.  Discuss principle of air cleaning systems, such as air capacity
    test, in-place testings, and laboratory test to determine the iodine
    collection efficiency. Discuss applicable guidance documents and
    acceptable values

10. Review the air cleaning system section from your assigned facility=s
    UFSAR and Technical Specification. Discuss the principal components
    and their use.

11. Identify the various sources of liquid and gaseous radwaste at your
    assigned facility by review of the UFSAR. Discuss the UFSAR
    described waste streams, and technologies associated with liquid and
    gaseous radwaste processing for the facility.

**TASK:**

29568. Review and evaluate the effluent sampling techniques, such as
       radioactive waste water, air iodine, and air particulate.

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2.  Review and evaluate the different measuring techniques, such as
    proportional counter, gamma spectroscopy, and liquid scintillation
    counter.

3.  Review and evaluate the calibration techniques and results for the
    measurement laboratory=s gamma spectroscopy, proportional counter,
    and liquid scintillation counter.

4.  Review and evaluate understand the measurement laboratory=s
    analytical techniques (e.g., gamma, Sr-89/90, H-3, and gross
    alpha/beta).

5.  Review and evaluate the laboratory=s QA/QC Policy and its
    implementation, including inter-laboratory and intra-laboratory
    comparison.

6.  Review and evaluate the minimum detectable activity (MDA) and lower
    limits of detection (LLDs) used by the licensee.

7.  Review and evaluate the projected public dose calculation
    methodologies (all pathways) listed in Regulatory Guide 1.109.

8.  Review and evaluate the characteristics of radiation monitoring
    systems (RMS), such as gamma-scintillation detector,
    beta-scintillation detector, GM, and ion chamber used by the
    licensee. Identify capabilities and limitations.

9.  Review and evaluate principle air cleaning system tests, such as air
    capacity test, in-place testings, and laboratory test to determine
    the iodine collection efficiency. Compare with applicable regulatory
    limits

10. Conduct a walk down of the air cleaning systems described in the
    UFSAR and Technical Specification. Compare your observations to the
    licensee=s descriptions.

11. Identify the various sources of liquid and gaseous radwaste, waste
    streams, and technologies associated with liquid and gaseous
    radwaste processing for the facility. Compare that with the
    descriptions presented in the UFSAR.

**OTHER IMPORTANT REFERENCES RELATED TO ISA-HP-8**

Environmental Measurement Laboratory, HASL-300 Procedures Manual U.S.
Department of Energy, New York, NY

NCRP Report No. 58, AA Handbook of Radioactivity Measurements
Procedures@

RG 1.21 \"Measuring and Reporting Radioactivity in Solid Wastes and
Releases of Radioactive Materials in Liquid and Gaseous Effluents from
Light-Water-Cooled Nuclear Power Plants\"

RG 1.52 \"Design, Testing, and Maintenance Criteria for Post-Accident
Engineered Safety Feature Atmosphere Cleanup System Air Filtration and
Adsorption Units of Light-Water-Cooled Nuclear Power Plants\" (Endorsed
ANSI N509-1976 and N510-1975)

RG 1.97 \"Instrumentation for Light-Water-Cooled Nuclear Power Plants to
Assess Plant Conditions During and Following an Accident\"

RG 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\"

RG 1.111 \"Methods for Estimating Atmospheric Transport and Dispersion
of Gaseous Effluents in Routine Releases from Light-Water-Cooled
Reactors\"

RG 4.15 \"Quality Assurance for Radiological Monitoring Programs (Normal
Operations) - Effluent Streams and the Environment\"

ANSI N13.1-1982 \"Guide to Sampling Airborne Radioactive Materials in
Nuclear Facilities\"

ANSI N45.2.23-1978 \"Qualification of Quality Assurance Program Audit
Personnel for Nuclear Power Plant Personnel\"

ANSI/ANS N55.4-1979 \"Gaseous Radioactive Waste Processing Systems for
Light-Water-Cooled Reactor Plants\"

ANSI/ANS N55.6-1979 \"Liquid Radioactive Waste Processing System for
Light-Water-Cooled Reactor Plants\"

ANSI/ASME N509-1980 \"Nuclear Power Plant Air Cleaning Units and
Components\"

ANSI/ASME N510-1980 \"Testing of Nuclear Air-Cleaning Systems\"

IE Bulletin No. 80-10 \"Contamination of Nonradioactive Systems and
Resulting Potential for Unmonitored, Uncontrolled Release of
Radioactivity to the Environment,\" May 6, 1980

IE Circular No. 80-18 \"10 CFR50.59 Safety Evaluations for Changes to
Radioactive Waste Treatment Systems,\" August 22, 1980

IE Information Notice No. 82-43 \"Deficiencies in LWR Air
Filtration/Ventilation Systems,\" November 16, 1982

IE Information Notice No. 82-49 \"Correction for Sample Conditions for
Air and Gas Monitoring,\" December 16, 1982

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

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Health Physics Inspector Individual Study Activity

**TOPIC**: (ISA-HP-9) Significance Determination Process - Occupational
Radiation Safety

**PURPOSE**: The risk significance of inspection findings are evaluated
using the significance determination process (SDP). The purpose of this
training is familiarize you with the Occupational Radiation 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.

**COMPETENCY**

**AREAS**: INSPECTION

TECHNICAL AREA EXPERTISE

REGULATORY FRAMEWORK

**LEVEL OF**

**EFFORT** 36 hours

**REFERENCES**: 1. NRC IMC 0609, ASignificance Determination Process@,

> Appendix C.

2.  SDP Branch on ALARA

    a.  NRC Inspection Report (50-483-0017), and NRC Responses to
        Licensee Contestation for White Findings at Callaway (EA-00-208,
        May, 2001)

    b.  NUREG-0713 (Latest Volume), [Occupational Radiation Exposure at
        Commercial Nuclear Power Reactors and Other
        Facilities]{.underline}

    c.  10 CFR Part 20, Standards for Protection Against Radiation;
        Final Rule, Supplementary Information (Statements of
        Consideration), Section VI., Subpart B (Radiation Protection
        Programs), pp 23366-23367, Federal Register Notice/Vol. 56, No.
        98, 5/21/91.

        3.  SDP Branch on Overexposure

            a.  Regulatory Guide 8.38, [Control of Access To High and
                Very High Radiation Areas in Nuclear Power
                Plants]{.underline}.

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        3.  SDP Branch on Substantial Potential for Overexposure (SPFO)

            a.  NUREG/BR-0195, ANRC ENFORCEMENT MANUAL@, Section 8.4.2.

            b.  HPPOS 232, NUREG/CR-5569, Rev. 1.

            c.  NRC Information Notice No. 88-63 (August 15, 1988); IN
                88-63, Supplement 1 (October 5, 1990)

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    a.  IN 88-63, Supplement 2 (June 25, 1991) and IN 02-03 (January 10,
        2002)

    b.  Davis-Besse Special Inspection Reports 02-06 and 02-16, ADAMS
        Accession \# ML030070606

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5.  SDP Branch on Ability to Assess Dose

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r.  Davis-Besse Special Inspection Reports 02-06 and 02-16, ADAMS
    Accession \# ML030070606

**EVALUATION**

**CRITERIA**: At the completion of this activity, you should be able to:

1.  Explain the purpose, objectives and applicability of the SDP
    process.

2.  Compare and contrast an Aobservation@ and a Afinding@.

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 bases for the four branches of the Occupational SDP.

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:

    a.  Process ALARA findings for a job through the branch, using
        example training scenarios.

    b.  Explain why a WHITE finding may not be an adequate basis for
        issuing a notice of violation against Part 20.1101(b)

    c.  Explain the basis for the AGreater than 4 Occurrences@ logic
        gate and the rationale for issuing a WHITE finding.

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1.  For the Overexposure Branch of the SDP, you should be able to
    process exposure findings through the branch, using example training
    scenarios.

2.  For the Substantial Potential for Overexposure (SPFO) Branch of the
    SDP, you should be able to:

    a.  Discuss and explain the SPFO enforcement policy tool, and give
        examples of SPFO findings.

    b.  Process exposure findings through the branch, using example
        training scenarios from NRC Information Notices and other
        sources.

    c.  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 millicuries of Co-60), are treated differently
        relative to SPFO.

    d.  Explain why an event with a SPFO that occurs in a Very High
        Radiation area merits a yellow finding.

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1.  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. Read MC 0609 and focus on Appendix C.

2.  Read the references pertinent to each SDP branch.

3.  Determine a facility=s current rolling-three-year-average (RTYA) and
    determine the median collective dose for both PWRs and BWRs.

