DOE-HDBK-1122-2009 Chg Notice 2, Radiological Control Technician Training (Part 6 of 9)

Part 6 of 9 Radiological Control Technician Training Site Academic Training Study Guide Phase I Coordinated and Conducted for the Office of Health, Safety and Security U.S. Department of Energy DOE-HDBK-1122-2009 Radiological Control Technician Study Guide ii This page intentionally left blank DOE-HDBK-1122-2009 Radiological Control Technician Study Guide iii Table of Contents Page Module 2.01 Radiological Documentation ............................................................................. 2.01-1 Module 2.02 Communication Systems ................................................................................... 2.02-1 Module 2.03 Counting Errors and Statistics ........................................................................... 2.03-1 Module 2.04 Dosimetry .......................................................................................................... 2.04-1 Module 2.05 Contamination Control ...................................................................................... 2.05-1 Module 2.06 Airborne Sampling Program/Methods .............................................................. 2.06-1 Module 2.07 Respiratory Protection ....................................................................................... 2.07-1 Module 2.08 Radiological Source Control ............................................................................. 2.08-1 Module 2.09 Environmental Monitoring ................................................................................ 2.09-1 Module 2.10 Access Control and Work Area Setup............................................................... 2.10-1 Module 2.11 Radiological Work Coverage ............................................................................ 2.11-1 Module 2.12 Shipment and Receipt of Radioactive Material................................................. 2.12-1 Module 2.13 Radiological Incidents and Emergencies........................................................... 2.13-1 Module 2.14 Personnel Decontamination............................................................................... 2.14-1 Module 2.15 Radiological Considerations for First Aid......................................................... 2.15-1 Module 2.16 Radiation Survey Instrumentation..................................................................... 2.16-1 Module 2.17 Contamination Monitoring Instrumentation...................................................... 2.17-1 Module 2.18 Air Sampling Equipment................................................................................... 2.18-1 Module 2.19 Counting Room Equipment............................................................................... 2.19-1 DOE-HDBK-1122-2009 Radiological Control Technician Study Guide iv This page intentionally left blank DOE-HDBK-1122-2009 Module 2.01 Radiological Documentation Study Guide 2.01-1 Course Title: Radiological Control Technician Module Title: Radiological Documentation Module Number: 2.01 Objectives: → 2.01.01 List the types of records/reports that the Radiological Control group is responsible for maintaining at your site.

→ 2.01.02 Describe the types of records and reports used at your site by the Radiological Control Group, to include but should not be limited to: a. Radiological Work Permits b. Survey Reports c. Analysis Reports d. Radiological Deficiency Reports e. ALARA Documentation f. Exposure Reports → 2.01.03 Explain the requirements for the records management system, such as QC, auditability/retrievability, management information at your site. INTRODUCTION 10 CFR 835 establishes radiation protection standards, limits, and program requirements for protecting individuals from ionizing radiation resulting from the conduct of DOE activities. It is important to maintain the proper documentation to ensure that these standards and requirements are being met. An RCT plays a vital role in supporting these requirements through proper documentation. PURPOSE AND REQUIREMENTS Radiological control records are needed to demonstrate the effectiveness of the overall Radiation Protection program at DOE facilities. The records are used to document radiological safety afforded to personnel on-site. Radiological control records become valuable tools in work planning, evaluating past trends, and guiding future performance goals. These records may become the basis for public disclosures, legal proceedings, medical assessment and audits to show compliance with company, state or federal requirements. Because of this, it is important that these records be of high quality, readily retrievable, and managed for the prescribed retention period. It is suggested that these records be cross-referenced, when applicable, to aid in their retrieval. DOE-HDBK-1122-2009 Module 2.01 Radiological Documentation Study Guide 2.01-2 References: 1. 10 CFR Part 835 (2007) "Occupational Radiation Protection. 2. "U.S. Department of Energy Radiological Control Standard"(2008). RADIOLOGICAL RECORDS MANAGEMENT PROGRAM Each DOE site must have a radiological records management program to ensure that auditable records and reports are controlled through the stages of creation, distribution, use, arrangement, storage, retrieval, media conversion and disposition. The radiological records management program should include the following: a. Radiological Policy Statements b. Radiological Control Procedures c. Individual Radiological Doses d. Internal and External Dosimetry Policies and Procedures (including Bases Documents) e. Personnel Training (course records and individual records) f. ALARA Records g. Radiological Instrumentation Test, Repair and Calibration Records h. Radiological Surveys i. Area Monitoring Dosimetry Results j. Radiological Work Permits k. Radiological Performance Indicators and Assessments l. Radiological Safety Analysis and Evaluation Reports m. Quality Assurance Records n. Radiological Incident and Occurrence Reports (and Critique Reports, if applicable) o. Accountability records for sealed radioactive sources p. Records for release of material to Controlled Areas q. Reports of loss of radioactive material. (Insert site specific information here.) RADIOLOGICAL RECORD KEEPING STANDARDS 2.01.01 List the types of records/reports that the Radiological Control group is responsible for maintaining at your site. DOE-HDBK-1122-2009 Module 2.01 Radiological Documentation Study Guide 2.01-3

Record keeping standards have been set by the Department of Energy. In addition to the requirement of being accurate and legible, all radiological records must include the following: • identification of the facility, specific location, function, and process • signature or other identifying code of the preparer and date • legible entries in black ink • corrections identified by a single line-out, initialed, and dated • supervisory signature to indicate review and proper completion of the forms. Each radiological control organization should maintain a file of names, signatures and initials for future identification of the person who signed or initialed a record. In addition, radiological control records should not include: • records that are corrected using opaque substances • records that contain shorthand or other nonstandardized terms. TYPES OF RADIOLOGICAL RECORDS Various types of records are included in the radiological records management program. These records fall into the following categories: Employment History Records Records detailing an employee's previous and on-going radiological work assignments, yearly doses at DOE and non-DOE facilities must be maintained. Where practical, the association between the radiation dose and job function must be preserved for trend analyses and future worker health studies. Personnel Radiological Records Occupational Radiation Dose Records must be maintained for all contractor, subcontractor, and Federal employees who are part of the personnel dosimetry program. These records include results of personnel extremity, skin, eye and whole body external dose measurements. The records also contain internal dose information, including in vivo measurements, urine, fecal and other specimen analysis and dose assessments. A complete record of radiological incidents and occurrences involving personnel dose must also be retained. The investigation and counseling of personnel radiological concerns must also be documented and maintained within the radiological control record management program. DOE-HDBK-1122-2009 Module 2.01 Radiological Documentation Study Guide 2.01-4 Medical Records Reports of periodic medical examinations and evaluations, respirator fit-testing results and records of medical treatment performed in support of the radiological control program should be maintained. Sites are encouraged to maintain nonoccupational radiation doses, such as therapeutic or large amounts of diagnostic radiation doses for employees. Radiological Training and Qualification Records Records of training and qualification in radiological control are permanently maintained to demonstrate that a person received appropriate information to perform the work assignment in a safe manner. Qualification standard records are retained for on-the-job, practical and formal classroom training. Training and qualification records are available to first-line supervision and management to aid in making work assignments. Included in the training records maintained, are quizzes, tests, responses and acknowledgements of training, with the date and signature of the person trained. Instrumentation and Calibration Records

Records of calibration, modification or maintenance, and periodic operational checks of fixed, portable, and laboratory radiation measuring equipment must be maintained. These records include frequencies, method, dates, personnel, training and traceability of calibration sources. Records of additional tests and checks of instrumentation used in conjunction with a suspected overexposure, questionable indication, or unusual occurrence are also be retained. Radiological Control Procedures Sites should maintain radiological control procedures, policies, ALARA records, work procedures, Radiological Problem Reports, Radiation Work Permits, and supporting data as part of their radiological control record management program. Specific requirements for each of these documents can be found in the DOE RCS. 2.01.02 Describe the types of records and reports used at your facility by the Radiological Control group, to include but should not be limited to: a. Radiological Work Permits b. Survey Reports c. Analysis Reports d. Radiological Deficiency Reports e. ALARA Documentation f. Exposure Reports: DOE-HDBK-1122-2009 Module 2.01 Radiological Documentation Study Guide 2.01-5 (Insert site specific information here.) RECORDS MANAGEMENT All records are required to be stored in a manner that ensures that they can be retrieved, in addition to being able to maintain their integrity and security. Once a record has been created, reviewed, and signed by appropriate supervision, the record should be considered complete and must not be modified. Subsequent errors identified in a completed record should be corrected by creating a supplemental record that includes traceability for the correction. Radiological Control records should be protected from temperature extremes, moisture, infestation, electromagnetic fields, excessive light, stacking, theft, and vandalism. Protective measures should include vaults, file rooms with fixed fire suppression, fire-rated cabinets, duplicate storage, or a combination of these. (Insert site specific information here.) RADIOLOGICAL REPORTING All personnel that are monitored by a personnel dosimetry program shall be provided an annual report of their radiation exposure. A person may also receive a current radiation dose record upon special request. Terminating employees will be given an exposure report within 90 days of their last day of employment summarizing their radiation dose for the total period of employment at the reporting facility. 2.01.03 Explain the requirements for the records management system, such as QC, auditability/retrievability, management information at your facility. DOE-HDBK-1122-2009 Module 2.02 Communications Systems Study Guide 2.02-1 Course Title: Radiological Control Technician Module Title: Communication Systems Module Number: 2.02 Objectives : 2.02.01 Explain the importance of good communication. 2.02.02 Identify two methods of communication and be able to determine different types of each. 2.02.03 Describe different types of communication systems. 2.02.04 Describe the FCC and DOE guidelines regarding proper use of communication systems. 2.02.05 Describe general attributes of good communications. 2.02.06 Explain the importance of knowing how to contact key personnel. → 2.02.07 Identify the communication systems available at your site and methods

available to contact key personnel. → 2.02.08 Describe the emergency communication systems available at your site. References: (Add any site-specific references.) IMPORTANCE OF COMMUNICATION Good communication is important in everyday life to make sure our message is clear, understood, and received. A clear concise communication eliminates confusion and the possibility of misunderstanding. It is important that the receiver understand the communication without unnecessary interpretation or guess work. For a communication to be completed there must be a receiver. The receiver is the person or group that the communication is intended. For a good communication process there must be a clear concise message, a medium of transmission (i.e. telephone, telegraph, E-mail, letter, signal flag, etc.), and a receiver. If a response is required by the receiver, this can serve as confirmation of reception of the communication, however, a response alone does not indicate the communication was understood correctly. Misunderstanding of 2.02.01 Explain the importance of good communication.c. Energy DOE-HDBK-1122-2009 Module 2.02 Communications Systems Study Guide 2.02-2 communication can potentially cause personal as well as physical damage to equipment and surroundings. In all communication processes, a sender of the communication must not assume knowledge that is needed for safe execution of the desired response. The communication must contain all pertinent information. Assuming or hoping the receiver has a given understanding of a process can lead to an unsafe condition. This is especially true in emergency situations that require immediate action or response. Make sure in all communications that desired responses are not outside the abilities or scope of the individual or group. METHODS OF COMMUNICATION In today's atmosphere of technology, there are methods of communications that seemed unlikely just 20 years ago. Who would have thought that a car phone would be as common as a home telephone. In general, communication can be broken into two groups, verbal and nonverbal. Verbal methods of communication include talking directly to another person, telephone conversation, voice mail, video tele-conferencing, and various other available mediums. Verbal methods generally allow discussion of details followed by questions and/or an immediate response. Verbal communication allows flexibility in the message along with added information without too much difficulty in transmission. Nonverbal methods of communication include signs, letters, signals, gestures, documents, E-mail, and various other available mediums. Nonverbal methods can limit the amount of information transmitted due to the difficulty in the transmission method. COMMUNICATION SYSTEMS There are several communication systems available at most DOE sites. These may include public address system, telephones, two-way radio, pagers, computer mail system, and computer based bulletin boards. Following is a brief generic description of each of these communication systems. The description is not meant to be all inclusive, but a cursory overview of key aspects of each system. 2.02.02 Identify the two methods of communication and be able to determine different types of each. 2.02.03 Describe different types of communication systems. DOE-HDBK-1122-2009

Module 2.02 Communications Systems Study Guide 2.02-3 Public Address The public address system consists of loud speakers and calling stations located throughout an area to provide audible notification to all personnel within the area. The public address system may be used for routine messages, contacting groups or individuals, items of interest to the general population, and emergency notifications or warnings. The public address system should be administratively controlled to ensure effectiveness in contacting facility personnel and availability during emergency conditions. Telephones Telephones provide a means for point to point communication. The telephone may be considered semiprivate when compared to the public address system, however while on a DOE facility, all calls are subject to monitoring for security reasons. The telephone system may offer the ability to leave a voice message whenever the receiving party is unavailable. The telephone system provides communication, but is subject to usage by other individuals which may impede your contacting the person or persons needed in an emergency situation. Two-way Radio Two-way radio communication provides a direct link to other individuals on your frequency or net. Although "traffic" on the radio may impair your message from being clearly understood, usage is controlled by possession of a radio with the correct frequency. Radio communication is subject to interference by outside sources, which may garble or mask the message. This may be of significance during emergency situations when location or type of emergency in progress must be relayed to response teams. Two-way radios do provide mobility and access while at remote locations. Pagers Pagers a small electronic devices capable of receiving signals from the telephone system to alert the carrier of intended communication from another party. Pagers provide access to personnel while away from the work location. Most pagers provide only a voice message or phone number to contact. Pagers normally do not allow the carrier to respond directly to the page verbally. Pagers provide a means of contacting personnel when there whereabouts are unknown, but are assumed to be within the site boundaries or very nearby. Computer Mail Systems Computer mail systems provide communication between computer terminals. Most systems are linked via a local area network. This links enables users to DOE-HDBK-1122-2009 Module 2.02 Communications Systems Study Guide 2.02-4 contact individuals or groups directly and leave written messages for these individuals to receive. Computer mail systems enable the user to contact receivers directly while other users are unaware. Computer Bulletin Boards Computer bulletin boards provide communication to anyone with access to the bulletin board. The user provides messages or information without knowing who will receive the information. Usually messages and information of general subject matter or routine information that apply to most users are available on a bulletin board. Most bulletin boards are controlled with minimal requirements for access. Bulletin boards provide a means for communicating with a large diverse group. FCC AND DOE RULES AND REGULATIONS When using communication systems licensed by the Federal Communications Commission and operated by the Department of Energy, one cannot:

• Use profane, indecent, or obscene language. • Willfully damage or permit radio equipment damage. • Cause malicious interference with any radio communications. • Intercept and use or publish the contents of any radio message without the permission of the proper authorities. • Make unnecessary or unidentified transmissions. • Transmit without first making sure that the transmission will not cause harmful interference. • Make any adjustments, repairs, or alterations to a radio transmitter without licensing by the FCC or acceptable equivalent. • Transmit a call sign, letter, or numeral which has not been assigned to your station. • Rebroadcast another transmission (i.e. radio station music). 2.02.04 Describe the FCC and DOE guidelines regarding proper use of communication systems. DOE-HDBK-1122-2009 Module 2.02 Communications Systems Study Guide 2.02-5 GENERAL ATTRIBUTES OF GOOD COMMUNICATIONS • Minimize the use of abbreviations and acronyms. Only abbreviations and acronyms from and approved list should be used in facility communication. • Make all oral instructions clear and concise. Do not include multiple actions in a verbal instruction which may get confused or misunderstood. • Ensure the identity of the person(s) is/are clearly understood. Identify yourself and your position, and ensure that you know to whom you are speaking. • Use clear, precise terminology. Do not use slang terms. Avoid words that sound alike. Use commonly agreed upon terms. Employ the phonetic alphabet for clarification. (See Table 1) Table 1. Phonetic Alphabet and Numbers A - Alpha J - Juliett S - Sierra 1 - One B - Bravo K - Kilo T - Tango 2 - Two C - Charlie L - Lima U - Uniform 3 - Three D - Delta M - Mike V - Victor 4 - Fower E - Echo N - November W - Whiskey 5 - Fife F - Foxtrot O - Oscar X - X-Ray 6 - Six G - Golf P - Papa Y - Yankee 7 - Seven H - Hotel Q - Quebec Z - Zulu 8 - Eight I - India R - Romeo . - Point 9 - Niner 0 - Zero 2.02.05 Describe general attributes of good communications. DOE-HDBK-1122-2009 Module 2.02 Communications Systems Study Guide 2.02-6 • Repeat back messages, either paraphrased or verbatim. • Speak distinctly and deliberately. • Acknowledge all communications. CONTACT OF KEY PERSONNEL The importance of knowing how to contact key personnel can not be understated. The importance lies in getting the knowledgeable people at the location they are needed. This can apply to emergency situations, routine circumstances, or nonroutine circumstances. The ability of the RCT to contact key personnel can reduce personnel injury, equipment damage, uncontrolled radioactive release, unrestricted movement of controlled materials, and other important actions. The RCT must be aware of the location of communication equipment, phone numbers or pager numbers, and/or emergency numbers regardless of location. Familiarity with the working environment will reduce the time needed to contact key personnel. The RCT must be aware of the location, situation, and personnel or equipment involved. This information must be relayed without misinterpretation to key personnel to afford proper response. SITE COMMUNICATION SYSTEMS (Insert site specific information here.) SITE EMERGENCY COMMUNICATIONS (Insert site specific information here.)

2.02.06 Explain the importance of knowing how to contact key personnel. 2.02.07 Identify the communication systems available at your site and methods available to contact key personnel. 2.02.08 Describe the emergency communication systems available at your site. DOE-HDBK-1122-2009 Module 2.02 Communications Systems Study Guide 2.02-7 SUMMARY This lesson has covered topics related to effective communications, contacting key personnel, and emergency communications. As an RCT you should be aware of your location and what communication systems are available to you while working on any job or situation. DOE-HDBK-1122-2009 Module 2.02 Communications Systems Study Guide 2.02-8 This page intentionally left blank. DOE-HDBK-1122-2009 (Revised 2013) Module 2.03 Counting Errors and Statistics Student's Material Course Title: Radiological Control Technician Module Title: Counting Errors and Statistics Module Number: 2.03 Objectives: (This document, Study Material, is referred to as Study Guide in the Program Management Guide) 2.03.01. Identify five general types of errors that can occur when analyzing radioactive samples, and describe the effect of each source of error on sample measurements. 2.03.02. State two applications of counting statistics in sample analysis. 2.03.03. Define the following terms: a. mode b. median c. mean 2.03.04. Given a series of data, determine the mode, median, or mean. 2.03.05. Define the following terms: a. variance b. standard deviation 2.03.06. Given the formula and a set of data, calculate the standard deviation. 2.03.07. State the purpose of a Chi-squared test. → 2.03.08. State the criteria for acceptable Chi-squared values at your site. 2.03.09. State the purpose of creating quality control (QC) charts. → 2.03.10. State the requirements for maintenance and review of QC charts at your site. 2.03.11. State the purpose of calculating warning and control limits. 2.03.12. State the purpose of determining efficiencies and correction factors. 2.03.13. Given counting data and source assay information, calculate efficiencies and correction factors. DOE-HDBK-1122-2009 (Revised 2013) Module 2.03 Counting Errors and Statistics Student's Material 2.03.14. State the meaning of counting data reported as x ± y. 2.03.15. Given counting results and appropriate formulas, report results to desired confidence level. 2.03.16. State the purpose of determining background. → 2.03.17. State the method and requirements for determining background for counting systems at your site. 2.03.18. State the purpose of performing sample planchet maintenance. → 2.03.19. State the method and requirements for performing planchet maintenance for counting systems at your site. 2.03.20. Explain methods to improve the statistical validity of sample measurements. 2.03.21. Define "detection limit," and explain the purpose of using detection limits in the analysis of radioactive samples. → 2.03.22. Given the formula and necessary information, calculate detection limit values for counting systems at your site. 2.03.23. State the purpose and method of determining crosstalk. → 2.03.24. State the criteria for acceptable values of crosstalk for counting systems at your site. 2.03.25. State the purpose of performing a voltage plateau.

→ 2.03.26. State the method of performing a voltage plateau on counting systems at your site. DOE-HDBK-1122-2009 (Revised 2013) Module 2.03 Counting Errors and Statistics Student's Material References: 1. "Advanced Health Physics Course Prestudy Guide," United States Nuclear Regulatory Commission, General Physics Corporation, 1987. 2. Chase & Rabinowitz, "Principles of Radioisotope Methodology," 3rd Edition, Burgess Publishing, 1987. 3. Gollnick, Daniel A., "Basic Radiation Protection Technology," 6th Edition, Pacific Radiation Corporation, Altadena, CA, 2012. 4. Knoll, Glenn F., "Radiation Detection and Measurement," 4th Edition, John Wiley & Sons, New York, 2010. 5. "Webster's New World Dictionary," 4thd College Edition, Webster's New World, Cleveland & New York, 2003. 6. Moe, Harold, "Operational Health Physics Training," ANL-88-26; DOE; Argonne National Laboratory, Chicago, 1988. 7. "Introduction to Low-background Counting Systems," Oxford-Tennelec Instruments. 8. Environmental Implementation Guide for Radiological Survey Procedures--Draft; DOE; February 1997. 9. Accuracy and Detection Limits for Bioassay Measurements in Radiation Protection - Statistical Considerations; U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research; April 1986. 10. “Manual for Conducting Radiological Surveys in Support of License Termination – Draft Report for Comment,” NUREG/CR-5849/ORAU-92/C57, June 1992. 11. “Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM),” NUREG-1575, Rev. 1 / EPA 402-R-97-016, Rev. 1 / DOE/EH-0624, Rev. 1, August 2000. 12. Multi-Agency Radiation Survey and Assessment of Materials and Equipment Manual (MARSAME) (NUREG-1575, Supplement 1), January 2009. DOE-HDBK-1122-2009 (Revised 2013) Module 2.03 Counting Errors and Statistics Student's Material INTRODUCTION Radiological sample analysis involves observation of a random process, one that may or may not occur, and estimation of the amount of radioactive material present based on that observation. All over the country radiation protection personnel are using activity measurements to make decisions that may affect the health and safety of workers at those facilities and their surrounding environments. This unit presents an overview of measurement processes, and statistical evaluation of both measurements and equipment performance. In addition, this unit addresses some of the actions to take to minimize the sources of error in count room operations. 2.03.01 Identify five general types of errors that can occur when analyzing radioactive samples, and describe the effect of each source of error on sample measurements. GENERAL SOURCES OF ERROR Assuming the counting system is calibrated correctly, there are five general sources of error associated with counting a sample: 1. Self-absorption 2. Backscatter 3. Resolving time 4. Geometry 5. Random disintegration of radioactive atoms (statistical variations). Self-Absorption When a sample has an abnormally large amount of material on the sample media (or the sample itself is large), it could introduce a counting error due to self-absorption, which is the absorption of the emitted radiation by the sample material itself. Self-absorption could occur for: • Liquid samples with a high solid content • Air samples from a high dust area • Use of improper filter paper may introduce a type of self-absorption, especially in

alpha counting (i.e., absorption by the media, or filter). DOE-HDBK-1122-2009 (Revised 2013) Module 2.03 Counting Errors and Statistics Student's Material Personnel counting samples should ensure the correct sample media is used, and that the sample does not become too heavily loaded with sample material. Count room personnel should be routinely checking samples for improper media or heavily loaded samples. Backscatter Counting errors due to backscatter occur when the emitted radiation traveling away from the detector is reflected, or scattered back, to the detector by the material in back of the sample. The amount of radiation that is scattered back will depend upon the type and energy of the radiation and the type of backing material (reflector). The amount of backscattered radiation increases as the energy of the radiation increases and as the atomic number of the backing material increases. Generally, backscatter error is only a consideration for particulate radiation, such as alpha and beta particles. Because beta particles are more penetrating than alpha particles, backscatter error will be more pronounced for beta radiation. The ratio of measured activity of a beta source counted with a reflector compared to counting the same source without a reflector is called the backscatter factor (BF). counts − with − reflector (Equation 1) BF = counts − without − reflector Normally, backscatter error is accounted for in the efficiency or conversion factor of the instrument. However, if different reflector materials, such as aluminum and stainless steel, are used in calibration and operation, an additional unaccounted error is introduced. (This additional error will be about 6% for stainless steel versus aluminum.) Count room personnel must be aware of the reflector material used during calibration of the counting equipment. Any deviation from that reflector material will introduce an unaccounted error and reduce confidence in the analysis results. Resolving Time Resolving time (or dead time) is the time interval which must elapse after a detector pulse is counted before another pulse can be counted. Any radiation entering the detector during the resolving time will not be recorded as a pulse; therefore, the information on that radiation interaction is lost. As the activity, or decay rate, of the sample increases, the amount of information lost during the resolving time of the detector is increased. As the losses from resolving time increase, an additional error in the measurement is introduced. Typical resolving time losses are shown in Table 1. Table 1. Typical Resolving Time Losses DOE-HDBK-1122-2009 (Revised 2013) Module 2.03 Counting Errors and Statistics Student's Material Count rate (cpm) GM Tube 1 Proportional 2 Scintillation 3 20,000 1.7% < 1% < 1% 40,000 3.3% < 1% < 1% 60,000 5.0% < 1% < 1% 100,000 8.3% < 1% 1.0% 300,000 25.0% < 1% 3.5% 500,000 42.0% 1.5% 5.8% 1 GM tube: 50µs resolving time 2 Proportional detector: 2µs resolving time 3 Scintillation detector: 7µs resolving time DOE-HDBK-1122-2009 (Revised 2013) Module 2.03 Counting Errors and Statistics Student's Material Resolving time losses can be corrected by using the equation: R (Equation 2) R = o 1− R τ o where: R = "true" count rate, in cpm Ro = observed count rate, in cpm τ = resolving time of the detector, in minutes ("tau") Count room personnel should be aware of the limitations for sample count rate, based

upon procedures and the type of detector in use, to prevent the introduction of additional resolving time losses. This is especially true for counting equipment that uses GM detectors. Instructor note: A detector, or detection system, can be characterized by a paralyzable or non­ paralyzable behaviour. In a non-paralyzable detector, an event happening during the dead time since the previous event is simply lost, so that with an increasing event rate the detector will reach a saturation rate equal to the inverse of the dead time. In a paralyzable detector, an event happening during the dead time since the previous one will not just be missed, but will restart the dead time, so that with increasing rate the detector will reach a saturation point where it will be incapable of recording any event at all. A semi-paralyzable detector exhibits an intermediate behaviour, in which the event arriving during dead time does extend it, but not by the full amount, resulting in a detection rate that decreases when the event rate approaches saturation. Geometry Geometry related counting errors result from the positioning of the sample in relation to the detector. Normally, only a fraction of the radiation emitted by a sample is emitted in the direction of the detector because the detector does not surround the sample. If the distance between the sample and the detector is varied, then the fraction of emitted radiation which hits the detector will change. This fraction will also change if the orientation of the sample under the detector (i.e., side-to-side) is varied. An error in the measurement can be introduced if the geometry of the sample and detector is varied from the geometry used during instrument calibration. This is especially critical for alpha counting, where any change in the sample-to-detector distance also increases (or decreases) the chance of attenuation of the alpha particles by the air between the sample and detector. Common sources examples of geometry-related errors include: DOE-HDBK-1122-2009 (Revised 2013) Module 2.03 Counting Errors and Statistics Student's Material • Piling smears and/or filters on top of each other in the same sample holder (moves the top sample closer to the detector and varies the calibration geometry). • Using deeper or shallower sample holders than those used during calibration (changes the sample-to-detector distance). • Adjusting movable bases in the counting equipment sliding drawer (changes the sample-to-detector distance). • Using too many or inappropriate sample holders or planchets (changes the sample-to-detector distance). Sources not fixed in position can change geometry and reduce reproducibility. • Plexiglas shelving in counting chamber is improperly set. Random Disintegration The fifth source of general counting error is the random disintegration of the radioactive atoms and constitutes the remainder of the lesson. STATISTICS Statistics is a branch of mathematics that deals with the organization, analysis, collection, and interpretation of statistical data. No definition of statistical data is given. However, Webster's does define a statistic as "an estimate of a variable, as an average or a mean, made on the basis of a sample taken from a larger set of data." This last definition is applicable to our discussion of counting statistics. After all, when we take samples, we use the data derived from analysis of those samples to make determinations about