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 Anear-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 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:** Health Physics Inspector Proficiency Level
Qualification Signature Card, Item ISA-HP-9

Health Physics Inspector Individual Study Activity

**TOPIC**: (ISA-HP-10) Significance Determination Process - Public
Radiation Safety: Radioactive Material Control

**PURPOSE**: The risk significance of inspection findings are evaluated
using the significance determination process (SDP). The purpose of this
training is familiarize you with the Radioactive Material Control
portion of the Public Radiation Safety 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.

**COMPETENCY**

**AREAS**: INSPECTION

REGULATORY FRAMEWORK

TECHNICAL AREA EXPERTISE

**LEVEL OF**

**EFFORT**: 24 hours

**REFERENCES:** 1. NRC IMC 0609, ASignificance Determination Process@,
Appendix D, Radioactive Material Control Branch.

2.  10 CFR Part 20, Standards for Protection Against Radiation.

> a\. Subpart F, Surveys and Monitoring; 20.1501(a) and 20.1501(b)
>
> b\. Subpart K, Waste Disposal.

3.  NUREG/CR-5569, Revision 1, Health Physics Positions Data Base.

4.  Health Physics Position (HPPOS) numbers; 42, 43, 44, 45, 48, 71, 72,
    73, 79, 106, 138, 171, 189, 190, 221, 250, 300.

5.  NRC Circular 81-07, Control of radioactively contaminated material.

6.  NRC Information Notice 83-05, Obtaining approval for disposing of
    very-low-level radioactive waste - 10 CFR Section 20.302 (current 10
    CFR 20.2002).

7.  NRC Information Notice 85-92, Survey of wastes before disposal from
    nuclear reactor facilities.

8.  NRC Information Notice 86-90, Requests to dispose of very low-level
    radioactive waste pursuant to 10 CFR 20.302 (current 10 CFR
    20.2002).

9.  NRC Information Notice 88-22, Disposal of sludge from onsite sewage
    treatment facilities at nuclear power stations.

10. NRC Inspection Report, and NRC Responses to Licensee Contestation
    for White Findings at Comanche Peak

**EVALUATION**

**CRITERIA**: At the completion of this activity, you should be able to:

36624. Explain the purpose, objectives and applicability of the
       Radioactive Material Control SDP process.

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2.  Define the safety significance, and give examples of Green, White,
    Yellow and Red findings of this SDP branch.

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3.  Process a finding through the SDP branch.

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4.  Explain the NRR Ano 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
> Ahot 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 Ano
    detectable@ criteria.

8.  Describe the minimum detection sensitivity criteria that licensees
    must use for radiation surveys of potentially contaminated material.

**TASKS:** 1. Read MC 0609D, Radioactive Material Control.

2.  Read each of the 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 any 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 results in the inspection report. Discuss any differences you may
> have with a qualified health physics inspector.

4.  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

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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**: Health Physics Proficiency Level Qualification
Signature Card Item ISA-HP-10

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Health Physics Inspector Individual Study Activity

**TOPIC**: (ISA-HP-11) Significance Determination Process - Public
Radiation Safety: Effluent Release Program

**PURPOSE**: The risk significance of inspection findings are evaluated
using the significance determination process (SDP). The purpose of this
training is to familiarize you with the Radioactive Effluent portion of
the Public Radiation Safety SDP, explain the bases of this branch,
provide technical and policy references, and present practical
exercises/scenarios that challenge you to use the SDP.

**COMPETENCY**

**AREAS**: INSPECTION

REGULATORY FRAMEWORK

TECHNICAL AREA EXPERTISE

**LEVEL OF**

**EFFORT**: 32 hours

**REFERENCES**: See list at end of this activity

**EVALUATION**

**CRITERIA**: At the completion of this activity, you 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 Aimpaired ability to assess dose.@

> 5\. Explain the similarity and differences between Appendix I to 10
> CFR Part 50 and 40 CFR Part 190.
>
> 6\. Describe what information is contained in a licensee=s Offsite
> Dose Calculation Manual.

7.  Compare and contrast the dose assessment methodology in 10 CFR Part
    20 to that in Appendix I of 10 CFR Part 50.

8.  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.

**TASKS:** 1. Read MC 0609D, Effluent Release Program.

2.  Read each of the references to become familiar with the accepted
    methodologies for the control and release of effluents.

3.  Go to the ROP web-site and review any 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:** Health Physics Proficiency Level Qualification
Signature Card Item ISA-HP-11

**REFERENCES FOR ISA-HP-11**

1\. NRC IMC 0609, ASignificance Determination Process@, Appendix D,
Effluent Release Program Branch.

2.  10 CFR Part 50, Criterion 60, Control of releases of radioactive
    materials to the environment.

3.  10 CFR Part 50, Criterion 61, Fuel storage and handling and
    radioactivity control.

4.  10 CFR Part 50, Criterion 63, Monitoring fuel and waste storage.

5.  10 CFR Part 50, Criterion 64, Monitoring radioactivity releases.

6.  10 CFR Part 20, Standards for Protection Against Radiation.

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p.  Subpart D - Radiation Dose Limits for Individual Members of the
    Public.

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7.  Appendix I to 10 CFR Part 50, Numerical Guides For Design Objectives
    And Limiting Conditions For Operation To Meet The Criterion AAs Low
    As Is Reasonably Achievable@ For Radioactive Material In
    Light-Water-Cooled Nuclear Power Reactor Effluents.

8.  NUREG/CR-5569, Revision 1, Health Physics Positions Data Base.

> a\. Health Physics Position (HPPOS) numbers; 4, 6, 7, 8, 40, 88, 102,
> 122, 170, 171, 212, 223, 229, 326.

9.  NUREG-0543, Methods for Demonstrating Compliance With the EPA
    Uranium Fuel Cycle Standard (40 CFR Part 190).

10. NUREG-0133, Preparation of Radiological Effluent Technical
    Specifications For Nuclear Power Plants.

11. NUREG-1301, Offsite Dose Calculation Manual Guidance: Standard
    Radiological Effluent Controls for Pressurized Water Reactors.

12. NUREG-1302, Offsite Dose Calculation Manual Guidance: Standard
    Radiological Effluent Controls for Boiling Water Reactors.

13. 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

14. Safety Guide 23, Onsite Meteorological Programs

15. 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

16. Regulatory Guide 1.143, Design Guidance for Radioactive Waste
    Management Systems, Structures, and Components Installed in
    Light-Water-Cooled Nuclear Power Plants.

17. Regulatory Guide 4.15, Quality Assurance for Radiological Monitoring
    Programs (Normal Operations) - Effluent Streams and the Environment.

18. Regulatory Guide 1.110, Cost-Benefit Analysis for Radwaste Systems
    for Light-Water-Cooled Reactors.

19. Regulatory Guide 1.111, Methods for Estimating Atmospheric Transport
    and Dispersion of Gaseous Effluents in Routine Releases from
    Light-Water-Cooled Reactors.

20. Regulatory Guide 1.112, Calculation of Releases of Radioactive
    Materials in Gaseous and Liquid Effluents from Light-Water-Cooled
    Power Reactors.

21. Regulatory Guide 1.113, Estimating Aquatic Dispersion of Effluents
    from Accidental and Routine Reactor Releases for the Purpose of
    Implementing Appendix I.

22. NRC Information Notice 80-10, Contamination of Nonradioactive
    Systems and Resulting Potential for Unmonitored, Uncontrolled
    Release to the Environment.

23. NRC Information Notice 79-21, Prevention of Unplanned Releases of
    Radioactivity.

24. NRC Information Notice 91-56, Potential Radioactive Leakage to Tank
    Vented to Atmosphere.

25. Generic Letter 79-03 and Generic Letter 79-06, Offsite Dose
    Calculation Manual.

26. Generic Letter 84-12, Compliance with 10 CFR Part 61 and
    Implementation of Radioactive Effluent Technical Specifications and
    Attendant Process Control Program.

27. Generic Letter 89-01, Implementation of Programmatic and Procedural
    Controls for Radiological Effluent Technical Specifications.

Health Physics Inspector Individual Study Activity

**TOPIC**: (ISA-HP-12) Significance Determination Process - Public
Radiation Safety: Environmental Monitoring Program

**PURPOSE**: The risk significance of inspection findings are evaluated
using the significance determination process (SDP). The purpose of this
training is familiarize you with the Environmental Monitoring Program
portion of the Public Radiation Safety 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.

**COMPETENCY**

**AREAS:** INSPECTION

REGULATORY FRAMEWORK

TECHNICAL AREA EXPERTISE

**LEVEL**

**OF EFFORT**: 24 hours

**REFERENCES:** 1. NRC IMC 0609, ASignificance Determination Process@,
Appendix D, Environmental Monitoring Program.

2.  10 CFR Part 50, Criterion 60, Control of releases of radioactive
    materials to the environment.