conditions in an area, in water, in air, etc., assuming that the sample is representative. So, we have estimated conditions (a variable) on the basis of a sample (our smear, water sample, air sample) taken from a larger set of data. Over the years, various methods and observations have identified three models which can be applied to observations of events that have two possible outcomes (binary processes). Luckily, we can define most observations in terms of two possible outcomes. For example, look at the following table: Table 2: Probability of Success Trial Definition of Success Probability of Success Tossing a coin "heads" 1/2 DOE-HDBK-1122-2009 (Revised 2013) Module 2.03 Counting Errors and Statistics Student's Material Rolling a die observing a given radioactive nucleus for a time, t "a six" The nucleus decays during the observation 1/6 1-e -λt For each of the processes that we want to study, we have defined a trial (our test), a success and a failure (two possible outcomes), and have determined the probability of observing our defined success. Now, to study these processes, we can use proven, statistical models to evaluate our observations for error. Consider the possibilities when throwing two dice. There are 36 possible outcomes when throwing two dice, as indicated in Table 3. Table 3. Possibilities in Rolling Dice Result Possibilities No. of Possibilities 2 1&1 1 3 1&2,2&1 2 4 1&3,2&2,3&1 3 5 1&4,2&3,4&1,3&2 4 6 1&5,2&4,3&3,4&2,5&1 5 7 1&6,2&5,3&4,4&3,5&2,6&1 6 8 2&6,3&5,4&4,5&3,6&2 5 9 3&6,4&5,5&4,6&3 4 10 4&6,5&5,6&4 3 11 5&6,6&5 2 12 6&6 1 If, in our study of this process, we define a success as throwing a number between 2 and 12, the outcome is academic. All trials will be successful, and we can describe the probabilities of throwing any individual number between the range of 2 and 12 inclusive would add up to 1. If we define a success as throwing a particular number, we can define the probability of our success in terms of the number of possible outcomes that would give us that number in comparison to the total number of possible outcomes. If we were to take two dice, roll the dice a large number of times, and graph the results in the same manner, we would expect these results to produce a curve such as the one shown in Figure 1. DOE-HDBK-1122-2009 (Revised 2013) Module 2.03 Counting Errors and Statistics Student's Material The area under the curve can be mathematically determined and would correspond to the probability of success of a particular outcome. For example, to determine the probability of throwing a particular number between 2 and 12 we would calculate the area under the curve between 2 and 12. The results of that calculation are 36. This is what statistics is all about; random binomial processes that should produce results in certain patterns that have been proven over the years. The three models that are used are distribution functions of binomial processes with different governing parameters. These functions and their restrictions are: • Binomial Distribution This is the most general of the statistical models and is widely applicable to all processes with a constant probability. It is not widely used in nuclear applications because the mathematics is too complex. • Poisson Distribution A simplified version of binomial distribution is the Poisson (pronounced "pwusówn") distribution, which is valid when the probability of success, P(x), is small. If we

graphed a Poisson distribution function, we would expect to see the predicted number of successes at the lower end of the curve, with successes over the entire range if sufficient trials were attempted. Thus, the curve would appear as seen in Figure 2. DOE-HDBK-1122-2009 (Revised 2013) Module 2.03 Counting Errors and Statistics Student's Material The Poisson model is used mainly for applications involving counting system background and detection limits, where the population (i.e., number of counts) is small. This will be discussed in greater detail later. • Gaussian Distribution Also called the "normal distribution," the Gaussian (pronounced "Gowziun") distribution is a further simplification which is applicable if the average number of successes is relatively large, but the probability of success is still low. A graph of a Gaussian distribution function is shown in Figure 3. DOE-HDBK-1122-2009 (Revised 2013) Module 2.03 Counting Errors and Statistics Student's Material Note that the highest number of successes is at the center of the curve, the curve is a bell shaped curve, and the relative change in success from one point to the adjacent is small. Also note that the mean, or average number of successes, is at the highest point, or at the center of the curve. The Gaussian, or normal, distribution is applied to counting applications where the mean count or success is expected to be greater than 20. It is used for counting system calibrations and operational checks, as well as for normal samples containing activity. It may or may not include environmental samples (i.e., samples with very low activity). 2.03.02 State the two purposes for statistical analysis of count room operations. APPLICATION OF STATISTICAL MODELS Application of specific statistical methods and models to nuclear counting operations is termed counting statistics and is essentially used to do two things: • Predict the inherent statistical uncertainty associated with a single measurement, thus allowing us to estimate the precision associated with that measurement. • Serve as a check on the normal function of nuclear counting equipment. DOE-HDBK-1122-2009 (Revised 2013) Module 2.03 Counting Errors and Statistics Student's Material 2.03.03 Define the following terms: a. mode b. median c. mean DEFINITIONS Mode An individual data point that is repeated the most in a particular data set. Median The center value in a data set arranged in ascending order. Mean The average value of all the values in a data set. 2.03.04 Given a series of data, calculate mode, median, or mean. DETERMINATION OF MODE, MEDIAN, AND MEAN • Determination of the Mode: In the set of test scores above, a score of 95 occurs (i.e., is repeated) more often than any other score. • Determination of the Median: In the same set of test scores, this is the score in the middle - where one half of the scores are below, and the other half are above the median. The median for the test scores in Figure 4 is 90. • Determination of the Mean: This is found by adding all of the values in the set together, and dividing by the number of values in the set. The mean of the nine test scores is 89. DOE-HDBK-1122-2009 (Revised 2013) Module 2.03 Counting Errors and Statistics Student's Material Mean determination is often expressed using special symbols, as illustrated in the following equation: Σx (Equation 3) x = i n 2.03.05 Define the following terms: a. variance

b. standard deviation 2.03.06 Given the formula and a set of data calculate the standard deviation. VARIANCE AND STANDARD DEVIATION Using the Gaussian distribution model depicted in Figure 5 (below) we need to define the terms "variance" and "standard deviation," which are both used as descriptors of the spread of the population (or the data set) in a normal distribution. E f data points around the mean , it tells how much the data "varies" from the mean. al distribution, the standard deviation is recise than the variance is The standard deviation of a sample is defined mathematically are located close to the mean, the curve will be tall and steep value for a standard ot as steep and have a larger numerical value for a standard , it has been determined mathematically that 68.2% of the area in the data point located at the mean ± (plus or minus) one 95.4% of the area under the curve falls between the data point deviations (2 ated at ± processes is that if the distribution (as depicted in Figure 5) is ng function with a mean observable success >20 (Gaussian the observed successes (or counts) will be within ±1 standard ean. the observed successes (or counts) will be within ±2 standard ean. D Counting Errors an t' amount of scatter e sample . In other word an. Mathematically, in a nor uare root . A term more epresented by a (pronounced “sigma”). athematically = If most of the data points tall and steep and have a low numerica re scattered, the curve will be lower and r a standard In a Gaussian distributio .2% of the area it inus) one ) the data point two standard curve falls point lo hat this means to us in terms of countin in Figure 5) is representative of a count (Gaussian 68.2% of the tim in ±1 standard deviation of the 95.4% of the tim in ±2 standard deviations of the Stude is defined as t . In other words, it tells how much the data "varies" from the m the s , (pronounced “sigma”). The standard deviation of a sample is defined If most of the data points are located close to the mean, the curve will be deviation. If data points curve will be lower and not as steep and have a larger numerical value f In a Gaussian distribution, it has been determined mathematically that 6 ithin the data point located at the mean ± (plus or ); 95.4% of the area under the curve falls betwee ), 99.97% of the area under the ) etc. to us in terms of counting processes is that if the distribution (as depicte representative of a counting function with a mean observable success >2 68.2% of the time the observed successes (or counts) will be wit 95.4% of the time the observed successes (or counts) will be wit DO OE-HDBK-1122-2009 (Revised 2013) Module 2.03 Counting Errors and d Statistics Studennt's Material Variance The amount of scatter o of data points around the mean is defined as th he sample variance. In other words s, it tells how much the data "varies" from the me ean. Standard Deviation Mathematically, in a norm mal distribution, the standard deviation is the sq quare root of the variance. A term more p precise than the variance is standard deviation, r represented by a σ (pronounced “sigma”). The standard deviation of a sample is defined m mathematically as: (Equation 4) σ = √(∑(xi ­ ) 2 / (n - 1)) where: σ = standard deviation of a sample xi = sample counts for each data point = mean n = number of data points If most of the data points are located close to the mean, the curve will be tall and steep

and have a low numericall value for a standard deviation. If data points a are scattered, the curve will be lower and n not as steep and have a larger numerical value fo or a standard deviation. In a Gaussian distribution n, it has been determined mathematically that 68 8.2% of the area under the curve falls with hin the data point located at the mean ± (plus or mminus) one standard deviation (1 σ ); ; 95.4% of the area under the curve falls between n the data point located at ± two standard deviations (2 σ ), 99.97% of the area under the curve falls between the data point loccated at ± three standard deviations (3 σ ) etc. WWhat this means to us in terms of counting g processes is that if the distribution (as depicted d in Figure 5) is representative of a counti ing function with a mean observable success >200 (Gaussian distribution): • 68.2% of the time e the observed successes (or counts) will be with hin ±1 standard deviation of the m mean. • 95.4% of the time e the observed successes (or counts) will be with hin ±2 standard deviations of the m mean. DOE-HDBK-1122-2009 (Revised 2013) Module 2.03 Counting Errors and Statistics Student's Material • 99.97% of the time the observed successes (or counts) will be within ±3 standard deviations of the mean. Remember, the area of the curve represents the probability of success in a random process. In radiation protection this random process is the decay of a radioactive sample. The known statistical distribution is used in radiation protection when setting up a counting system and in evaluating its operation by means of daily pre-operational source checks. In performing the calibration of the system, a radioactive source with a known activity is counted twenty times for one minute each time. Using the data from the twenty counts, the mean and standard deviation can be calculated. The mean can then be used to determine the efficiency of the system while allowing for a certain number of standard deviations during operation. The twenty counts can also be used to perform another required test of the system's performance, the chi-squared test (see below). Example 2.03-1 Calculate the mean and sample standard deviation for the following data set: 193, 188, 202, 185, 179, 217, 191 199, 201, 214, 193, 232, 199, 210, 196, 211, 191, 203, 201, 195 2.03.07 State the purpose of a Chi-squared test. CHI-SQUARED TEST The Chi-squared test (pronounced "ki") is used to determine the precision of a counting system. Precision is a measure of exactly how a result is determined without regard to its accuracy. It is a measure of the reproducibility of a result, or in other words, how often that result can be repeated, or how often a "success" can be obtained. This test results in a numerical value, called the Chi-squared value (X 2 ), which is then compared to a range of values for a specified number of observations or trials. This range represents the 2 expected (or predicted) probability for the chosen distribution. If the X value is lower than the expected range, this tells us that there is not a sufficient degree of randomness in the observed data. If the value is too high, it tells us that there is too much randomness in the observed data. DOE-HDBK-1122-2009 (Revised 2013) Module 2.03 Counting Errors and Statistics Student's Material The Chi-squared test is often referred to as a "goodness-of-fit" test. If it does NOT fit a curve

indicating sufficient randomness, then the counting instrument may be malfunctioning. The Chi-squared value is calculated as follows: ∑ ( x − x) 2 i (Equation 5) X 2 = x Example 2.03-2 Using the data from Example 2.03-1, determine the Chi-squared value for the data set. 2.03.08 State the criteria for acceptable Chi-squared values at your site. (Insert site-specific information here.) Assuming a given set of data passes the Chi-squared test, the data can then be used to prepare quality control charts for use in verifying the consistent performance of the counting system. 2.03.09 State the purpose of creating quality control (QC) charts. 2.03.10 State the requirements for maintenance and review of QC charts at QUALITY CONTROL CHARTS Quality control charts may be prepared using source counting data obtained during system calibration. Obviously since this test verifies that the equipment is still operating within an expected range of response, we cannot change the conditions of the test in mid-stream. QC charts, then, enable us to track the performance of the system while in use. Data that can be used for quality control charts include gross counts, counts per unit time, and efficiency. Most nuclear laboratories use a set counting time corresponding to the normal counting time for the sample geometry being tested. DOE-HDBK-1122-2009 (Revised 2013) Module 2.03 Counting Errors and Statistics Student's Material When the system is calibrated and the initial calculations performed, the numerical values of the mean ± 1, 2, and 3 standard deviations are also determined. Using graph paper or a computer graphing software, lines are drawn all the way across the paper at those points corresponding to the mean, the mean plus 1, 2, and 3 standard deviations, and the mean minus 1, 2, and 3 standard deviations. The daily control check results are then plotted on the control chart to see if the results fall outside the limits (Insert site-specific information here.) 2.03.11 State the purpose of calculating warning and control limits. SYSTEM OPERATING LIMITS The values corresponding to ±2 and ±3 standard deviations may be called the upper and lower warning and control limits, respectively (or other terms such as action limits) . The results of the daily source counts are graphed in many countrooms. Most of the time results will lie between the lines corresponding to ±1 standard deviation (68.2%). We also know that 95.4% of the time our count will be between ±2 standard deviations and that 99.97% of the time our count will be between ±3 standard deviations. Counts that fall outside the warning limit (±2 σ ) are not necessarily incorrect. Statistical distribution models say that we should get some counts in that area. Counts outside the warning limits indicate that something MAY be wrong. The same models say that we will also get some outside the control limits (±3 σ ). However, not very many measurements will be outside those limits. We use 3 σ as the control – a standard for acceptable performance. In doing so we say that values outside of ±3 σ indicate unacceptable performance, even though those values may be statistically valid. True randomness also requires that there be no patterns in the data that are obtained; some will be higher than the mean, some will be lower, and some will be right on the mean. When patterns do show up in quality control charts, they are usually indicators of systematic error. For example:

• Multiple points outside two sigma • Repetitive points (n out of n) outside one sigma • Multiple points, in a row, on the same side of the mean DOE-HDBK-1122-2009 (Revised 2013) Module 2.03 Counting Errors and Statistics Student's Material • Multiple points, in a row, going up or down. The assumption is made that systematic error is present in our measurements, and that our statistical analysis has some potential for identifying its presence. However, industry assumption is that systematic error that is present is very small in comparison to random error. Quality control charts should be maintained in the area of the radioactivity counting system such that they will be readily accessible to those who operate the system. These charts can then be used by operators to determine if routine, periodic checks (typically daily) have been completed before system use. 2.03.12 State the purpose of determining efficiencies and correction factors. 2.03.13 Given counting data and source assay information, calculate efficiencies and correction factors. COUNTER EFFICIENCY A detector intercepts and registers only a fraction of the total number of radiations emitted by a radioactive source. The major factors determining the fraction of radiations emitted by a source that are detected include: • The fraction of radiations emitted by the source which travel in the direction of the detector window • The fraction emitted in the direction of the detector window which actually reach the window • The fraction of radiations incident on the window which actually pass through the window and produce an interaction • The fraction scattered into the detector window All radiation detectors will, in principle, produce an output pulse for each particle or photon which interacts within its active volume. Because of the factors outlined above, only a fraction of the disintegrations occurring in a source result in counts being reported by the detector. Therefore, there is only a certain fraction of the disintegrations occurring that results in counts reported by the detector. Using a calibrated source with a known activity, a precise figure can be determined for this fraction. This value can then be used as a ratio in order to relate the number of pulses counted to the number of particles and/or photons emitted by the source. This ratio is called the efficiency. DOE-HDBK-1122-2009 (Revised 2013) Module 2.03 Counting Errors and Statistics Student's Material The detector efficiency gives us the fraction of counts detected per disintegration, or c/d. Since activity is the number of disintegrations per unit time, and the number of counts are detected in a finite time, the two rates can be used to determine the efficiency if both rates are in the same units of time. Counts per minute (cpm) and disintegrations per minute (dpm) are the most common. Thus, the efficiency, E, can be determined as shown in Equation 6. Used in this manner the time units will cancel, resulting in counts/disintegration (c/d). cpm c (Equation 6) E = = dpm d The efficiency obtained in the formula above will be in fractional or decimal form. To calculate the percent efficiency, the fraction can be multiplied by 100. For example, an efficiency of 0.25 would mean 0.25 × 100, or 25%. Example 2.03-3 A source is counted and yields 2840 counts per minute. If the source activity is known to be 12,500 dpm calculate the efficiency and percent efficiency.