3.  10 CFR Part 50, Criterion 64, Monitoring radioactivity releases.

4.  Generic Letter 79-65, Radiological Environmental Monitoring Program
    Requirements - Enclosing Branch Technical Position, Revision 1.

5.  Generic Letter 89-01, Implementation of Programmatic and Procedural
    Controls for Radiological Effluent Technical Specifications.

6.  Generic Letter 89-01, Supplement 1, NUREG-1301 and 2, Offsite Dose
    Calculation Manual Guidance: Standard Radiological Effluent Controls
    for PWRs and BWRs.

7.  NUREG-1302, Offsite Dose Calculation Manual Guidance: Standard
    Radiological Effluent Controls for Boiling Water Reactors.

8.  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.

9.  Regulatory Guide 4.1, Programs for Monitoring Radioactivity in the
    Environs of Nuclear Power Plants.

10. Regulatory Guide 4.8, Environmental Technical Specifications for
    Nuclear Power Plants. Also see its revision, Branch Technical
    Position, Revision 1, November 1979.

11. Regulatory Guide 4.15, Quality Assurance for Radiological Monitoring
    Programs (Normal Operations) - Effluent Streams and the Environment.

**EVALUATION**

**CRITERIA:** At the completion of this activity, you should be able to:

1.  Explain the purpose, objectives and applicability of the
    Environmental Monitoring program.

2.  Be able to 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 link between the radiological effluent monitoring
    program and the radiological environmental monitoring program.

> 5\. Describe the different environmental pathways the program is
> designed to assess.
>
> 6\. Describe the significance of the pathways monitored by the
> radiological environmental monitoring program.

7.  Describe the minimum detection sensitivity criteria that licensees
    must use to analyze their environmental samples.

8.  Describe the process a licensee needs to do to make changes to the
    scope of their radiological environmental monitoring program.

**TASKS:** 1. Read MC 0609D, Environmental Monitoring Program.

2.  Read the reference materials to become familiar with the
    characteristics of an acceptable environmental monitoring program.

3.  Go to the ROP web-site and review any Green and White findings in
    the area of public radiation safety.

> 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.

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:** Health Physics Proficiency Level Qualification
Signature Card Item ISA-HP-12

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Health Physics Inspector Individual Study Activity

**TOPIC**: (ISA-HP-13) Significance Determination Process - Public
Radiation Safety: Transportation Branch

**PURPOSE**: The risk significance of inspection findings are evaluated
using the significance determination process (SDP). The purpose of this
activity is to familiarize you with the transportation branch of the
Public Radiation SDP, explain the bases of the branch, provide technical
and policy references, and present practical exercises/scenarios that
will require you to use the this branch of the SDP.

**COMPETENCY**

**AREAS**: INSPECTION

REGULATORY FRAMEWORK

TECHNICAL AREA EXPERTISE

**LEVEL**

**OF EFFORT**: 36 hours

**REFERENCES**: 1. IMC 0609, ASignificance Determination Process@,
Appendix D.

> 2\. Transportation SDP Branch

3.  NRC Inspection Report, and NRC Responses to Licensee Contestation
    for White Findings

4.  NUREG-1600, AU.S.-Specific Schedules of Requirements for Transport
    of Specified Types of Radioactive Material Consignments, 11/98.

5.  NUREG-1608, ACategorizing and Transporting Low Specific Activity
    Materials and Surface Contaminated Objects@, 7/98

6.  NUREG/CR-6407, AClassification of Transportation Packing and Dry
    Spent Fuel Storage System Components According to Importance to
    Safety@, 2/96.

7.  NUREG/CR-5569, AHealth Physics Positions Data Base@, Section 2.17
    (Transportation and Shipping), 2/94.

8.  NUREG/CR-6204, AQuestions and Answers Based on Revised 10 CFR Part
    20, Sections 20.1904 and .1906, pp 50-51, 5/94.

9.  10 CFR Part 20, Standards for Protection Against Radiation, Appendix

10. 10 CFR Part 61, Licensing Requirements for Land Disposal of
    Radioactive Wastes.

11. 10 CFR Part 71, Packaging and Transportation of Radioactive Material

12. 49 CFR Parts 171-178, Subchapter C, Hazardous Material Regulations

13. Radiation Limits SDP Sub-Branch (No References Given)

14. Package Breach SDP Sub-Branch (No References Given)

15. Certificate of Compliance SDP Sub-Branch

> a\. NUREG-0383, Volume 2, Rev. 23 (latest version), ADirectory of
> Certificates of Compliance for Radioactive Materials Packages@.

16. Low Level Burial Ground SDP Sub-Branch (No References Given)

17. Notifications or Emergency Information SDP Sub-Branch

> a\. NRC Information Notice 92B62: Emergency Response Information
> Requirements for Radioactive Material Shipments**;** August 24, 1992.
>
> b\. NRC Information Notice 93-07: CLASSIFICATION OF TRANSPORTATION
> EMERGENCIES; February 1, 1993.

**EVALUATION**

**CRITERIA**: At the completion of this activity, you 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.

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.

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4.  For the Breach of Package Sub-Branch of the SDP you should be able
    to:

> a\. Define Abreach@, Anormal conditions of transport@, and AType A@
> and how these terms factor in when determining whether a loss of
> containment of a Type A package is deemed a breach finding, for
> purposes of this SDP.

b.  Define Abreach@, Abeyond normal conditions of transport@, and AType
    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 a YELLOW finding is made by the NRC when a
> State licensed low-level burial ground bans a power plant (will not
> accept shipments). Explain how the NRC issuance of this finding might
> be related to maintaining openness.
>
> b\. Discuss and explain why waste under-classification merits a WHITE
> finding.

c.  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:

l.  For Block N1, give examples of non-compliances that result in GREEN
    and WHITE findings.

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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.

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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:

l.  Define what a Certificate of Compliance (COC) is, what it does and
    discuss typical COC requirements (or conditions) for a Type B
    shipping container.

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a.  Process findings through the branch, using example training
    scenarios

**TASKS** 1. Read MC 0609 and focus on Appendix C.

2.  Read the references pertinent to each SDP sub-branch.

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
    contents) that would result in only a GREEN finding.

5.  Read 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.

2.  Read 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.

3.  Read 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.

4.  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 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 appropriate reference NUREG, sort and list six components
> relative to their importance to safety.
>
> 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.

**Documentation**: Health Physics Proficiency Level Signature Card Item
ISA-HP-13

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Health Physics Inspector Individual Study Activity

**TOPIC:** (ISA-HP-14) Performance Indicator - Occupational Radiation
Safety

**PURPOSE:** 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. This activity
will provide you with a detailed knowledge of the definition of an
Occupational Radiation Safety PI, the ability to identify an individual
operational occurrence, and correctly classify the occurrence as a PI,
or not.

**COMPETENCY**

**AREA:** INSPECTION

REGULATORY FRAMEWORK

TECHNICAL AREA EXPERTISE

**LEVEL OF**

**EFFORT:** 18 Hours

**REFERENCES:** 1. NEI 99-02 Rev. 2, [Regulatory Assessment Performance
Indicator Guideline]{.underline}, Section 2.5.

> 2\. NEI 99-02 Rev. 0, [Regulatory Assessment Performance Indicator
> Guideline]{.underline}, Frequently Asked Questions, pages 112 - 120,
> or
>
> NRC Web site for Archived Occupational FAQs:
> NRR/OVERSIGHT/ASSESS/archived faqs

3.  Standard Technical Specification 6.12, High Radiation Area Access
    Control.

4.  

5.  Regulatory Guide 8.38, [Control Of Access To High And Very High
    Radiation Areas In Nuclear Power Plants]{.underline} .

> 6\. NUREG/CR-6204, [Question and Answers Based on Revised
> 10 CFR 20]{.underline}, Q&As Nos. 49,92, 218,373,385,,423,
> 441,447,448,487,488, and 489 .
>
> 7\. NUREG/CR-5569, Rev., [1 Health Physics Positions Data
> Base]{.underline}, Section 2.5, pages 62 to 68.
>
> 8\. Manual Chapter 0608, Performance Indicator Program
>
> 9\. NRR Radiation Protection Technical Readings on PIs

**EVALUATION**

**CRITERIA:** At the completion of this activity, you should be able to:

14980. Identify which attributes of a Radiation Protection Program are
       covered by the PI in this area.

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9.  Discuss the several types of operational occurrences included in the
    PI definition, including which non-conformance with High Radiation
    Area Technical Specifications are pertinent.