By algebraic manipulation, Equation 6 can be solved for the disintegration rate (see Equation 7). The system efficiency is determined as part of the calibration. When analyzing samples, a count rate is reported by the counting system. Using Equation 7, the activity, (A), of the sample can be determined in dpm, and then converted to any other units of activity (e.g., Ci, Bq). cpm cpm (Equation 7) dpm = E ⎯⎯→ dpm A = E Example 2.03-4 A sample is counted on a system with a 30% efficiency. If the detector reports 4325 net counts per minute what is the activity of the sample in dpm? A correction factor (CF), which is simply the inverse of the efficiency, is used by multiplying it by the net count rate to determine the activity, as in Equation 8. DOE-HDBK-1122-2009 (Revised 2013) Module 2.03 Counting Errors and Statistics Student's Material 1 (Equation 7) CF = E Example 2.03-5 An instrument has an efficiency of 18%. What is the correction factor? This count-rate correction factor should not be confused with a geometry correction factor used with some radiation instruments, such as the beta correction factor for a Cutie Pie (RO-3C). 2.03.14 State the meaning of counting data reported as x.xx ± yy. 2.03.15 Given counting results and appropriate formulas, report results to desired confidence level. ERROR CALCULATIONS The error present in a measurement governed by a statistical model can be calculated using known parameters of that model. Nuclear laboratories are expected to operate at a high degree of precision and accuracy. However, since we know that there is some error in our measurements, we are tasked with reporting measurements to outside agencies in a format that identifies that potential error. The format that is used should specify the activity units and a range in which the number must fall. In other words, the results would be reported as a given activity plus or minus the error in the measurement. Since nuclear laboratories prefer to be right more than they are wrong, counting results are usually reported in a range that would be correct 95% of the time, or at a 95% confidence level. In order to do this, the reported result should be in the format: (Equation 9) x.xx ± yy (K σ ) where: x.xx = measured activity, in units of dpm, Ci, or Bq yy = associated potential (or possible) error in the measurement K = multiple of counting error σ = standard deviation at stated confidence level (CL) DOE-HDBK-1122-2009 (Revised 2013) Module 2.03 Counting Errors and Statistics Student's Material Note: Use of K σ is only required for confidence levels other than 68% (see Table 4). Therefore: σ = 1× σ 68% CL (optional) 1.64 σ = 1.64 × σ 90% CL (sometimes used) 2 σ = 1.96 × σ 95% CL (normally used) For example, a measurement of 150 ± 34 dpm (2 σ ) indicates the activity as 150 dpm; however, it could be as little as 116 dpm or as much as 184 dpm with 95% confidence (at 2 σ ). The calculations of the actual range of error are based on the standard deviation for the distribution. In the normal (or Gaussian) distribution, the standard deviation of a single count is defined as the square root of the mean, or σ = x . The error, e, present in a single count is some multiplier, K, multiplied by the square root of that mean, i.e., some multiple times the standard deviation, K σ . The value of K used is based on the confidence level that is desired, and is derived from the area of the curve included at that confidence level (see Figure 5).

Common values for K include: Table 4. Counting Error Multiples Error Confidence Level K Probable 50% 0.6745 Standard 68% 1.0000 9/10 90% 1.6449 95/100 95% 1.9600 99/100 99% 2.5750 To calculate the range to the point at which you would expect to be right 95% of the time, you would multiply the standard deviation by 1.96, and report the results of the measurement as x.xx dpm ± yy dpm (2 σ ). Note that using a 68% or 50% confidence level introduces an expected error a large percentage of the time. Therefore, for reasonable accuracy a higher confidence level must be used. The simple standard deviation (σ) of the single count (x) is usually determined as a count rate (counts per unit time). This is done by dividing the count rate (R) by the count time (T). Subscripts can be applied to distinguish sample count rates from background count rates. (Equation 10) σ = K R T DOE-HDBK-1122-2009 (Revised 2013) Module 2.03 Counting Errors and Statistics Student's Material Example 2.03-6 The count rate for a sample was 250 cpm. Assume 10 minute counting time, zero background counts and a 25% efficiency. Report sample activity at a 95% C.L. 2.03.16 State the purpose of determining background. BACKGROUND Determination of Background Radioactivity measurements cannot be made without consideration of the background. Background, or background radiation, is the radiation that enters the detector concurrently with the radiation emitted from the sample being analyzed. This radiation can be from natural sources, either external to the detector (i.e., cosmic or terrestrial) or radiation originating inside the detector chamber that is not part of the sample. In practice, the total counts are recorded by the counter. This total includes the counts contributed by both the sample and the background. Therefore, the contribution of the background will produce an error in radioactivity measurements unless the background count rate is determined by a separate operation and subtracted from the total activity, or gross count rate. The difference between the gross and the background rates is called the net count rate (sometimes given units of ccpm, or corrected counts per minute). This relationship is seen in the following equation: (Equation 11) R = R − RS S +B B where: RS = net sample count rate (cpm) RS+B = gross sample count rate (cpm) RB = background count rate (cpm) The background is determined as part of the system calibration by counting a background (empty) sample holder for a given time. The background count rate is determined in the same way as any count rate, where the gross counts are divided by the count time, as seen in Equation 12 below. (Equation 12) R B = N B TB where: RB = background count rate (counts per time, i.e., cpm) DOE-HDBK-1122-2009 (Revised 2013) Module 2.03 Counting Errors and Statistics Student's Material NB = gross counts, background TB = background count time In practice, background values should be kept as low as possible. As a guideline, background on automatic counting systems should not be allowed to exceed 0.5 cpm alpha and 1 cpm beta-gamma. If system background is above this limit the detector should be cleaned or replaced. Reducing Background Typically, the lower the system background the more reliable the analysis of samples will be. In low-background counting systems the detector housing is surrounded by lead shielding so as to reduce the background. Nonetheless, some background still manages

to reach the detector. Obviously, little can be done to reduce the actual source of background due to natural sources. On many systems a second detector is incorporated to detect penetrating background radiation. When a sample is analyzed the counts detected by this second detector during the same time period are internally subtracted from the gross counts for the sample. Background originating inside the detector chamber can be, for the most part, more easily controlled. The main contributors of this type of background are: • Radiation emitted from detector materials • Radioactive material on inside detector surfaces • Radioactive material on the sample slide assembly • Contamination in or on the sample planchet or planchet carrier There are, unfortunately, trace levels of radioactivity in the materials of which detectors and their housings are made. This is simply a fact of life in the atomic age. The contribution to background from such materials is negligible, but should nonetheless be acknowledged. Radioactive material can be transferred from contaminated samples to the inside surfaces of the detector chamber during counting. This usually occurs when samples having gross amounts of material on them are counted in a low-background system. During the insertion and withdrawal of the sample into the detector chamber, loose material can be spread into the chamber. In order to prevent this, these samples should be counted using a field survey instrument or a mini-scaler. Low-background systems are designed for counting lower-activity samples. Counting of a high-activity sample on these systems should be avoided unless it is a sealed radioactive source. DOE-HDBK-1122-2009 (Revised 2013) Module 2.03 Counting Errors and Statistics Student's Material (Insert site-specific information here.) Note: The following is provided as an example as a typical counting system. Radioactive material can also be transferred from contaminated samples to the slide assembly upon which samples are inserted into, and withdrawn from, the detector chamber. This can be prevented in the same way as stated above. In addition, when loading and stacking samples for counting, ensure that the slide assembly cover is in place. The slide assembly should also be cleaned on a routine basis (i.e., weekly). When loading and unloading samples into and from planchets, material from the samples can be spread to the planchet and/or the carrier. Most smears and air samples are 47-mm diameter and are counted in a planchet that is almost the same size. The planchet is placed in a carrier which surrounds and supports the planchet and allows for automatic sample exchange by the counting system. When a sample is counted, the entire carrier is placed under the detector window inside the detector chamber. Any contamination on the carrier (or in the planchet) is counted with, and attributed to, the sample. A paper disc can be placed in the bottom of the planchet as a step in preventing transfer of material from samples to the planchet. Care should be taken when loading and unloading samples such that material remains on the sample media. 2.03.17 State the method and requirements for determining background for counting systems at your site. (Insert site-specific information here.) 2.03.18 State the purpose of performing planchet maintenance. PLANCHET MAINTENANCE Planchets and carriers should be inspected, cleaned, and counted on a routine basis. All in-use

planchets and carriers must read less than established site limits. Planchets exceeding these limits should be decontaminated and recounted as necessary. By maintaining planchets clean and as free from contamination as possible, sample result reliability will be increased because the amount of error introduced in the sample analysis will be reduced. 2.03.19 State the method and requirements for performing planchet maintenance for counting systems at your site. DOE-HDBK-1122-2009 (Revised 2013) Module 2.03 Counting Errors and Statistics Student's Material PROPAGATION OF ERROR The error present in a measurement includes the error present in the sample count and the error present in the background count. The total error in the measurement is calculated by squaring the error in the background and adding that to the square of the error in the sample count, and taking the square root of the sum, as shown in Equation 13. (Equation 13) e = e 2 + e 2 S S + B B where: eS = error present in the measurement (sample) eS+B = error in sample count (sample plus background) eB = error present in background count Since we normally use this equation in terms of a count rate, the formula is slightly modified as follows, and the error stated as the sample standard deviation (σ S): R RS +B B(Equation 14) Kσ S = K + TS TB where: RS+B = gross sample count rate (sample plus background) RB = background count rate TS = sample count time TB = background count time K = confidence level multiple (see Table 4) The error in the sample count is the standard deviation of the count, which is the square root of that count (see Equation 13 above). Example 2.03-7 An air sample is counted and yields 3500 counts for a 2-minute count period. The system background is 10 cpm determined over a 50-minute count time. Determine the error in the sample and report the net count rate to 95% confidence level. If the sample counting time and the background counting time is the same, the formula can be simplified even more to: DOE-HDBK-1122-2009 (Revised 2013) Module 2.03 Counting Errors and Statistics Student's Material R + RS +B B(Equation 15) Kσ = K s T Example 2.03-8 A long-lived sample is counted for one minute and gives a total of 562 counts. A one minute background gives 62 counts. Report net sample count rate to 95% CL. 2.03.20 Explain the methods used to improve the statistical validity of count room measurements. IMPROVING STATISTICAL VALIDITY OF COUNT ROOM MEASUREMENTS Minimizing the statistical error present in a single sample count is limited to several options. If we look at the factors present in the calculation below (same as Equation 14), we can see that there are varying degrees of control over these factors. The standard deviation is calculated here in terms of count rate. R RS +B Bσ rate = + TS TB RS+B is the sample count rate. We really have no control over this. RB is the background count rate. We do have some control over this. On any counting equipment the background should be maintained as low as possible. In most of our counting applications, however, the relative magnitude of the background count rate should be extremely small in comparison to the sample count rate if proper procedures are followed. This really becomes an issue when counting samples for free release or environmental samples. However, some reduction in error can be obtained by increasing the background counting time, as

discussed below. TB and TS are the background and sample counting times, respectively. These are the factors that we have absolute control over. In the previous section we talked about the reliability of the count itself. We have been able to state that a count under given circumstances may be reproduced with a certain confidence level, and that the larger the number of counts the greater the reliability. The condition we have been assuming is that our count is taken within a given time. In order to get more precise results, many counts must be observed. Therefore, if we have low count rates, the counting time must be increased in order to obtain many counts, thereby making the result more precise (or reproducible). DOE-HDBK-1122-2009 (Revised 2013) Module 2.03 Counting Errors and Statistics Student's Material The total counting time required depends upon both the sample and background count rates. For high sample activities the sample count time can be relatively short compared to the background count time. For medium count rates we must increase the sample count time in order to increase precision. As the sample activity gets even lower, we approach the case where we must devote equal time to the background and source counts. In other words, by counting low activity samples for the same amount of time as that of the background, we increase the precision of our sample result. However, we must never count a sample for a period of time longer than the system background. In summary, by minimizing the potential error present, we improve statistical validity of our measurements. 2.03.21 Define "detection limit," and explain the purpose of using detection limits in the analysis of radioactive samples. DETECTION LIMITS The detection limit of a measurement system refers to the statistically determined quantity of radioactive material (or radiation) that can be measured (or detected) at a preselected confidence level. This limit is a factor of both the instrumentation and technique/procedure being used. The two parameters of interest for a detector system with a background response greater than zero are (see Figure 6): LC Critical detection level: the response level at which the detector output can be considered "above background" LD Minimum significant activity level, i.e., the activity level that can be seen with a detector with a fixed level of certainty These detection levels can be calculated by the use of Poisson statistics, assuming random errors and systematic errors are separately accounted for, and that there is a background response. For these calculations, two types of statistical counting errors must be considered quantitatively in order to define acceptable probabilities for each type of error: Type I - occurs when a detector response is considered above background when in fact it is not (associated with LC) Type II - occurs when a detector response is considered to be background when in fact it is greater than background (associated with LD) 2.71 DOE-HDBK-1122-2009 (Revised 2013) Module 2.03 Counting Errors and Statistics Student's Material If the two probabilities (areas labeled I and II in Figure 6) are assumed to be equal, and the background of the counting system is not well-known, then the critical detection level (LC) and the minimum significant activity level (LD) can be calculated. The two values would be derived using the equations LC = kσB and LD = k 2

+ 2kσB, respectively. If 5% false positives (Type I error) and 5% false negatives (Type II error) are selected to be acceptable levels, i.e., 95% confidence level, then k = 1.645 and the two equations can be written as: RB RB(Equation 16) LC = 1.645 + TB TS R R3 B B(Equation 17) LD = + 3.29 + TS TB TS where: LC = Critical detection level LD = a priori detection limit [minimum significant activity level RB = background count rate TB = background count time TS = sample count time 2.71 DOE-HDBK-1122-2009 (Revised 2013) Module 2.03 Counting Errors and Statistics Student's Material The minimum significant activity level, LD, is the a priori (before the fact) activity level that an instrument can be expected to detect 95% of the time. In other words, it is the smallest amount of activity that can be detected at a 95% confidence level. When stating the detection capability of an instrument, this value should be used. The critical detection level, LC, is the lower bound on the 95% detection interval defined for LD, and is the level at which there is a 5% chance of calling a background value "greater than background." This value (LC) should be used when actually counting samples or making direct radiation measurements. Any response above this level should be counted as positive and reported as valid data. This will ensure 95% detection capability for LD. If the sample count time (TS) is the same as the background count time (TB), then equations 16 and 17 can be simplified as follows: (Equation 18) LC = 2.32 RB T R (Equation 19) LD = 3 + 4.65 B T T where: T = count time (sample and background) Therefore, the full equations for LC and LD must be used for samples with count times differing from the background determination time (95% CL used). These equations assume that the standard deviation of the sample planchet/carrier background during the sample count (the planchet/carrier assumed to be 0 activity) is equal to the standard deviation of the system background (determined using the background planchet/carrier). The critical detection level, LC, is used when reporting survey results. It is used to say that at a 95% confidence level, samples above this value are radioactive. This presupposes, then, that 5% of the time clean samples will be considered radioactive. The minimum significant activity level, LD, [also referred to as the LLD (Lower Limit of Detection) in some texts] is calculated prior to counting samples. This value is used to determine minimum count times based on release limits and airborne radioactivity levels. In using this value we are saying that at a 95% CL, samples counted for at least the minimum count time calculated using the LD that are positive will indeed be radioactive (above the LC). This presupposes, then, that 5% of the time samples considered clean will actually be radioactive. Example 2.03-9 A background planchet is counted for 50 minutes and yields 16 counts. Calculate the critical detection level and the minimum significant activity level for a 0.5 minute sample count time. DOE-HDBK-1122-2009 (Revised 2013) Module 2.03 Counting Errors and Statistics Student's Material 2.03.22 Given the formula and necessary information, calculate detection limit values for counting systems at your site. MINIMUM DETECTABLE ACTIVITY (MDA) The minimum significant activity level, LD , can be used to evaluate whether the measurement

process is adequate to meet requirements. For example, the results of Example 2.03-9 (in units of cpm) can be converted using counting efficiency and area of swipe to determine the adequacy of the measurement system for contamination surveys for removable contamination performed to ensure that the removable surface contamination values specified in Appendix D of 10 CFR 835 are not exceeded (in units of dpm/100cm 2 ). In most cases, swipes to determine the removable contamination levels will be counted in the field or submitted to the counting lab for analysis where the background radiation levels are sufficiently low enough to ensure that LD, and thus MDA, are below the limits in 10 CFR 835 Appendix D. (Equation 19b) MDAremovable = LD x (100/Aswipe) / e where: MDAremovable = Activity level in dpm/100 cm 2 e = Detector efficiency in counts per disintegration Aswipe = Area of surface swiped in cm 2 Static Count MDAs Similarly, one can calculate a MDA for a static field measurement to evaluate total surface contamination values. In this case, an adjustment is needed to account for the size of the detector. To determine the MDA for static counts, i.e., the probe is stationary for a prescribed period of time, Equation 19c is used. (Equation 19c) MDAtotal = LD x (100/Aprobe) / e where: MDAtotal = Activity level in dpm/100 cm 2 e = Detector efficiency in counts per disintegration Aprobe = Surface area of probe in cm 2 Scanning MDA DOE-HDBK-1122-2009 (Revised 2013) Module 2.03 Counting Errors and Statistics Student's Material The ability to identify a small region or area of slightly elevated radiation during surface scanning is dependent upon the RCT’s skill in recognizing an increase in the audible output of the instrument. Experience has shown that a 25% to 50% increase may be easily identifiable at ambient background levels of several thousand counts per minute, whereas, at ambient levels of a few counts per minute, a two to three fold increase in the audible signal is required before a change is readily recognizable. The detection sensitivity of scanning is dependent upon a number of other factors, such as detector scan speed, surface characteristics, size of elevated activity region, surveyor efficiency, level of activity, and detector/surface distance. The ability to detect an elevated region of activity using a particular survey scanning technique would need to be determined empirically and is beyond the scope of this training. (Insert site-specific information here.) 2.03.23 State the purpose and method of determining crosstalk. CROSSTALK Discrimination Crosstalk is a phenomenon that occurs on proportional counting systems (such as a Tennelec) that employ electronic, pulse-height discrimination, thereby allowing the simultaneous analysis for alpha and beta-gamma activity. Discrimination is accomplished by establishing two thresholds, or windows, which can be set in accordance with the radiation energies of the nuclides of concern. Recall that the pulses generated by alpha radiation will be much larger than those generated by beta or gamma. This makes the discrimination between alpha and beta-gamma possible. Beta and gamma events are difficult to distinguish; hence, they are considered as one by such counting systems. In practice, the lower window is set such that electronic noise and ultra-low-energy photon events are filtered out. Any pulse generated whose size is greater than the setting

for the lower window is considered an event, or a count. The upper window is then set such that any pulses which surpass the upper discriminator setting will be considered an alpha count (see Figure 7). DOE-HDBK-1122-2009 (Revised 2013) Module 2.03 Counting Errors and Statistics Student's Material For output purposes, the system routes each count to a series of channels which simply keep a total of the counts routed to them. Channel A is for alpha counts, Channel B is beta-gamma counts, and Channel C is total counts. As a sample is being counted, all valid counts registered (i.e., those which surpass the lower discriminator setting) are routed to the C-channel. In addition, if the count was considered an alpha count (i.e., it surpassed the upper discriminator setting) it is routed to the A-channel; else it is tallied in the B-channel. In effect, what occurs is that the number of beta-gamma counts (Channel B) are determined by subtracting the number of alpha counts (Channel A) from the total counts (Channel C), or B = C - A. Origin of Crosstalk Now that we understand the process involved, there is a dilemma that stems from the fact that events are identified by the system as either alpha or beta-gamma according to the size of the pulse generated inside the detector. The system cannot really tell what type of radiation has generated the pulse. Rather, the pulse is labeled as "alpha" or "beta­ gamma" by comparing the size of the pulse to the discriminator setting. It is the setting of the discriminator that poses the dilemma. Alpha particles entering the detector chamber generally are attenuated by the detector fill-gas because of their high LET, thereby producing a large pulse. Low-energy beta particles and photons will also lose all their energy within the detector gas, but DOE-HDBK-1122-2009 (Revised 2013) Module 2.03 Counting Errors and Statistics Student's Material nevertheless produce a smaller pulse because of their lower energies. High energy beta particles can still retain some of their energy even after having produced a pulse while traversing the detector volume. Rather than leaving the detector, as would a photon, the beta is reflected off of the detector wall and reenters the volume of gas, causing ionizations and generating a second pulse. These two pulses can be so close together that the detector sees them as one large pulse. Because of the large pulse size it can surpass the upper discriminator setting and is, therefore, counted as an alpha, and not as a beta. The result is that alpha activity can be reported for a sample when in fact there was little or no alpha present. Conversely, if a true alpha-generated pulse is not large enough so as to exceed the upper discriminator, it would be counted as a beta-gamma event. This is crosstalk. The solution is not a simple one. The setting of the upper discriminator depends on the radiations and energies of the sources and samples being analyzed. If high energy beta radiations are involved, a significant portion of them could be counted as alpha events if the setting is too low. If the setting is too high, lower-energy alpha events could be counted as beta-gamma. Typically, the setting of the discriminator will usually be some "happy medium." A discussion of how this can be dealt with is in order. Calibration Sources and Crosstalk For calibrations of Tennelec counting systems, the manufacturer provides the following

general recommendations for discriminator settings: First, using a Strontium-90 beta source, set the upper (A) discriminator such that there is 1% beta-to-alpha crosstalk. Then, using a Polonium-210 alpha source, set the A+B discriminator such that there is less than 3% alpha-to-beta crosstalk. Energies of sources used to calibrate counting systems should be the same as, or as close as possible to, the energies of radionuclides in the samples analyzed. Wherever possible they should be a pure emitter of the radiation of concern. For beta-gamma sources the most popular isotope in radiation protection is Sr-90. It has a relatively long half-life of 29.1 years, but emits betas of only 546 keV. However, Sr-90 decays to Yttrium-90, another beta emitter which has a short half-life of only 2.67 days and emits a 2.281 MeV beta. Y-90 decays to Zirconium-90m which emits a 2.186 Mev gamma almost instantaneously to become stable. The daughters reach equilibrium with the strontium parent within a number of hours after source assay. Hence, for every Sr beta emitted a Y beta is also emitted, thereby doubling the activity. These sources are often listed as Sr/Y-90 for obvious reasons. This makes Sr/Y-90 sources an excellent choice and they are used by many sites for calibrations and performance testing. Po-210 is essentially a pure alpha emitter. This is primarily the reason why it is recommended for calibrations and performance testing. It yields a strong alpha, but it also has a short half-life. A comparison of some alpha emitters is given in Table 5. DOE-HDBK-1122-2009 (Revised 2013) Module 2.03 Counting Errors and Statistics Student's Material Table 5. Alpha Emitters Isotope Half-Life Energy (MeV) Po-210 138.38 days 5.3044 Pu-239 2.4E4 years 5.156, 5.143, 5.105 Ra-226 1.60E3 years 4.78, 4.602 Th-230 7.54E4 years 4.688, 4.621 Natural U 4.4E9 years (avg.) 4.2 (avg.) 2.03.24 State the criteria for acceptable values of crosstalk for counting systems at your site (Insert site-specific information here.) VOLTAGE PLATEAUS Very simply put, a voltage plateau is a graph that indicates a detector's response to a specific energy particle with variations of high voltage. The x-axis represents the high voltage and the y-axis the response (i.e., counts). The resulting curve gives an indication of detector quality, and can indicate problems with the counting gas should they be present. The curve can also be used to determine the optimum operating high voltage for the system. Most automatic low-background counting systems provide several different analysis modes. These modes count samples at certain pre-determined voltages. Counting systems generally provide three analysis modes: • ALPHA ONLY • ALPHA THEN BETA • ALPHA AND BETA (SIMULTANEOUS) There are usually two voltage settings used in conjunction with these analysis modes: • Alpha voltage (lower) • [Alpha plus] Beta voltage (higher) DOE-HDBK-1122-2009 (Revised 2013) Module 2.03 Counting Errors and Statistics Student's Material Recall that in a proportional counter the amount of voltage determines the amount of gas multiplication. Because of the high LET of alpha radiation, at a lower voltage, even though the gas amplification will be lower, alpha pulses will still surpass the lower discriminator and some will even pass the upper discriminator. Because of the lower gas amplification beta-gamma pulses will not be large enough to be seen. Therefore, any counts reported for the sample will be

alpha counts. In the ALPHA ONLY mode, the sample is counted once, at the alpha voltage. Counts may appear in either the A or B channels. Upon output, the A and B channels will be added together and placed in Channel A and, therefore, reported as alpha counts; the B channel will be cleared to zero, thereby resulting in no beta-gamma counts. In the ALPHA THEN BETA mode, the sample is counted twice. The first count interval determines the alpha counts using the alpha voltage. The second count is done at the beta voltage. The determination of alpha and beta-gamma counts in this mode is based strictly on the operating characteristics of the detector at the different voltages. For this reason, the A and B counts are summed during both counting intervals to attain the total counts. The separation of alpha and beta-gamma counts is then calculated and reported according to the following formula: A + B α = 1 1 (Equation 20) CF α β = ( A − B ) − α2 2 where: α = reported gross alpha counts β = reported gross beta-gamma counts A1,B1 = accumulated channel counts respectively, 1st interval A2,B2 = accumulated channel counts respectively, 2nd interval CFα = alpha correction factor (ratio of alpha efficiency at alpha voltage to efficiency at beta voltage) In the ALPHA AND BETA (SIMULTANEOUS) mode, the sample is counted once using the beta voltage. Alpha events are reported in the A channel, while beta-gamma counts are reported in the B channel. This is the mode used most often. As can be seen, the setting of the two voltages will have a direct impact on the number of counts reported for a given sample. The determination of what these voltage settings should be must be done such that the optimum performance of the detector is obtained for those voltage regions. This is the purpose of a plateau. 2.03.26 State the method of performing a voltage plateau on counting systems at your site. DOE-HDBK-1122-2009 (Revised 2013) Module 2.03 Counting Errors and Statistics Student's Material (Insert site-specific information here.) In conjunction with initial system setup and calibration by the vendor, two voltages plateaus are performed--alpha voltage and beta voltage. For P-10 gas the alpha plateau is started at about 400 volts and the beta plateau at about 900 volts. The plateaus are developed by plotting the gross counts accumulated for each radionuclide. Each time that a count is completed, the high voltage is incremented a specific amount, typically 25 to 50 volts, and another count is accumulated. This is repeated until the end of the range is reached, typically about 1800 volts. With the high voltage set at the starting point, few or no counts are observed because of insufficient ion production within the detector. As the voltage is increased, a greater number of pulses are produced with sufficient amplitude to exceed the discriminator threshold, and are then accumulated in the counter. There will be a high voltage setting where the increase in counts levels off (see Figure 8). This area is the detector plateau. Further increases in high voltage result in little change in the overall count rate. The plateau should remain flat for at least 200 volts using a Sr/Y-90 source, and this indicates the plateau length. Between 1750 and 1850 volts the count rate will start to increase dramatically. This is the avalanche region, and the high voltage should not be increased any further. DOE-HDBK-1122-2009 (Revised 2013)

Module 2.03 Counting Errors and Statistics Student's Material The region where the counts level off is called the knee of the plateau. The operating voltage is chosen by viewing the plateau curve and selecting a point 50 to 75 volts above the knee and where the slope per 100 volts is less than 2.5%. This ensures that minor changes in high voltage will have negligible effects on the sample count. Poor counting gas or separation of the methane and argon in P-10 can result in a very high slope of the plateau. Upon initial system setup and calibration the vendor determines and sets the optimum operating voltages for the system. Thereafter, plateaus should be generated each time the counting gas is changed. SUMMARY This lesson addressed the measures used to minimize error and the fundamentals of binomial statistics, as well as the application of these fundamentals in a nuclear counting environment. Completion of the unit does not qualify the student to perform any tasks independently. DOE-HDBK-1122-2009 Module 2.04 Dosimetry Study Guide 2.04-1 Course Title: Radiological Control Technician Module Title: Dosimetry Module Number: 2.04 Objectives: 2.04.01 Identify the DOE external exposure limits for general employees. 2.04.02 Identify the DOE limits established for the embryo/fetus of a declared pregnant female general employee. → 2.04.03 Identify the administrative exposure control guidelines at your site, including those for the: a. General employee b. Member of the public/minor c. Incidents and emergencies d. Embryo/fetus → 2.04.04 Identify the requirements for a female general employee who has notified her employer in writing that she is pregnant. 2.04.05 Determine the theory of operation of a thermoluminescent dosimeter (TLD). 2.04.06 Determine how a TLD reader measures the radiation dose from a TLD. 2.04.07 Identify the advantages and disadvantages of a TLD compared to a film badge. → 2.04.08 Identify the types of beta-gamma TLDs used at your site. → 2.04.09 Identify the types of neutron TLDs used at your site. → 2.04.10 Determine the requirements for use of TLDs used at your site. → 2.04.11 Determine the principle of operation, and the types used, for the personnel neutron dosimeters used at your site. → 2.04.12 Determine the principle of operation of self-reading dosimetry (SRD) used at your site. → 2.04.13 Determine the principle of operation, and guidelines for use, for the alarming dosimeters used at your site. → 2.04.14 List the types of bioassay monitoring methods at your site. DOE-HDBK-1122-2009 Module 2.04 Dosimetry Study Guide 2.04-2 INTRODUCTION Radiation dosimetry is the branch of science that attempts to quantitatively relate specific measures made in a radiation field to chemical and/or biological changes that the radiation would produce in a target. Dosimetry is essential for quantifying the incidence of various biological changes as a function of the amount of radiation received (dose-effect relationships), for comparing different experiments, for monitoring the radiation exposure of individuals, and for surveillance of the environment.