10. Define Aunintended occupational exposure@ and Aradiation safety
    barriers@.

11. Distinguish between an individual occurrence and concurrent
    operational occurrences.

12. Discuss the minimum significance Athresholds@ of an occurrence in
    the PI definition.

13. Discuss the Athresholds@ associated with the colored PI significance
    bands.

14. Discuss how Asignificant@ 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.

3.  Review the materials in References 5 and 6.

4.  Review the discussion on Aaccessible@ areas and adequate controls
    for High and Very High Radiation Areas in Reference 4, above.

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:** Health Physics Proficiency Level Qualification
Signature Card Item

ISA-HP-14

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**HEALTH PHYSICS INSPECTOR**

**ON-THE-JOB ACTIVITIES**

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Health Physics Inspector On-the-Job Activity

**TOPIC:** (OJT-HP-1) Inspecting Access Controls to Radiologically
Significant Areas

**PURPOSE:** The purpose of this activity is to familiarize you with
inspection activities in inspection procedure 71121.01 AAccess Controls
to Radiologically Significant Areas@. The objectives of this procedure
are:

-   To review and assess licensee=s performance in implementing physical
    and administrative controls for airborne radioactivity areas,
    radiation areas, high radiation areas (HRAs), and worker adherence
    to these controls.

-   To observe access controls to radiation and high radiation areas
    \<1000 mrem/hr and areas that are subject to transient dose rates.
    Review the controls that have been established and confirm that
    workers follow established rules. Review the high radiation area
    incidents found in the performance indicators (PIs) and in the
    licensee=s corrective action program during the current assessment
    period.

-   To verify that the licensee is maintaining adequate controls over
    HRAs (with dose rates greater than 25 rem/h) and all very high
    radiation areas (VHRA).

**COMPETENCY**

**AREA:** TECHNICAL AREA EXPERTISE

ASSESSMENT AND ENFORCEMENT

COMMUNICATION

INSPECTION

**LEVEL OF**

**EFFORT:** 32 hours

**REFERENCES:** All the references used in ISA-HP-3 are appropriate for
this module. Most of the efforts in this OJT will rely on licensee
procedures and related regulatory documents (Technical Specifications,
FSAR, etc.). Any references used other than licensee procedures will be
used to determine the regulatory stance that has been historically
applied by the NRC for a given situation. References selected should
support the actual inspection effort.

> Licensee ARadiation Protection Plan@
>
> Licensee ARadiological Access Control@ procedure
>
> Licensee ARadiological Postings@ procedure
>
> Licensee ARadiation Work Permit@ procedure
>
> Licensee Implementing Procedures for High and Very High
>
> Radiation Areas
>
> Licensee ATechnical Specifications@
>
> (Administrative controls for high radiation areas)
>
> Licensee AFinal Safety Analysis Report or Updated Safety Analysis
> Report@
>
> Davis-Besse Special Inspection Reports 02-06 and 02-16, ADAMS
> Accession \# ML030070606

**EVALUATION**

**CRITERIA**: Upon completion of the tasks in this OJT, you should be
able to:

61560. Describe areas that are considered risk significant and what kind
       of work would increase the risk significance.

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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 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.

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. 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.

**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 instrument on and check batteries, check high voltage or
> instrument zero if applicable, physical condition and perform a
> response test with appropriate source.
>
> d\. Verify that 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
    areas, locked high radiation areas. Review licensee procedures for
    very high radiation areas, and identify the Aadditional measures@
    implemented, as required by 10 CFR20.1602.

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 surface- disturbing
    work. (A qualified inspector can identify an example physical
    setting and work through this with you.)

9.  Review plant procedures for internal dose assessment. Walk the
    procedure through for a hypothetical uptake that involves a mixture
    of activation products, fission products and transuranics. (Examples
    could include but are not limited to: handling of dried purification
    filters, surface disturbing work such as grinding, or eddy current
    testing of a letdown heat exchanger.)

10. 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.

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13. Perform a physical inspection of licensee spent fuel pool/transfer
    canal or moisture separator pit and determine licensee controls for
    high activity items stored in pool. Determine what controls are in
    place to prevent accidentally exposing high activity items.

> Note: some licensees do not do long-term storage of items suspended
> from side of pool in which case determine the controls that they would
> use for short term transient storage prior to disposition.

14. Review a licensee 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.

15. 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**: Health Physics Inspector Proficiency Level Signature
Card Item OJT-HP-1

**Additional References Related to Access Control**

**to Radiologically Significant Areas:**

IE Circular 76-03, \"Radiation Exposures in Reactor Cavities@

IE Bulletin 78-08, ARadiation Levels From Fuel Element Transfer Tubes@

IN 80-22, \"Breakdowns in Contamination Control Programs@

IN 82-31, \"Overexposure of Diver During Work in Fuel Storage Pool@

IN 84-82, AGuidance For Posting Radiation Areas@

IN 85-06, AContamination of Breathing Air Systems@

IN 86-44, AFailure to Follow Procedures when Working in High Radiation
Areas@

IN 88-63, \"High Radiation Hazards from Irradiated Incore Detectors and
Cables@

IN 88-79, AMisuse of Flashing Lights for High Radiation Area Controls@

IN 90-33, ASources of Unexpected Occupational Radiation Exposures at
Spent Fuel Storage Pools@

IN 90-44 ADose‑Rate Instruments Under responding to the True Radiation
Fields@

IN 90-47, AUnplanned Radiation Exposures to Personnel Extremities Due to
Improper Handling of Potentially Highly Radioactive Source@

IN 92-75, AUnplanned Intakes of Airborne Radioactive Material By
Individuals at Nuclear Power Plants@

IN 97-36, AUnplanned Intakes by Worker of Transuranic Airborne
Radioactive Materials And External Exposure Due to Inadequate Control of
Work@

NUREG/CR 5569, Rev.1 Health Physics Positions (HPPOS 002,
010,014,015,016, 036,055, 066, 068, 128, 138, 180, 210, 234, 236, 237,
242, 244, 245, 323 and 328)

NUREG/CR-6204, Questions and Answers based on Revised 10 CFR 20
(Questions 27,49,52,57,92,132,373,385,423,447, 448, 487, 489)

Health Physics Inspector On-the-Job Activity

**TOPIC**: (OJT-HP-2) Inspecting ALARA Planning and Controls

**PURPOSE**: The purpose of this activity is to familiarize you the
inspection activities in inspection procedure 71121.02 AALARA Planning
and Controls@. The objective of this procedure is to assess performance
with respect to maintaining individual and collective radiation
exposures as low as is reasonably achievable. This 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

ASSESSMENT AND ENFORCEMENT

COMMUNICATION

INSPECTION

**LEVEL**

**OF EFFORT:** 36 hours

**REFERENCES**: All the references used in ISA-HP-4 are appropriate for
this module. Most of the efforts in this OJT will rely on licensee
procedures and related regulatory documents (Technical Specifications,
FSAR, etc..). Any references used other than licensee procedures will be
used to determine the regulatory stance that has been historically
applied by the NRC for a given situation. References selected should
support the actual inspection effort.

1.  10 CFR 20

2.  Licensee dose printout of individual doses for a work group
    identified by qualified inspector

3.  Licensee source term trending documentation/ ALARA review.

4.  Licensee source term reduction program procedures

5.  Completed licensee ALARA package for a significant job identified by
    qualified inspector.

6.  Licensee historical or current outage backbone schedule that
    identifies significant work milestones.

7.  Licensee temporary shielding procedure

8.  Licensee hot spot reduction procedure

9.  Licensee source term control/ reduction procedure ( May be
    fragmented into several procedures)

10. Licensee AStrategic Primary Water Plan@ section describing shutdown
    chemistry or other procedure describing the chemistry controls
    involved in shutting down the plant. (The AStrategic Water Plans@
    are licensee process optimization documents that are generated in
    response to EPRI and NEI initiatives/commitments.)

11. Licensee ADeclared Pregnant Woman@ Procedure

**EVALUATION**

**CRITERIA:** Upon completion of the tasks in this OJT, you should be
able to:

64140. Explain why engineering controls are used to reduce or eliminate
       respiratory protection requirements.

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1.  Explain why there are still some variations in the amount of dose
    received for a population of workers even though skill levels are
    same and work assignments are comparable.

2.  Describe licensee source term trending techniques.

3.  Describe key elements of a source term reduction program and
    identify any initiatives that are being missed.

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.

9.  Describe the process used to determine and assign fetal dose.
    Describe a woman=s rights with respect to declaring her pregnancy.

TASKS: 1. Estimate dose reductions achieved by use of engineering
controls.

t.  Reduction in TEDE (Total Effective Dose Equivalent) dose by avoiding
    having respiratory protection

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a.  Reduction in TEDE dose that would have been incurred had the
    licensee used neither engineering controls nor respiratory
    protection.