External dosimetry is the science dealing with the measurement of a radiation field incident to the body and the evaluation of the equivalent dose resulting from energy deposited within the body by radiation. External dose is usually a derived or inferred quantity since it is not possible to directly measure the exact dose to any organ or tissue. Any measurement must be compared to a known quantity to derive dose and equivalent dose. This process is called "calibration". Internal dosimetry is the analysis and measurement of radionuclides in humans or bioassay samples and the evaluation of intakes and doses from those measurements. It involves evaluation of bioassay data, evaluation of the intake, distribution, retention, and elimination of radionuclides, and evaluation of various absorbed doses and equivalent dose quantities. Internal dosimetry is inherently indirect in nature. It is not possible to determine the exact organ absorbed dose, equivalent dose or effective dose in a living human being resulting from an intake of radioactive materials. Internal dose is usually a derived or inferred quantity, obtained by evaluation of indirect measurements and computational models. This is particularly true for alpha- and beta- emitting radionuclides in the body which have low photon emission abundances. Direct measurements of internalized photon-emitting radionuclides in organs also may be difficult because of attenuation and scattering by overlying tissues. The capability to accurately measure and analyze radioactive materials and workplace conditions, and determine personnel radiation exposure is fundamental to the safe conduct of radiological operations. Accordingly, DOE shall ensure radiological measurements, analyses, worker monitoring results and estimates of public exposures are accurate and appropriately made. 10 CFR 835 prescribes the requirements for both external and internal dose monitoring. It is the responsibility of all workers to wear personnel monitoring devices where required by Radiological Work Permits, signs, procedures or by radiological control personnel. They are also expected to report immediately the loss, damage or unexpected exposure of personnel monitoring devices or off-scale readings of self- reading dosimeters to the Radiological Control Organization (RCO). All employees are expected to keep track of their radiation exposure status and avoid exceeding radiological Administrative Control Levels. Additionally, all should notify the RCO DOE-HDBK-1122-2009 Module 2.04 Dosimetry Study Guide 2.04-3 of off-site occupational radiation exposures so that worker dosimetry records can be updated. References: 1. "Basic Radiation Protection Technology"; Gollnick, Daniel; 5th ed.; Pacific Radiation Corporation; 2008. 2. ANL-88-26 (1988) "Operational Health Physics Training"; Moe, Harold; Argonne National Laboratory, Chicago. 3. "DOE Radiological Control Standard"; U.S. Department of Energy, 2008. 4. 10 CFR Part 835 (2007) "Occupational Radiation Protection". DOE-HDBK-1122-2009 Module 2.04 Dosimetry Study Guide 2.04-4 DOSIMETRY TERMS Absorbed Dose (D): Energy absorbed by matter from ionizing radiation per unit mass of irradiated material at the place of interest in that material. The absorbed dose is expressed in units of rad (or gray) (1 rad = 0.01 gray).

Equivalent Dose (HT): The product of average absorbed dose (DT,R) in rad (or gray) in a tissue or organ (T) and a radiation (R) weighting factor (wR). For external dose, the equivalent dose to the whole body is assessed at a depth of 1 cm in tissue; the equivalent dose to the lens of the eye is assessed at a depth of 0.3 cm in tissue, and the equivalent dose to the extremity and skin is assessed at a depth of 0.007 cm in tissue. Equivalent dose is expressed in units of rem (or Sv). Whole Body: For the purposes of external exposure, head, trunk (including male gonads), arms above and including the elbow, or legs above and including the knee. Extremity: Hands and arms below the elbow or feet and legs below the knee. Committed Equivalent Dose (HT,50): The equivalent dose calculated to be received by a tissue or organ over a 50-year period after the intake of a radionuclide into the body. It does not include contributions from radiation sources external to the body. Committed equivalent dose is expressed in units of rem (or Sv). Radiation Weighting Factor (wR ): A modifying factor used to calculate the equivalent dose from the average tissue or organ absorbed dose; the absorbed dose (expressed in rad or gray) is multiplied by the appropriate radiation weighting factor. The radiation weighting factors to be used for determining equivalent dose in rem are as follows: DOE-HDBK-1122-2009 Module 2.04 Dosimetry Study Guide 2.04-5 Table 1. RADIATION WEIGHTING FACTORS, wR Type and energy range Radiation weighting factor Photons, electrons and muons, all energies 1 Neutrons, energy < 10 keV2, 3 5 Neutrons, energy 10 keV to 100 keV2, 3 10 Neutrons, energy > 100 keV to 2 MeV2, 3 20 Neutrons, energy > 2 MeV to 20 MeV2, 3 10 Neutrons, energy > 20 MeV2, 3 5 Protons, other than recoil protons, energy > 2 MeV 5 Alpha particles, fission fragments, heavy nuclei 20 1. All values relate to the radiation incident on the body or, for internal sources, emitted from the source. 2. When spectral data are insufficient to identify the energy of the neutrons, a radiation weighting factor of 20 shall be used. 3. When spectral data are sufficient to identify the energy of the neutrons, the following equation may be used to determine a neutron radiation weighting factor value: wR = 5 + 17 exp ⎥ ⎦ ⎤ ⎢ ⎣ ⎡− 6 ))2(ln( 2 nE Where En is the neutron energy in MeV. Committed Effective Dose (E50): The sum of the committed equivalent doses to various tissues or organs in the body (HT,50), each multiplied by the appropriate tissue weighting factor (wT)--that is, E50 = ΣwTHT,50 + wRemainderHRemainder,50. Where wRemainder is the tissue weighting factor assigned to the remainder organs and tissues and HRemainder,50 is the committed DOE-HDBK-1122-2009 Module 2.04 Dosimetry Study Guide 2.04-6 equivalent dose to the remainder organs and tissues. Committed effective dose is expressed in units of rem (or Sv). Total Effective Dose (TED): The sum of the effective dose (for external exposures) and the Committed Effective Dose (for internal exposures). Annual Limit on Intake (ALI):

The derived limit for the amount of radioactive material taken into the body of an adult worker by inhalation or ingestion in a year. ALI is the smaller value of intake of a given radionuclide in a year by the reference man (ICRP Publication 23) that would result in a committed effective dose of 5 rems (0.05 sieverts (Sv)) (1 rem = 0.01 Sv) or a committed equivalent dose of 50 rems (0.5 Sv) to any individual organ or tissue. ALI values for intake by ingestion and inhalation of selected radionuclides are based on International Commission on Radiological Protection Publication 68, Dose Coefficients for Intakes of Radionuclides by Workers, published July, 1994 (ISBN 0 08 042651 4). Derived Air Concentration (DAC): For the radionuclides listed in appendix A of this part, the airborne concentration that equals the ALI divided by the volume of air breathed by an average worker for a working year of 2000 hours (assuming a breathing volume of 2400 m3). For the radionuclides listed in appendix C of this part, the air immersion DACs were calculated for a continuous, non-shielded exposure via immersion in a semi-infinite cloud of radioactive material. Except as noted in the footnotes to appendix A of this part, the values are based on dose coefficients from International Commission on Radiological Protection Publication 68, Dose Coefficients for Intakes of Radionuclides by Workers, published July, 1994 (ISBN 0 08 042651 4) and the associated ICRP computer program, The ICRP Database of Dose Coefficients: Workers and Members of the Public, (ISBN 0 08 043 8768). Bioassay: The determination of kinds, quantities, or concentrations, and, in some cases, locations of radioactive material in the human body, whether by direct measurement or by analysis, and evaluation of radioactive materials excreted or removed from the human body. In Vivo: A direct bioassay measurement of radioactivity in living tissue, for example, a whole body count or chest count. DOE-HDBK-1122-2009 Module 2.04 Dosimetry Study Guide 2.04-7 In Vitro: The bioassay measurement of radioactivity by means of internal representative sampling in order to estimate the radioactivity in tissue. Examples are analysis of urine and fecal collections. Background: Radiation from: naturally occurring radioactive materials which have not been technologically enhanced, cosmic sources, global fallout as it exists in the environment (such as from the testing of nuclear explosive devices), radon and its progeny in concentrations or levels existing in buildings or the environment which have not been elevated as a result of current or prior activities, and consumer products containing nominal amounts of radioactive material or producing nominal amounts of radiation. Declared Pregnant Worker: A woman who has voluntarily declared to her employer, in writing, her pregnancy for the purpose of being subject to the occupational exposure limits to the embryo/fetus as provided in 10 CFR 835.206. This declaration may be revoked, in writing, at any time by the declared pregnant worker. DOE LIMITS Limits are the legal maximum values stated in 10 CFR 835. To exceed these values is to violate the law. Programs must be in place to ensure that exposures to ionizing radiation are kept below these levels. To accomplish this, Administrative Control Levels are selected well below the regulatory limits. These control levels are usually multi-tiered with increasing levels of authority required to approve higher Administrative Control Levels.

Annual dose limits are based on a calendar year (January 1st through December 31st). For assigning internal doses received from intakes (committed equivalent dose and committed effective dose), the total 50-year committed dose received is assigned to the time of the intake even though the actual dose is proportionally received over the 50-year period. 2.04.01 Identify the DOE external exposure limits for occupational workers. DOE-HDBK-1122-2009 Module 2.04 Dosimetry Study Guide 2.04-8 General Employees General employees are DOE employees or DOE contractors. A Radiological Worker is a general employee whose job assignment involves operation of radiation producing devices or working with radioactive materials, or who is likely to be routinely occupationally exposed above 0.1 rem (0.001 sievert) per year total effective dose. Radiological workers from other DOE or DOE contractor facilities may receive occupational exposure to ionizing radiation as a radiological worker if they: • Provide a record of current Radiological Worker I or II standardized core training, • Receive site-specific Radiological Worker I or II training at the facility where they will be working, and • Provide their radiation dose record or a written estimate for the current year. Table 2 lists the various legal limits for exposure to ionizing radiation. There are four general categories listed: whole body, lens of the eyes, extremities and organ/tissue/skin. These limits are also covered in 10 CFR 835.208 and the Radiological Control Standard (RCS). Exposures should be well below the limits in this table and maintained as low as reasonably achievable. DOE-HDBK-1122-2009 Module 2.04 Dosimetry Study Guide 2.04-9 Table 2 - Summary of Dose Limits TYPE OF EXPOSURE ANNUAL LIMIT General Employees: Whole body (internal + external) Lens of Eye Extremity (hands and arms below the elbow: feet and legs below the knee) Any organ or tissue (other than lens of eye) and skin 5 rem (0.05 sievert) 15 rem (0.15 sievert) 50 rem (0.5 sievert) 50 rem (0.5 sievert) Declared Pregnant Worker Embryo/Fetus 0.5 rem (0.005 sievert) Per gestation period Minors (under age 18) and Students Whole body (internal + external) Extremity/Skin Lens of Eye 0.1 rem (0.001 sievert) 5 rem (0.05 sievert) 1.5 rem (0.001 sievert) Members of the public: Whole body (internal + external) 0.1 rem (0.001 sievert) Notes: 1. Internal dose to the whole body should be calculated as committed effective dose. The committed effective dose is the resulting dose committed to the whole body from internally deposited radionuclides over a 50-year period after intake. 2. The annual limit of exposure to “any organ or tissue” is based on the committed does to that organ or tissue resulting from internally deposited radionuclides over a 50-year period after intake plus any external effective dose to that organ during the year. 3. Exposures due to background radiation, therapeutic and diagnostic medical procedures, and participation in medical research programs should not to be included in either personnel radiation dose records or assessment of dose against the limits in this table. DOE-HDBK-1122-2009

Module 2.04 Dosimetry Study Guide 2.04-10 Minors//Public Minors are individuals less than 18 years of age. The public are defined as individuals not occupationally exposed to radiation or radioactive materials. An individual is not a "member of the public" during any period in which the individual receives an occupational dose. Occupational dose is an individual's dose due to exposure to ionizing radiation (external and internal) as a result of that individual's work assignment. Occupational dose does not include exposure received as a medical patient, background radiation, or participation in medical research programs. The DOE limit for exposure to minors and the public is stated in 10 CFR 835.207 and 835.208 and are listed in Table 2. Embryo/Fetus of Declared Pregnant Workers After a female general employee voluntarily notifies her supervisor in writing that she is pregnant, for the purposes of embryo/fetal dose protection, she should be considered a declared pregnant worker. The employer should provide the option of a mutually agreeable reassignment of work tasks, without loss of pay or promotional opportunity, such that further occupational radiation exposure is unlikely. For a declared pregnant worker who chooses to continue radiological work: • The dose limit for the embryo/fetus for the entire gestation period (from conception to birth) is 0.5 rem (0.005 sievert) {10 CFR 835.206}. • Efforts should be made to avoid exceeding 0.05 rem (0.0005 sievert) per month to the pregnant worker {10 CFR 835.206}. If the dose is likely to approach 0.05 rem/month (0.0005 sievert/month), additional dosimetry will be assigned to monitor the dose to the embryo/fetus. If the dose to the embryo/fetus is determined to have already exceeded 0.5 rem (0.005 sievert) when a worker notifies her employer of her pregnancy, the worker should not be assigned to tasks where additional occupational radiation exposure is likely during the remainder of the gestation period. 2.04.02 Identify the DOE limits established for the embryo/fetus of a female occupational worker. DOE-HDBK-1122-2009 Module 2.04 Dosimetry Study Guide 2.04-11 Emergency Exposures Emergency Exposure Situations For emergency situations, general employees could be allowed to exceed specified dose limits. The level of exposure permitted will depend upon the severity of the emergency situation. Exposures up to 2 times the annual dose limits could be permitted to protect against property loss. Higher exposures, up to 5 times the annual dose limits or greater, could be permitted to save lives and protect public health. The potential amount of exposure to rescue personnel should be evaluated, and an exposure objective should be established for the rescue mission. The DOE requires that the details of any exposure in excess of the annual dose limits be documented in the occupational exposure record of the affected employee. In addition, the incident must be investigated and the results reported to DOE. Departmental requirements for occurrence reporting and processing provide a mechanism for such investigations and reports. The employee must not be allowed to receive further exposure until approval is first obtained from the contractor management and responsible DOE field organization. Also, the employee must receive counseling from the appropriate health experts regarding the consequences of receiving additional occupational exposure that year and the affected employee must agree, before returning to radiological work.

Operations which have been suspended as a result of a dose in excess of the limits specified in § 835.202, except those received in accordance with § 835.204, may only be resumed with the approval of the DOE. The operation that caused the exposure must cease pending a finding by DOE that the conditions that caused the exposure had been eliminated. Planned Special Exposures A planned special exposure may be authorized for a radiological worker to receive doses in addition to and accounted for separately from the doses received under the normal occupational limits specified in Sec. 835.202(a) provided that each of the following conditions are satisfied: 1. The planned special exposure is considered only in an exceptional situation when alternatives that might prevent a radiological worker from exceeding the limits in 835.202(a) are unavailable or impractical; 2. The contractor management (and employer, if the employer is not the contractor) specifically requests the planned special exposure, in writing; and DOE-HDBK-1122-2009 Module 2.04 Dosimetry Study Guide 2.04-12 3. Joint written approval is received from the appropriate DOE Headquarters program office and the Secretarial Officer responsible for environment, safety, and health matters. Prior to requesting an individual to participate in an authorized planned special exposure, the individual's dose from all previous planned special exposures and all doses in excess of the occupational dose limits should be determined. An individual should not receive a planned special exposure that, in addition to these doses determined, would result in a dose exceeding: 1. In a year, the numerical value of the dose limits established in 835.202(a); or 2. Over the individual's lifetime, five times the numerical value of the dose limits provided in 835.202(a). Prior to a planned special exposure, written consent should be obtained from each individual involved. Each individual consent should include: 1. The purpose of the planned operations and procedures to be used; 2. The estimated doses and associated potential risks and specific radiological conditions and other hazards which might be involved in performing the task: and 3. Instructions on the measures to be taken to keep the dose ALARA considering other risks that may be present. Records of the conduct of a planned special exposure should be maintained and a written report submitted within 30 days after the planned special exposure to the approving organizations. The dose from these planned special exposures is not to be considered in controlling future occupational dose as part of the normal occupational dose of the individual under 835.202(a). Equivalency of Dosimetric Terms Under certain circumstances, when an individual conducts multiple activities involving both activities under 10 CFR 835.1(b)(1) and excluded activities, e.g., activities involving NRC licensed activities, it is not clear as to how to apply using different dose coefficients and weighting factors to calculate the overall cumulative total effective dose for workers. Accordingly DOE has stated that, for the purpose of compliance with 10 CFR 835.1(b)(1), DOE considers the following terms to be equivalent: DOE-HDBK-1122-2009 Module 2.04 Dosimetry Study Guide

2.04-13 Table 3 Equivalency of Dosimetric Terms Dosimetric Term Prior to 2007 Amendment to 10 CFR 835 DOE Amended Dosimetric Term Committed effective dose equivalent Committed effective dose Committed dose equivalent Committed equivalent dose Cumulative total effective dose equivalent Cumulative total effective dose Deep dose equivalent Equivalent dose to the whole body Dose equivalent Equivalent dose Effective dose equivalent Effective dose Lens of the eye dose equivalent Equivalent dose to the lens of the eye Quality factor Radiation weighting factor Shallow dose equivalent Equivalent dose to the skin or Equivalent dose to any extremity Weighting factor Tissue weighting factor Total effective dose equivalent Total effective dose SITE ADMINISTRATIVE GUIDELINES 2.04.03 Identify the administrative exposure control guidelines at your site, including those for the: a. Radiation worker b. Non-radiation worker c. Incidents and emergencies d. Embryo/Fetus DOE-HDBK-1122-2009 Module 2.04 Dosimetry Study Guide 2.04-14 Radiological Workers (Insert site specific information here) Non-radiation Worker (Insert site specific information here) Exposure from Incidents or Emergencies (Insert site specific information here) Embryo/Fetus of a Declared Pregnant Worker (Insert site specific information here) SITE EXPOSURE REQUIREMENTS FOR THE UNBORN CHILD (Insert site specific information here) TYPES OF DOSIMETRY As a result of irradiation, some solid substances undergo changes in some of their physical properties. These changes amount to storage of the energy from the radiation. Since the energy is stored, these materials can be used for dosimeters. The features that have been studied include: Optical density changes Optical density changes involve a change in the color of some types of plastics and glass. In glass, the dose range is 103 to 106 rads (10 to 104 gray). The range for plastics is 106 to 109 rads (104 to 107 gray). Film badges, a type of optical density dosimetry, provides low range monitoring 10 mR to 10 R for personnel and high range monitoring 1 R to 1,000 R for accident readings. 2.04.04 Identify the requirements for a female radiation worker who has notified her employer in writing that she is pregnant. DOE-HDBK-1122-2009 Module 2.04 Dosimetry Study Guide 2.04-15 Thermoluminescence Thermoluminescence (TL) is the ability of some materials to convert the energy from radiation to a radiation of a different wavelength, normally in the visible light range. There are two categories of thermoluminescence. Fluorescence This is emission of light during or immediately after irradiation (within fractions of a second) of the phosphor. This is not a particularly useful reaction for TLD use. Phosphorescence This is the emission of light after the irradiation period. The delay time can be from a few seconds to weeks or months. This is the principle of operation used for thermoluminescent dosimeters. The property of thermoluminescence of some materials is the main method used for personnel dosimeters at DOE facilities. TLD OPERATION TLDs use phosphorescence as their means of detection of radiation.

Electrons in some solids can exist in two energy states, called the valence band and the conduction band. The difference between the two bands is called the band gap. Electrons in the conduction band or in the band gap have more energy than the valence band electrons. Normally in a solid, no electrons exist in energy states contained in the band gap. This is a "forbidden region." In some materials, defects in the material exist or impurities are added that can trap electrons in the band gap and hold them there. These trapped electrons represent stored energy for the time that the electrons are held. (See figure 1) This energy is given up if the electron returns to the valence band. 2.04.05 Determine the theory of operation of a thermoluminescent dosimeter (TLD). DOE-HDBK-1122-2009 Module 2.04 Dosimetry Study Guide 2.04-16 In most materials, this energy is given up as heat in the surrounding material, however, in some materials a portion of energy is emitted as light photons. This property is called luminescence. (See figure 2) TLD READER Heating of the TL material causes the trapped electrons to return to the valence band. When this happens, energy is emitted in the form of visible light. The light output is detected and measured by a photomultiplier tube and a dose is then calculated. A typical basic TLD reader contains the following components: (See figure 3) • Heater - raises the phosphor temperature 2.04.06 Determine how a TLD reader measures the radiation dose from a TLD. DOE-HDBK-1122-2009 Module 2.04 Dosimetry Study Guide 2.04-17 • Photomultiplier Tube - measures the light output • Meter/Recorder - display and record data A glow curve can be obtained from the heating process. The light output from TL material is not easily interpreted. Multiple peaks result as the material is heated and electrons trapped in "shallow" traps are released. This results in a peak as these traps are emptied. The light output drops off as these traps are depleted. As heating continues, the electrons in deeper traps are released. This results in additional peaks. Usually the highest peak is used to calculate the dose. The area under the curve represents the radiation energy deposited on the TLD. A simple glow curve is shown in figure 4. DOE-HDBK-1122-2009 Module 2.04 Dosimetry Study Guide 2.04-18 After the readout is complete, the TLD is annealed at a high temperature. This process essentially zeroes the TL material by releasing all trapped electrons. The TLD is then ready for reuse. ADVANTAGES AND DISADVANTAGES OF TLDs Advantages (as compared to film dosimeter badges) includes: • Able to measure a greater range of doses • Doses may be easily obtained • They can be read on site instead of being sent away for developing • Quicker turnaround time for readout • Reusable 2.04.07 Identify the advantages and disadvantages of a TLD compared to a film badge. DOE-HDBK-1122-2009 Module 2.04 Dosimetry Study Guide 2.04-19 Disadvantages • Each dose cannot be read out more than once • The readout process effectively "zeroes" the TLD

SITE BETA/GAMMA TLDs (Insert site specific material here) SITE NEUTRON TLDs (Insert site specific material here) DOE EXTERNAL DOSIMETRY REQUIREMENTS Personnel dosimetry should be provided to and used by individuals as follows: 1. General employees who are expected to receive an effective dose to any portion of the whole body of 0.1 rem (0.001 sievert) or more in a year or an equivalent dose to the extremities, or organs and other tissues (including lens of the eye and skin) of 10 percent or more of the corresponding limits [835.402(a)(1)]; 2. Declared pregnant workers who are expected to receive from external sources an equivalent dose of 0.05 rem (0.0005 sievert) or more to the embryo/fetus during the gestation period [835.402(a)(2)]; 3. Occupationally exposed minors likely to receive from external sources an effective dose in excess of 50% of the limits [835.402(a)(3)]; 4. Members of the public who enter the controlled area and are likely to receive an effective dose of 0.05 rem (0.0005 sievert) or more in a year [835.402(a)(4)]; and 5. Individuals entering a high or very high radiation area radiation area [835.402(a)(5)]. 2.04.08 Identify the types of beta-gamma TLDs used at your site. 2.04.09 Identify the types of neutron TLDs used at your site. DOE-HDBK-1122-2009 Module 2.04 Dosimetry Study Guide 2.04-20 Neutron dosimetry shall be provided when an individual is likely to exceed the applicable threshold provided above due to neutron radiation [835.402(b)]. Dosimeters should be issued only to individuals knowledgeable of their proper use and worn only by those to whom the dosimeters were issued. To minimize the number of individuals in the dosimetry program, the issuance of dosimeters is discouraged to other than individuals entering radiological areas where there is a likelihood of external exposure in excess of the monitoring thresholds established in Article 511.1 of the Radiological Control Standard. Although issuing dosimeters to individuals who are not occupationally exposed to radiation can appear as a conservative practice, it creates the impression that the wearers are occupationally exposed to radiation. Implementation of an unnecessarily broad dosimetry program is not an acceptable substitute for development of a comprehensive workplace monitoring program. Individuals should return dosimeters for processing as scheduled or upon request, and should be restricted by line management from continued radiological work until dosimeters are returned. Individuals should wear their primary dosimeters on the chest area, on or between the waist and the neck, or in the manner prescribed by radiological control procedures or work authorizations. Film dosimeters should not be worn or taken off-site unless specifically authorized by the Radiological Control Manager or designee. The practice at some facilities of taking thermoluminescent dosimeters (TLDs) off-site is discouraged and should not be implemented where not in place. Individuals should not wear dosimeters issued by their resident facilities while being monitored by a dosimeter at another facility unless authorized by the Radiological Control Manager or designee. Individuals should not expose their dosimeters to security X-ray devices, excessive heat, or medical sources of radiation.

An individual whose dosimeter is lost, damaged, or contaminated should place work in a safe condition, immediately exit the area, and report the occurrence to the Radiological Control Organization. Reentry of the individual into radiological areas should not be made until a review has been conducted and management has approved reentry. 2.04.10 Determine the requirements for use of TLDs used at your site. DOE-HDBK-1122-2009 Module 2.04 Dosimetry Study Guide 2.04-21 SITE REQUIREMENTS FOR USE OF TLDs (Insert site specific material here) SITE PERSONNEL NEUTRON DOSIMETERS (Insert site specific material here) POCKET AND ELECTRONIC DOSIMETERS Pocket and electronic dosimeters are supplemental dosimeters that provide real-time indication of exposure to radiation and assist in maintaining personnel doses less than Administrative Control Levels. Supplemental dosimeters shall be issued to personnel prior to entry into a High or Very High Radiation Area [835.502(a)(2)]. Supplemental dosimeters should also be issued when planned activities could cause an individual to exceed 50 millirem or 10 percent of a facility Administrative Control Level from external radiation in 1 work day, whichever is greater or when required by a Radiological Work Permit. Pocket dosimeters should be selected with the lowest range applicable (typically 0-200 mR) for anticipated personnel exposures. Supplemental dosimeters should be worn simultaneously with the primary dosimeter and located on the chest area, on or between the waist and the neck. Supplemental dosimeters should be read periodically while in use and should not be allowed to exceed 75 percent of full scale. Work authorized by written authorization should be stopped when supplemental dosimeter readings indicate total exposure or rate of exposure substantially greater than planned. The Radiological Control Organization should be consulted prior to continuation of work. The energy dependence of supplemental dosimeters, particularly to low-energy beta radiation, should be considered in determining their applicability. For example, the SRPD (shown in figure 5) has a thick case that effectively shields most betas. Use of electronic dosimeters is encouraged for entry into High Radiation Areas or when planned doses greater than 0.1 rem (0.001 sievert) in 1 work day are expected. 2.04.11 Determine the principle of operation, and the types used, for the personnel neutron dosimeters used at your site. DOE-HDBK-1122-2009 Module 2.04 Dosimetry Study Guide 2.04-22 An electronic dosimeter provides an early warning of elevated exposure through the use of alarm set points at specified dose rates or integrated doses. When the dose results from the pocket or electronic dosimeters differ by more than 50 percent from the primary dosimeter result and the primary dosimeter result is greater than 0.1 rem (0.001 sievert), an investigation should be initiated to explain the difference. SITE SELF-READING DOSIMETERS (Insert site specific material here) Self Reading Pocket Dosimeters (SRPD) The direct reading pocket dosimeter consists of an ionization chamber sensitive to a desired radiation; a quartz fiber electrometer to measure the charge; and a microscope to read the fiber image off a scale (reticule). (See figure 5)

The electrometer embodies two electrodes, one of which is a moveable quartz fiber and the other a metal frame. When the electrometer is charged to a predetermined voltage, the electrodes assume a calibrated separation. 2.04.12 Determine the principle of operation of self-reading dosimetry (SRD) used at your site. DOE-HDBK-1122-2009 Module 2.04 Dosimetry Study Guide 2.04-23 As the dosimeter is exposed to radiation, ionization occurs in the surrounding chamber decreasing the charge on the electrode in proportion to the exposure. The deflection of the moveable quartz fiber electrode is projected by a light source through an objective lens to a calibrated scale and read through a microscope eyepiece. (See Figure 6) DOE-HDBK-1122-2009 Module 2.04 Dosimetry Study Guide 2.04-24 Illumination for the optical system is obtained by pointing the dosimeter at any convenient light source. Light passes through the clear glass bottom seal to illuminate the scale. The bottom is sealed by a bellows containing an insulated charging pin. When charging, the charging pin moves up to contact the electrometer closing the circuit. Sufficient voltage is applied to recharge the system. The entire dosimeter system is hermetically sealed in a protective barrel. DOE-HDBK-1122-2009 Module 2.04 Dosimetry Study Guide 2.04-25 SITE ALARMING DOSIMETRY (Insert site specific material here) INTERNAL DOSIMETRY REQUIREMENTS Per 10 CFR 835: for the purpose of monitoring individual exposures to internal radiation, internal dose evaluation programs (including routine bioassay programs) shall be conducted for: 1. General employees who, under typical conditions, are likely to receive 0.1 rem (0.001 sievert) or more committed effective dose from all occupational radionuclide intakes in a year; 2. Declared pregnant workers likely to receive an intake resulting in an equivalent dose to the embryo/fetus in excess of 10 percent of the limit (or 0.05 rem [0.0005 sievert]); 3. Occupationally exposed minors who are likely to receive a committed effective dose in excess of 50 percent of the applicable limit (or 0.05 rem [0.0005 sievert]) from all radionuclide intakes in a year; 4. Members of the public entering a controlled area likely to receive a committed effective dose in excess of 50 percent of the limit (or 0.05 rem [0.0005 sievert]) from all radionuclide intakes in a year. The estimation of internal dose should be based on bioassay data rather than air concentration values unless bioassay data are unavailable, inadequate, or internal dose estimates based on air concentration values are demonstrated to be as or more accurate. Personnel should participate in follow-up bioassay monitoring when their routine bioassay results indicate an intake in the current year with a committed effective dose of 0.1 rem (0.001 sievert) or more. Personnel whose routine duties may involve exposure to surface or airborne contamination or to radionuclides readily absorbed through the skin, such as tritium, should be considered for participation in the bioassay program. 2.04.13 Determine the principle of operation, and guidelines for use, for the alarming dosimeters used at your site.