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2.  Compare individual exposures for a selected workgroup and identify
    probable reasons for significant differences. Note: 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 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, shielding,
    filled system vs dry. 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.

7.  Review 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 lesson-learned
    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 Declared Pregnant Worker (or Woman) procedures. Review the
    exposure results and monitoring controls employed by licensee with
    respect to the requirements. (Program should have provisions for
    determining fetal dose from internal deposition as well as external
    radiation.)

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:** Health Physics Inspector Proficiency Level
Qualification Signature Card Item OJT-HP-2

Health Physics Inspector On-the-Job Activity

**TOPIC:** (OJT-HP-3) Inspecting Radiation Monitoring Instrumentation

**PURPOSE:** The purpose of this activity is to familiarize you with the
inspection activities in inspection procedure 71121.03, ARadiation
Monitoring Instrumentation@. The objectives of this procedure are: (1)
To determine the accuracy and operability of radiation monitoring
instruments that are used for the protection of occupational workers and
(2) To determine the adequacy of program to provide self-contained
breathing apparatus (SCBA) for entering and working in areas of unknown
radiological and/or potential immediately dangerous to life and health
(IDLH) areas.

**COMPETENCY**

**AREA**: TECHNICAL AREA EXPERTISE

INSPECTION

**LEVEL OF**

**EFFORT:** 36 hours

**REFERENCES:** All the references used in ISA-HP-5 are appropriate for
this module. Most of the efforts in this OJT will rely on licensee
procedures and related regulatory documents (Technical Specifications,
FSAR, etc..). Any references used other than licensee procedures will be
used to determine the regulatory stance that has been historically
applied by the NRC for a given situation. References selected should
support the actual inspection effort.

1.  Licensee fixed and portable instrumentation procedures

2.  Licensee SCBA training and maintenance procedures

3.  Licensee whole body counting and internal dose assessment
    procedures.

**EVALUATION**

**CRITERIA:** Upon Completion of this activity you should be able to:

12976. Describe the extent that emergency planning and maintenance rule
       inspections (in your region) address radiological instruments
       well enough to preclude repeating inspection activities while
       ensuring adequate inspection coverage.

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1.  Describe the plant activities that have permanently installed area
    radiation monitors based on examination of the plants Final Safety
    Analysis Report (FSAR).

2.  Describe the instrumentation used for high risk jobs including
    survey meter types and designations, remote readout area monitors
    and continuous air monitors.

3.  Describe the fixed and portable instrumentation used for personnel
    release from radiologically controlled area and how it is used.

4.  Describe the methodology used to calibrate instruments, verify their
    operability and determine if appropriate calibration sources are
    used. alarm set points if applicable.

5.  Describe the process used to resolve situations where instruments
    are found to be significantly out of calibration. What is the impact
    on surveys done with the instrument?

6.  Describe the likely consequences of having a substantially over or
    under responding instrument used to cover work and perform surveys.
    What would be the effect of significantly extending the source check
    frequencies of survey instruments?

7.  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.

8.  If the plants have individuals with internal exposures greater than
    100 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.

9.  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.

10. Describe the training program for SCBA users.

11. Describe the licensee SCBA air cylinder program for air quality
    testing hydro-testing, and replacement schedule.

**TASKS:**

13660. 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 for licensee site).

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1.  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.

2.  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 associated with jobs with the potential for workers to
    receive 100 mrem CEDE.

3.  Identify types of radiation detection instruments utilized for
    personnel release from the radiologically controlled area.

4.  Verify Calibration, Operability, and Alarm Set point (if applicable)
    of Types of Several Instruments and Equipment.

5.  Determine what actions are taken when, during calibration or source
    checks, an instrument is found significantly out of calibration
    (\>50%).

6.  Determine possible consequences of instrument use since last
    successful calibration or source check.

7.  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.

8.  Evaluate licensee internal exposures, if there are any internal
    exposures \>100 mrem CEDE, determine whether the affected personnel
    were properly measured utilizing calibrated equipment, data analyzed
    and internal exposures properly assessed in accordance with licensee
    procedures.

9.  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.

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14008. 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.

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1.  Review pertinent sections of 29 CFR 1910 and 49 CFR part 173 and
    part 178 for additional respirator and cylinder maintenance
    requirements.

2.  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:** Health Physics Inspector Proficiency Level
Qualification Signature Card Item OJT-HP-3.

**Additional references related to Radiation Monitoring
Instrumentation**

**Instruments**

10 CFR 20.1501

NUREG-1736, AConsolidated Guidance: 10 CFR 20 - Standards for Protection
Against Radiation@

Regulatory Guide 8.6, AStandard Test Procedure for Geiger-Muller
Counters@

IE-Bulletin 97-001, APotential for Erroneous Calibration, Dose Rate, or
Radiation Exposure Measurements with Certain Victoreen Model 530 and
53OSI Electrometer/Dose-Meters@

HPPOS-1, AProposed Guidance for Calibration and Surveillance
Requirements to Meet Item II.F.1 of NUREG-0737

HPPOS-88, ACorrections for Sample Conditions for Air and Gas Monitoring@

HPPOS-279, \" Technical Assistance Request Regarding Electronic
Calibration of Survey Instruments@

HPPOS-328, AProper Operation and Use of Alarm Dosimeters at Nuclear
Power Plants@

IE Notice 93-30, A NRC Requirements for Evaluation of Wipe Test Results;
Calibration of Count Rate Survey Instruments@

Q&A 147, ACalibration Frequency@

Q&A 209, ACalibration Frequency@

**Internal Exposure**

10 CFR 20.1202 -1204(h)

10 CFR 20.1700, ASubpart H@

NUREG-1400, AAir Sampling in the Workplace@

NUREG/CR-0041, Rev. 1, AManual of Respiratory Protection Against
Airborne Radioactive Materials@

> Regulatory Guide 8.9, AAcceptable Concepts, Models, Equations, and
> Assumptions for Bioassay Program (Revision 1)

Regulatory Guide 8.15, AAcceptable Programs for Respiratory Protection
(Revision 1)

**Self Contained Breathing Apparatus (SCBA)**

IE Circular 80-03, AProtection From Toxic Gas Hazards@

IE Notice 81-26, Part 4 APersonnel Entry Into Inerted Containment@

IE Notice 85-87, AHazards of Inerting Atmospheres@

IE Notice 86-46, AImproper Cleaning and Decontamination of Respiratory
Protection Equipment@

NRC Information Notice 98-20, AProblems with Emergency Preparedness
Respiratory Protection Programs@

HPPOS-147, ARespirator User=s Notice- Use of Unapproved Subassemblies
(NIOSH warns against the use of unapproved subassemblies (parts and
components) and unauthorized modification for/of approved respirators)

IE Notice 83-68, ARespirator User Warning: Defective Self-Contained
Breathing Apparatus Air Cylinders@

IE Notice 85-48, ARespirator User Warning: Defective Self-Contained
Breathing Apparatus Air Cylinders@

IE Notice 86-103, ARespirator Coupling Nut Assembly Failures@

Information Notice 89-47, APotential Problems with Worn or Distorted
Hose Clamps on Self-Contained Breathing Apparatus@

Information Notice 94-35, ANIOSH Respirator User Notices,@Inadvertent
Separation of the Mask-Mounted Regulator (MMR) from the Facepiece of the
Mine Safety Appliances (MSA) Company Self Contained Breathing
Apparatus(SCBA) and Status Update@

Information Notice 95-01, ADOT Safety Advisory: High Pressure Aluminum
Seamless and Aluminum Composite Hoop-Wrapped Cylinders@

Information Notice 97-66, AFailure to Provide Special Lenses for
Operators Using Respirator or SCBA During Emergency Operations@

Health Physics Inspector On-the-Job Activity

**TOPIC:** (OJT-HP-4) Radioactive Material Processing 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 7112202, Radioactive Material
Processing and Transportation, to allow you to independently conduct
inspections in this area. 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 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 programs.

**COMPETENCY**

**AREA:** INSPECTION

TECHNICAL AREA EXPERTISE

**LEVEL**

**OF EFFORT:** 40 HOURS

**REFERENCES:** Title 10, Code of Federal Regulations, Parts 20, 61, and
71

> Title 49, Code of Federal Regulations, Parts 100-189

Plant Updated Final Safety Analysis Report

Plant Annual Effluent Release Report

Offsite Dose Calculation Manual

> Inspection Procedure 71122, APublic Radiation Safety@

NRC Branch Technical Position, Waste Form Technical Position

> NRC Branch Technical Position on Concentration Averaging and
> Encapsulation
>
> NUREG - 1608, Categorizing and Transporting Low Specific Activity
> Materials and Surface Contaminated Objects
>
> NUREG-1660, U.S. - Specific Schedules of Requirements for Transport of
> Specified Types of radioactive material Consignments

**EVALUATION**

**CRITERIA:** Upon completion of the tasks, you 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
> 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 CFR61.55.

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.

10. For the same shipments identified in item 10, 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 CFR61
    and branch technical positions.

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3.  Discuss how the licensee maintains its regulatory documents, for its
    radioactive material processing and transportation program,
    up-to-date.

4.  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 does provide.

2.  Locate a copy of the radwaste section from the UFSAR. Conduct walk
    downs (considering ALARA and safety constraints) of as much of the
    station=s liquid and solid radwaste processing systems during your
    OJT activity. Assess the licensee=s conformance of the facility with
    its design documents including any 10 CFR50.59 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 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.

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 CFR20 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.

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 (preferability a non-exempt
    shipment). 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.

**DOCUMENTATION:** Health Physics Proficiency level Signature Card item
OJT-HP-4

**Other Important References Related to radioactive Material Processing
and Transportation**

Information Notice 86-20, Technical Position on Concentration Averaging

Information Notice 90-50, Minimization of Methane Gas in Plant Systems
and Radwaste Shipping Containers

Information Notice 92-62, Emergency Response Information Requirements
for Radioactive Material Shipments

Information Notice 92-72, Employee Training and Shipper Registration
Requirements for Transporting Radioactive Materials

Information Notice 95-09, Monitoring and Training of Shippers and
carriers of Radioactive Materials

NUREG/BR-0204, Instructions for Completing NRC=s Uniform Low-level
Radioactive Waste Manifest

NRC Generic Letter 81-38, Storage of Low-Level radioactive Wastes at
Power Reactor Sites

NRC Bulletin 80-10, Contamination of Non-Radioactive System and
Resulting Potential for Unmonitored, Uncontrolled Release to Environment

ANSI/ANS-40.37-1993, Mobile Radioactive Waste Processing Systems

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

NUREG/CR-6204, Questions and Answers Based on Revised 10 CFR Part 20

NUREG 1736, Consolidated Guidance: 10 CFR Part 20- Standards for
Protection Against Radiation

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Health Physics Inspector On-the-Job Activity

**TOPIC:** (OJT-HP-5) Radiological Environmental Monitoring Program
(REMP) and Radioactive Material Control 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 IP71122.03, Radiological
Environmental Monitoring Program and Radioactive Material Control
Program, to allow you to independently conduct inspections in these
areas. 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
CFR 50.34a.

> The radioactive material control program verifies that the licensee
> maintains a program to ensure that licensed radioactive material is
> controlled in accordance with the requirements of 10 CFR Part 20.
>
> Upon completion of this on-the-job activity, you will be able to
> perform assessments of reactor licensee=s Radiological Environmental
> Monitoring Program (REMP) and Radioactive Material Control Program.

**COMPETENCY**

**AREA:** INSPECTION

**LEVEL OF**

**EFFORT:** 40 HOURS

**REFERENCES:** Title 10, Code of Federal Regulations, Parts 20 and 50

> Plant Updated Final Safety Analysis Report
>
> IE Circular 81-07 and IE Information Notice 85-92
>
> Regulatory Guide 1.23, \@On-site Meteorological Program@
>
> Regulatory Guide 1.111, AMethods of Estimating Atmospheric Transport
> and Dispersion of Gaseous Effluents in Routine Release from
> Light-Water-Cooled Reactors@
>
> Regulatory Guide 4.1, APrograms for Monitoring Radioactivity in the
> Environs of Nuclear Power Plants@
>
> Regulatory Guide 4.13, APerformance, Testing, and Procedural
> Specifications for Thermoluminescence Dosimetry: Environmental
> Application@
>
> Regulatory Guide 4.15, AQuality Assurance for Radiological Monitoring
> Programs (Normal Operations) BEffluent Streams and the Environment@

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

> NUREG/CR-6204, Questions and Answers Based on Revised 10 CFR Part 20
>
> NUREG 1736, Consolidated Guidance: 10 CFR Part 20- Standards for
> Protection Against Radiation

**EVALUATION**

**CRITERIA:** Upon completion of the tasks, you 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 the different measuring techniques (proportional counter,
    gamma spectroscopy, liquid scintillation counter, and TLD reader)
    that you inspected. Compare and contrast the methods.

3.  Discuss data reduction techniques used by the licensee, including
    MDA and LLD.

4.  Discuss the Laboratory=s QA Policy and QC implementation, including
    inter-laboratory and intra-laboratory comparisons.

5.  Discuss the Land Use Census. Identify principle land uses.

6.  Discuss the Meteorological Monitoring Program, including calibration
    methodology for wind direction, wind speed, and delta temperature.

7.  Discuss the licensee=s determination of χ/Q and D/Q, and annual
    average data.

8.  Review the REMP Annual Report and discuss the soundness of the
    report.

9.  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 IE Information Notice 85-92.

**TASKS**:

1.  Review and become familiar with the documents listed in the
    reference list.

2.  Locate a copy of the REMP section from the UFSAR and ODCM.

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 the radiological analytical methodology for
    tritium, gamma, ambient radiation (using TLD), and gross alpha and
    beta. Compare the licensee=s technique to standard analytical
    techniques.

8.  Review and evaluate the meteorological monitoring requirements and
    evaluate the adequacy of the calibration results for wind direction,
    wind speed, and delta temperature.

9.  Conduct walk downs of as much of the REMP sampling stations as
    possible, including the meteorological monitoring tower and the
    control room. Compare this with that describe, as applicable, in the
    UFSAR, REMP or ODCM.

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2.  Review the sequential relationship between Radiological Effluent
    Controls and the licensee=s REMP.

3.  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.

4.  Review the Fundamentals of Laboratory QA/QC Policy and its
    implementation at the licensee=s facility.

5.  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.

**DOCUMENTATION**: Health Physics Proficiency Level Signature Card Item
OJT-HP-5

**Other Important References Related to REMP and Radioactive Materials
Control**

Environmental Measurement Laboratory, HASL-300 Procedures Manual U.S.
Department of Energy, New York, NY

NCRP Report No. 45, ANatural Background Radiation in the US@

NCRP Report No. 47, ATritium measurement Techniques@

NCRP Report No. 50, AEnvironmental Radiation Measurements@

NCRP Report No. 58, AA Handbook of Radioactivity Measurements
Procedures@

ANSI N545-1975, APerformance, Testing, and Procedural Specification for
TLD, Environmental Application@

ANSI N13.1-1969, AGuide to Sampling Airborne Radioactive materials in
Nuclear Facilities@

ANSI N13.4-1971, AAmerican National Standard for the Specification of
Portable X- or Gamma-Radiation Survey Instruments@

Inspection Procedure 71122,03 APublic Radiation Safety

Health Physics Inspector On-the-Job Activity

**TOPIC:** (OJT-HP-6) Radioactive Gaseous and Liquid Effluent Treatment
and Monitoring Systems

**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 7112201, Radioactive Gaseous and
Liquid Effluent Treatment and Monitoring Systems, to allow you to
independently conduct inspections in this area. 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 Adefense-in-depth@ against an
unmonitored, unanticipated, and unplanned release of radioactive
material to the environment.

> 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:** Title 10, Code of Federal Regulations, Parts 20 and 50

> Title 40, Code of Federal Regulations, Part 190
>
> Plant Updated Final Safety Analysis Report
>
> Inspection Procedure 71122, APublic Radiation Safety@
>
> NUREG/CR-5569, Rev. 1, Health Physics Positions Data Base
>
> NUREG/CR-6204, Questions and Answers Based on Revised 10 CFR Part 20
>
> NUREG 1736, Consolidated Guidance: 10 CFR Part 20- Standards for
> Protection Against Radiation

**EVALUATION**

**CRITERIA:** Upon completion of the tasks, you 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 different measuring techniques (proportional counter,
    gamma spectroscopy, and liquid scintillation counter) that you
    inspected and compare and contrast their capabilities and
    limitations.

3.  Discuss the Laboratory=s QA Policy and QC implementation, including
    inter-laboratory and intra-laboratory comparison. Compare it with
    NRC expectations.

4.  Discuss data reduction techniques, including MDA and LLD. Verify
    licensee MDA and LLD calculations for each release pathway.

5.  Discuss the radioactive liquid treatment systems and the effluent
    release pathways. Compare these to those presented in the UFSAR and
    ODCM

6.  Discuss the radioactive gas treatment systems and the gaseous
    effluent release pathways. Compare these to those presented in the
    UFSAR and ODCM.

7.  Discuss the radioactive liquid effluent radiation monitoring systems
    listed in the ODCM. Discuss the capabilities of the monitors and
    their conformance with UFSAR and ODCM descriptions.

8.  Discuss the radioactive gaseous effluent radiation monitoring
    systems listed in the ODCM. Discuss the capabilities of the monitors
    and their conformance with UFSAR and ODCM descriptions.

9.  Discuss the air cleaning systems and their functions (e.g.,
    Augmented Off-Gas, Reactor Building Air Cleaning System.)

10. 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.

11. Discuss the calibration techniques for the effluent flow rate
    measurement devices listed in the ODCM. Identify what type of flow
    the sampler is seeing show if 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
    facility as possible. Compare your observations with descriptions in
    the UFSAR

3.  Identify the various sources of liquid and gaseous radwaste, waste
    streams, and technologies associated with liquid and gaseous
    radwaste processing for the facility. Compare your observations with
    descriptions in the UFSAR

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```
4.  Review the licensee=s ODCM and identify the radioactive gaseous and
    liquid effluent pathways.

5.  Review and evaluate the projected public dose calculation
    methodologies (all pathways) listed in Regulatory Guide 1.109.
    (**\*Comment: it may take several days**.)

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```
1.  Review the licensee=s projected public dose assessment technique and
    compare its techniques against the RG 1.109.

2.  Review and evaluate the effluent radiation monitoring system channel
    calibration, functional test, and source check results. (**Comment:
    it may take several days**.)

3.  Review and evaluate the air cleaning system surveillance test
    results. (**Comment: it may take several days**.)

4.  Review and understand the calibration techniques and results for the
    measurement laboratory=s gamma spectroscopy, proportional counter,
    and liquid scintillation counter. (**Comment: it may take several
    days**.)

5.  Review and evaluate the implementation of the QA/QC in the
    measurement laboratory, including inter-laboratory and
    intra-laboratory comparisons.

6.  Review and understand the measurement laboratory=s analytical
    techniques (e.g., gamma, Sr-89/90, H-3, and gross alpha/beta).

7.  Review and understand the radioactive liquid and gaseous release
    permits.

8.  Review and understand the Annual Radioactive Effluent Release
    Report.

9.  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.

**DOCUMENTATION**: Health Physics Proficiency Level Signature Card Item
OJT-HP-6

**Other Important References Related to Effluent Treatment and
Monitoring**

Environmental Measurement Laboratory, HASL-300 Procedures Manual U.S.
Department of Energy, New York, NY

NCRP Report No. 58, AA Handbook of Radioactivity Measurements
Procedures@

RG 1.21 \"Measuring and Reporting Radioactivity in Solid Wastes and
Releases of Radioactive Materials in Liquid and Gaseous Effluents from
Light-Water-Cooled Nuclear Power Plants\"

RG 1.52 \"Design, Testing, and Maintenance Criteria for Post-Accident
Engineered Safety Feature Atmosphere Cleanup System Air Filtration and
Adsorption Units of Light-Water-Cooled Nuclear Power Plants\" (Endorsed
ANSI N509-1976 and N510-1975)

RG 1.97 \"Instrumentation for Light-Water-Cooled Nuclear Power Plants to
Assess Plant Conditions During and Following an Accident\"

RG 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\"

RG 1.111 \"Methods for Estimating Atmospheric Transport and Dispersion
of Gaseous Effluents in Routine Releases from Light-Water-Cooled
Reactors\"

RG 4.15 \"Quality Assurance for Radiological Monitoring Programs (Normal
Operations) - Effluent Streams and the Environment\"

ANSI N13.1-1982 \"Guide to Sampling Airborne Radioactive Materials in
Nuclear Facilities\"

ANSI N45.2.23-1978 \"Qualification of Quality Assurance Program Audit
Personnel for Nuclear Power Plant Personnel\"

ANSI/ANS N55.4-1979 \"Gaseous Radioactive Waste Processing Systems for
Light-Water-Cooled Reactor Plants\"

ANSI/ANS N55.6-1979 \"Liquid Radioactive Waste Processing System for
Light-Water-Cooled Reactor Plants\"

ANSI/ASME N509-1980 \"Nuclear Power Plant Air Cleaning Units and
Components\"

ANSI/ASME N510-1980 \"Testing of Nuclear Air-Cleaning Systems\"

IE Bulletin No. 80-10 \"Contamination of Nonradioactive Systems and
Resulting Potential for Unmonitored, Uncontrolled Release of
Radioactivity to the Environment,\" May 6, 1980

IE Circular No. 80-18 \"10 CFR50.59 Safety Evaluations for Changes to
Radioactive Waste Treatment Systems,\" August 22, 1980

IE Information Notice No. 82-43 \"Deficiencies in LWR Air
Filtration/Ventilation Systems,\" November 16, 1982

IE Information Notice No. 82-49 \"Correction for Sample Conditions for
Air and Gas Monitoring,\" December 16, 1982

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

PAGE INTENTIONALLY BLANK

**Health Physics Inspector Technical Proficiency Level**

**Signature Card and Certification**

+--------------------------------------------+------------+------------+
| Inspector Name:                            | Employee   | Su         |
| \_\_\_\_\_\_\_\_\_\_\_                     |            | pervisor=s |
| \_\_\_\_\_\_\_\_\_\_\_\_\_\_\_\_\_\_\_\_\_ | Initials / |            |
|                                            | Date       | Signature  |
|                                            |            | / Date     |
+--------------------------------------------+------------+------------+
| ***A. Training Courses***                  |            |            |
+--------------------------------------------+------------+------------+
| Power Plant Engineering (self-study of     |            |            |
| selected chapters)                         |            |            |
+--------------------------------------------+------------+------------+
| GE Technology (R-104B)                     |            |            |
+--------------------------------------------+------------+------------+
| Westinghouse Technology (R-104P)           |            |            |
+--------------------------------------------+------------+------------+
| Health Physics Technology (H-201)          |            |            |
+--------------------------------------------+------------+------------+
| Respiratory Protection (H-311)             |            |            |
+--------------------------------------------+------------+------------+
| Environmental Monitoring for Radioactivity |            |            |
| (H-111)                                    |            |            |
+--------------------------------------------+------------+------------+
| Radwaste Management (H-202)                |            |            |
+--------------------------------------------+------------+------------+
| Transportation of Radioactive Materials    |            |            |
| (H-308)                                    |            |            |
+--------------------------------------------+------------+------------+
| ***B. Individual Study Activities***       |            |            |
+--------------------------------------------+------------+------------+
| (ISA-HP-1) Code of Federal Regulations     |            |            |
| (CFR=s)                                    |            |            |
+--------------------------------------------+------------+------------+
| (ISA-HP-2) Licensee Documents for Health   |            |            |
| Physics Inspectors                         |            |            |
+--------------------------------------------+------------+------------+
| (ISA-HP-3) Access Controls to              |            |            |
| Radiologically Significant Areas           |            |            |
+--------------------------------------------+------------+------------+
| (ISA-HP-4) ALARA Planning and Controls     |            |            |
+--------------------------------------------+------------+------------+
| (ISA-HP-5) Radiation Monitoring            |            |            |
| Instrumentation                            |            |            |
+--------------------------------------------+------------+------------+
| (ISA-HP-6) Radioactive Material Processing |            |            |
| and Transportation                         |            |            |
+--------------------------------------------+------------+------------+
| (ISA-HP-7) Radiological Environmental      |            |            |
| Monitoring Program (REMP)                  |            |            |
|                                            |            |            |
| and Radioactive Material Control Program   |            |            |
+--------------------------------------------+------------+------------+
| (ISA-HP-8) Radioactive Gaseous and Liquid  |            |            |
| Effluent Treatment and                     |            |            |
|                                            |            |            |
| Monitoring Systems                         |            |            |
+--------------------------------------------+------------+------------+
| (ISA-HP-9) Significance Determination      |            |            |
| Process                                    |            |            |
|                                            |            |            |
| \- Occupational Radiation Safety           |            |            |
+--------------------------------------------+------------+------------+
| (ISA-HP-10) Significance Determination     |            |            |
| Process                                    |            |            |
|                                            |            |            |
| \- Public Radiation Safety: Radioactive    |            |            |
| Material Control                           |            |            |
+--------------------------------------------+------------+------------+
| (ISA-HP-11) Significance Determination     |            |            |
| Process                                    |            |            |
|                                            |            |            |
| \- Public Radiation Safety: Effluent       |            |            |
| Release Program                            |            |            |
+--------------------------------------------+------------+------------+
| (ISA-HP-12) Significance Determination     |            |            |
| Process                                    |            |            |
|                                            |            |            |
| \- Public Radiation Safety: Environmental  |            |            |
| Monitoring                                 |            |            |
|                                            |            |            |
| Program                                    |            |            |
+--------------------------------------------+------------+------------+
| (ISA-HP-13) Significance Determination     |            |            |
| Process                                    |            |            |
|                                            |            |            |
| \- Public Radiation Safety: Transportation |            |            |
| Branch                                     |            |            |
+--------------------------------------------+------------+------------+

+--------------------------------------------+------------+------------+
| (ISA-HP-14) Performance Indicator -        |            |            |
| Occupational Radiation Safety              |            |            |
+--------------------------------------------+------------+------------+
| *On-the-Job Activities*                    |            |            |
+--------------------------------------------+------------+------------+
| (OJT-HP-1) Inspecting Access Controls to   |            |            |
| Radiologically Significant                 |            |            |
|                                            |            |            |
| Areas                                      |            |            |
+--------------------------------------------+------------+------------+
| (OJT-HP-2) Inspecting ALARA Planning and   |            |            |
| Controls                                   |            |            |
+--------------------------------------------+------------+------------+
| (OJT-HP-3) Inspecting Radiation Monitoring |            |            |
| Instrumentation                            |            |            |
+--------------------------------------------+------------+------------+
| (OJT-HP-4) Radioactive Material Processing |            |            |
| and Transportation                         |            |            |
+--------------------------------------------+------------+------------+
| (OJT-HP-5) Radiological Environmental      |            |            |
| Monitoring Program (REMP)                  |            |            |
|                                            |            |            |
| and Radioactive Material Control Program   |            |            |
+--------------------------------------------+------------+------------+
| (OJT-HP-6) Radioactive Gaseous and Liquid  |            |            |
| Effluent Treatment and                     |            |            |
|                                            |            |            |
| Monitoring Systems                         |            |            |
+--------------------------------------------+------------+------------+

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

+-------------------------------------+--------------------------------+
| **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 Technology (R-104B)              |                                |
+-------------------------------------+--------------------------------+
| Westinghouse Technology (R-104P)    |                                |
+-------------------------------------+--------------------------------+
| Health Physics Technology (H-201)   |                                |
+-------------------------------------+--------------------------------+
| Respiratory Protection (H-311)      |                                |
+-------------------------------------+--------------------------------+
| Environmental Monitoring for        |                                |
| Radioactivity (H-111)               |                                |
+-------------------------------------+--------------------------------+
| Radwaste Management (H-202)         |                                |
+-------------------------------------+--------------------------------+
| Transportation of Radioactive       |                                |
| Materials (H-308)                   |                                |
+-------------------------------------+--------------------------------+
| ***B. Individual Study              |                                |
| Activities***                       |                                |
+-------------------------------------+--------------------------------+
| (ISA-HP-1) Code of Federal          |                                |
| Regulations (CFR=s)                 |                                |
+-------------------------------------+--------------------------------+
| (ISA-HP-2) Licensee Documents for   |                                |
| Health Physics                      |                                |
|                                     |                                |
| Inspectors                          |                                |
+-------------------------------------+--------------------------------+
| (ISA-HP-3) Access Controls to       |                                |
| Radiologically Significant          |                                |
|                                     |                                |
| Areas                               |                                |
+-------------------------------------+--------------------------------+
| (ISA-HP-4) ALARA Planning and       |                                |
| Controls                            |                                |
+-------------------------------------+--------------------------------+
| (ISA-HP-5) Radiation Monitoring     |                                |
| Instrumentation                     |                                |
+-------------------------------------+--------------------------------+

+-------------------------------------+--------------------------------+
| (ISA-HP-6) Radioactive Material     |                                |
| Processing and                      |                                |
|                                     |                                |
| Transportation                      |                                |
+-------------------------------------+--------------------------------+
| (ISA-HP-7) Radiological             |                                |
| Environmental Monitoring            |                                |
|                                     |                                |
| Program (REMP) and Radioactive      |                                |
| Material                            |                                |
|                                     |                                |
| Control Program                     |                                |
+-------------------------------------+--------------------------------+
| (ISA-HP-8) Radioactive Gaseous and  |                                |
| Liquid Effluent                     |                                |
|                                     |                                |
| Treatment and Monitoring Systems    |                                |
+-------------------------------------+--------------------------------+
| (ISA-HP-9) Significance             |                                |
| Determination Process               |                                |
|                                     |                                |
| \- Occupational Radiation Safety    |                                |
+-------------------------------------+--------------------------------+
| (ISA-HP-10) Significance            |                                |
| Determination Process               |                                |
|                                     |                                |
| \- Public Radiation Safety:         |                                |
| Radioactive Material                |                                |
|                                     |                                |
| Control                             |                                |
+-------------------------------------+--------------------------------+
| (ISA-HP-11) Significance            |                                |
| Determination Process               |                                |
|                                     |                                |
| \- Public Radiation Safety:         |                                |
| Effluent Release                    |                                |
|                                     |                                |
| Program                             |                                |
+-------------------------------------+--------------------------------+
| (ISA-HP-12) Significance            |                                |
| Determination Process               |                                |
|                                     |                                |
| \- Public Radiation Safety:         |                                |
| Environmental Monitoring Program    |                                |
+-------------------------------------+--------------------------------+
| (ISA-HP-13) Significance            |                                |
| Determination Process               |                                |
|                                     |                                |
| \- Public Radiation Safety:         |                                |
| Transportation Branch               |                                |
+-------------------------------------+--------------------------------+
| (ISA-HP-14) Performance Indicator - |                                |
| Occupational                        |                                |
|                                     |                                |
| Radiation Safety                    |                                |
+-------------------------------------+--------------------------------+
| ***C. On-the-Job Activities***      |                                |
+-------------------------------------+--------------------------------+
| (OJT-HP-1) Inspecting Access        |                                |
| Controls to Radiologically          |                                |
|                                     |                                |
| Significant Areas                   |                                |
+-------------------------------------+--------------------------------+
| (OJT-HP-2) Inspecting ALARA         |                                |
| Planning and Controls               |                                |
+-------------------------------------+--------------------------------+
| (OJT-HP-3) Inspecting Radiation     |                                |
| Monitoring                          |                                |
|                                     |                                |
| Instrumentation                     |                                |
+-------------------------------------+--------------------------------+
| (OJT-HP-4) Radioactive Material     |                                |
| Processing and                      |                                |
|                                     |                                |
| Transportation                      |                                |
+-------------------------------------+--------------------------------+
| (OJT-HP-5) Radiological             |                                |
| Environmental Monitoring            |                                |
|                                     |                                |
| Program (REMP) and Radioactive      |                                |
| Material                            |                                |
|                                     |                                |
| Control Program                     |                                |
+-------------------------------------+--------------------------------+
| (OJT-HP-6) Radioactive Gaseous and  |                                |
| Liquid Effluent                     |                                |
|                                     |                                |
| Treatment and Monitoring Systems    |                                |
+-------------------------------------+--------------------------------+

Supervisor=s Recommendation: Signature / Date:
\_\_\_\_\_\_\_\_\_\_\_\_\_\_\_\_\_\_\_\_\_\_\_\_\_\_\_\_\_\_

Division Director=s Approval: Signature / Date:
\_\_\_\_\_\_\_\_\_\_\_\_\_\_\_\_\_\_\_\_\_\_\_\_\_\_\_\_\_\_

Copies to: Inspector

HR Office

Supervisor

ATTACHMENT 1

Revision History For IMC 1245, Appendix C3

+--------+-------+---------------------+------+----------+--------------+
| Comm   | Issue | Description of      | Trai | Training | Comment      |
| itment | Date  | Change              | ning | Co       | Resolution   |
| Tr     |       |                     | Ne   | mpletion | Accession    |
| acking |       |                     | eded | Date     | Number       |
| Number |       |                     |      |          |              |
+--------+-------+---------------------+------+----------+--------------+
| N/A    | 10/   | This IMC has been   | None | N/A      | N/A          |
|        | 16/06 | revised to          |      |          |              |
|        |       | incorporate         |      |          |              |
|        | CN    | comments from the   |      |          |              |
|        | 0     | Commission in which |      |          |              |
|        | 6-027 | the term public     |      |          |              |
|        |       | confidence has been |      |          |              |
|        |       | change to openness  |      |          |              |
+--------+-------+---------------------+------+----------+--------------+
|        |       |                     |      |          |              |
+--------+-------+---------------------+------+----------+--------------+
|        |       |                     |      |          |              |
+--------+-------+---------------------+------+----------+--------------+
|        |       |                     |      |          |              |
+--------+-------+---------------------+------+----------+--------------+
|        |       |                     |      |          |              |
+--------+-------+---------------------+------+----------+--------------+