DOE-HDBK-1122-2009 Module 2.04 Dosimetry Study Guide 2.04-26 Personnel should submit bioassay samples, such as urine or fecal samples, and participate in bioassay monitoring, such as whole body or lung counting, at the frequency required by the bioassay program. Personnel should be notified promptly of positive bioassay results and the results of dose assessments and subsequent refinements. Dose assessment results should be provided in terms of rem or mrem. BIOASSAY ASSESSMENT METHODS Today's technology has not produced a device that allows accurate determination of internal exposure following the entry of radioactive materials into the body. The method that is used to determine internal dose contributions relies on calculation of dose to affected portions of the body based on the quantities of radioactive materials in the body. Thus, the real problem becomes one of quantifying the amount of material present. Bioassay is the term that is used to describe the assessment of the quantity of radioactive material present in the body. There are currently two types of bioassay measurements employed in nuclear industries: in vivo and in vitro. In vivo bioassay involves counting the living tissue, as described below. In vitro involves counting an excreted sample, such as urine. Bioassay programs are designed to fulfill two needs: 1) Evaluate effectiveness of contamination control practices • Routine bioassay programs utilize submission and analysis of samples from workers in facilities where the likelihood of intake exists • Primarily limited to urinalysis due to ease of sample collection • Also includes initial, routine, and termination whole body counts 2) Evaluate potential consequences of accidental inhalation or ingestion of large quantities of radioactive materials • Can involve all types of bioassay measurements with collection and analysis of nasal, urine, and fecal samples. • Whole body counts provide immediate indications for given radionuclides if individual(s) involved are free of external contamination. DOE-HDBK-1122-2009 Module 2.04 Dosimetry Study Guide 2.04-27 Quantification of materials actually in the body can be affected by the availability of measurements taken early after the incident. The elimination rate of some materials from the body falls off as the concentration in the body falls off, or with time. Accurate quantification of initial quantities, present, thus accurate dose assessment, can be dependent on availability of early data. Identification of the proper bioassay technique to use is aided by a knowledge of the types of contamination present in a particular work area. For example, if you know that the contamination in a facility typically includes radionuclides that cannot be detected with in vivo measurements, then it would be obvious that collection and measurement of urine or other samples is necessary. If the presence of gamma emitting nuclides is identified, consider the possibility of the presence of materials that do not decay with gamma emission. Periodic radionuclide assessment of contamination in facilities will provide information on relative radionuclide concentrations. Caution must be exercised in using information of this nature. Cycles of contamination should be used as an indicator only. Remember, fresh coolant does not have the same isotopic makeup of coolant that has decayed.

Contamination control measures cannot be too stringent during collection, handling, and analysis of bioassay samples. Cross-contamination can cause erroneous assumptions and inaccurate dose assessments. If procedural guidance is not sufficient to determine required actions, consult supervision. In Vivo Measurements In vivo techniques consist of direct measurements of gamma or X-radiation emanating from the body. This method is very useful for any radionuclide which emits (or has daughters which emit) photons of sufficient energy to escape the body. The photon flux density must be large enough for measurement in a reasonable time period, even though the quantity of material in the organ is very small. This method is possible only for those radionuclides emitting penetrating radiation, e.g., Co-60 and Cs-137 or bremsstrahlung, e.g., P-32 and Sr-90. Many radionuclides, Na-22, Fe-59, Co-60, Zn-65, Rb-86, Sr-85, Te-132, I-131, Cs-137, Ba-140, Ce-144, Au-198, U-235, Np-239, and Am-241 emit electromagnetic radiation of sufficient energy to be measured by external counting. If the counter has been calibrated previously, one may rapidly determine the identity and amount of any of these radionuclides. Such measurements are more acceptable to the subject than the provision of samples of excreta, although they do require him to be absent from work during the period of measurement. Direct counting of the individual without preparation beforehand (changing into clean clothes and external decontamination) may give misleading results, since this method measures all gamma emitting radionuclides in or on a subject; therefore, sensitive counts (lung) should be done immediately after the subject washes and changes into clean clothing. Radon DOE-HDBK-1122-2009 Module 2.04 Dosimetry Study Guide 2.04-28 daughters that cling to body hair due to their electrostatic charge are the chief source of bad lung counts. When this method errs, it usually does so by being too high, so that a negative result is likely to be a reliable indication that there is no internal contamination with gamma emitters. In external counting, the requirement for sensitivity and energy discrimination determines the complexity of the measuring equipment. Estimations of very small quantities require elaborate shielding of both the sensing element and the subject, sensitive detectors, and the best discrimination between gamma ray energies. However, a single moderately large, well-shielded sodium iodide crystal coupled with a multichannel analyzer can usually meet the need. This system in conjunction with a shielded chair or moving bed, is capable of determining: • I-131 in the thyroid gland. • Insoluble radionuclides in the chest. • Insoluble radionuclides in the intestine. • Insoluble radionuclides in wounds. These need not emit highly penetrating radiation, since much of the material may be on or near the surface, i.e., for wounds. Because large sodium iodide crystals do not have good collimation capabilities, it is usually not possible to measure specific organ contents directly. In some cases, solid state detector (GeLi) can be used for specific organ determination. However, the decreased sensitivity of this method limits the usefulness of these measurements. Small sodium iodide detectors are used for determining thyroid dose. Site In Vivo Methods

(Insert site specific material here) Advantages of In Vivo Measurements • No sample required • Results obtained quickly • Some equipment design allows field use • Time and manpower requirements minimized. Disadvantages of In Vivo Measurements 2.04.14 List the types of bioassay monitoring methods at your site. DOE-HDBK-1122-2009 Module 2.04 Dosimetry Study Guide 2.04-29 • Limited to detection and measurement of gamma emitters • Individual must be free of external contamination • Long count times for identification • Effects of background • Complex calibration procedure and calibration equipment • Expense • Quantification error due to differences in tissue structure from one person to another as compared to calibration phantom. In Vitro Measurements The amount of material present in the body is estimated using the amount of materials present in excretions or secretions from the body. Samples could include urine, feces, blood, sputum, saliva, hair, and nasal discharges. Calculation requires knowledge and use of metabolic models which allow sample activity to be related to activity present in the body. Resulting dose calculations to quantify committed and effective doses are estimates. This is due partly to use of default values for measurements that cannot be readily made such as mass of particular organs, volumes of particular fluids, etc., in lieu of actual values for individual involved. Remember that reference man is an average. Another contributing factor is the difference in metabolism from one individual to another. Urinalysis Indicates effectively that soluble radioactive material has been deposited in the blood for transport to various organs. A fraction of the material is normally removed from the blood by the kidneys and excreted. Later, material absorbed by various organs may be released to the blood through biological exchange processes, and then may be excreted in the urine. Certain compounds are determined to be insoluble because they are avidly retained in the lung. However, they also eventually appear in the urine. Particles are removed to the pharynx by the ciliary-mucus transport mechanism where they are swallowed, dissipated and partially absorbed in the gastrointestinal tract for transport to the blood. Other particles are removed by transport to the lymphatic system for subsequent release to the blood. Other particles slowly enter into a physical or chemical state which allows direct transport from the pulmonary region of the lung to the blood. All three cases lead to urinary excretion of the material. DOE-HDBK-1122-2009 Module 2.04 Dosimetry Study Guide 2.04-30 Taking samples of urine involves two special difficulties. One is the possibility of contamination if the sample is taken at work. The other is the problem of collecting a sample from which can be calculated the total excretion of radionuclide per unit time, usually per day. It is ordinarily not convenient to collect a full 24-hr sample of urine, so it is frequently necessary to estimate the fraction this is of the relatively constant daily urine excretion.

One of the advantages of measuring the radionuclide content of urine is that if a radionuclide is found in a carefully collected sample of urine, there can be no doubt that it was in extracellular body fluids. Furthermore, under the most favorable conditions, the amount of daily urinary excretion of radionuclide may be used directly to calculate total body content. One of the simplest examples of practical importance is tritium oxide which is present in the same concentration in urine as in extracellular fluids of the body. Almost all employees are willing to provide a limited number of urine samples; however, prolonged urine sampling involving samples taken both at home and at work will often meet with increasing employee resistance. Fecal Analysis An appreciable fraction of the particles entering the gastrointestinal tract may not be absorbed; these appear in the feces within twenty-four hours. Thus, fecal analysis is an excellent and relatively rapid indicator that an exposure has occurred. Fecal analysis is particularly useful for inhaled, insoluble materials that do not appear in the urine for weeks. For many highly insoluble materials, particles remaining in the pulmonary system continue to reach the mucus blanket, although at a greatly reduced rate. These particles are then transported by ciliary action to the gastrointestinal tract. Thus, fecal analysis can also contribute to the estimate of the lung burden. Two drawbacks to fecal analysis are: (1) there is considerable employee resistance to provide fecal samples and (2) there is very little correlation between fecal content and organ depositions. Thus, fecal analysis is primarily a qualitative method used only for detecting the intake of insoluble materials and providing indication of clearance of such materials from the lungs. Fecal sampling is normally done immediately following an incident because correlation is best when intake times are known. Sputum When obtainable, sputum may contain insoluble material initially deposited in the lung and later eliminated by ciliary action. However, clearance time for sputum is very rapid and samples must be taken immediately after an incident. DOE-HDBK-1122-2009 Module 2.04 Dosimetry Study Guide 2.04-31 Saliva May be analyzed to detect internal contamination, but the only practical case in which saliva can be used to estimate body content is that of tritium oxide, for which urine is the usual method. Nasal Discharge The presence of radionuclides in nasal discharge and nasal swabs generally gives an indication of the deposition of the coarsest inhaled particles in the nose. Measurement of the amounts present cannot always be used for quantitative estimation of the amount in the body, but it can be useful in detecting significant exposures and identifying the radionuclide involved in an accident. Site In Vitro Methods (Insert site specific material here) Advantages of In Vitro Measurements • Can be used for estimation of neutron doses using activation product concentration in hair and blood (P32 and Na24) • Can be used to quantify presence of materials which decay by alpha and beta emission to allow detection and measurement with external detector systems. Disadvantages of In Vitro Measurements • Requires sample submission and analysis • Time and manpower requirements BIOASSAY SCHEDULING PROGRAM

Contamination found in a given facility will depend on the materials that are used and produced in the facility. Thus, the materials that internal dosimetrists are primarily concerned with will change from one site to another as well. Baseline/Routine/Exit Evaluations (Insert site specific material here) 2.04.14 List the types of bioassay monitoring methods at your site. DOE-HDBK-1122-2009 Module 2.04 Dosimetry Study Guide 2.04-32 Special Evaluations (Insert site specific material here) Investigation Levels (Insert site specific material here) Medical Uses (Insert site specific material here) SUMMARY The method of operation of dosimeters is a vital knowledge for RCT. RC personnel are the first line of defense against abuse of these instruments and must ensure the proper wearing and use of them. Internal exposure involves a source (contaminant) inside the body. It is more difficult to measure; sophisticated whole body counters or indirect measurements of excreta samples are required to obtain an estimate. The exposure from the contaminant does not stop when the person leaves the radiation field and the contaminant continues to irradiate tissue all day and all night. If necessary, medical treatment is required to enhance the removal of the source material from the body. Alpha radiation poses the biggest problem. DOE-HDBK-1122-2009 Module 2.04 Dosimetry Study Guide 2.04-33 This page intentionally left blank. DOE-HDBK-1122-2009 Module 2.05 Contamination Control Study Guide 2.05-1 Course Title: Radiological Control Technician Module Title: Contamination Control Module Number: 2.05 Objectives: 2.05.01 Define the terms "removable and fixed surface contamination," state the difference between them and list common methods used to measure each. 2.05.02 State the components of a radiological monitoring program for contamination control and common methods used to accomplish them. 2.05.03 State the basic goal of a contamination control program and list actions that contribute to its success. 2.05.04 State the basic principles of contamination control and list examples of implementation methods. 2.05.05 List and describe the possible engineering control methods used for contamination control. 2.05.06 State the purpose of using protective clothing in contamination areas. 2.05.07 List the basic factors which determine protective clothing requirements for personnel protection. INTRODUCTION Contamination control is probably one of the most difficult and challenging tasks the Radiological Control Technician will encounter. To have a successful contamination control program, the radiological control staff must have considerable foresight, initiative, and experience. References: 1. DOE Radiological Control Standard, Articles 325, 337 and 338 (2008). 2. "The Health Physics and Radiological Health Handbook," Shleien; 1992. DOE-HDBK-1122-2009 Module 2.05 Contamination Control Study Guide 2.05-2 TYPES OF CONTAMINATION

Contamination is simply defined as radioactive material in an unwanted location, e.g., personnel, work areas, etc. Two types of contamination are possible, fixed and removable. Fixed contamination is radioactive surface contamination that is not easily transferred to other personnel or equipment through normal contact. Removable contamination is radioactive surface contamination that is easily transferred to other personnel or equipment through normal contact. Removable contamination is measured by a transfer test using a suitable sampling material. Common materials used for the monitoring are the standard paper disk smear or cloth smear. The standard technique involves wiping approximately 100 cm2 of the surface of interest using moderate pressure. A common sampling practice used to ensure a 100 cm2 sample is to wipe a 16 square inch "S" shape on the surface (i.e., 4 inches by 4 inches). Qualitative, large area wipe surveys may be taken using other materials, such as Masslin cloth or Kimwipe, to indicate the presence of removable contamination. These are commonly used when exact levels of contamination are not required. Fixed contamination is measured by use of a direct survey technique. This technique, commonly referred to as "frisking," indicates the total contamination on a surface apparent to the detector from both fixed and removable. When non-removable levels are to be recorded, the removable level must be subtracted from the total. Appendix D to 10 CFR 835 lists contamination levels in units of dpm/100 cm2. Typical evaluation of fixed contamination monitoring includes adjusting the portable instrument count rate (counts per minute) to account for the area of the monitoring instrument (probe area) and the instrument efficiency to obtain units of dpm/100 cm2. Footnote 4 to Appendix D states that when “removable contamination on objects of surface area less than 100 cm2 is determined, the activity per unit area shall be based on the actual area and the entire surface shall be wiped”. For example, for an object with a total surface area of 50 cm2, the entire object would be wiped and the count result would be divided by the counting efficiency and multiplied by an area adjustment factor of two (100 cm2/50 cm2) to get a result in units of dpm/100 cm2. A similar approach may be used for fixed contamination monitoring of objects with surface area less than 100 cm2. However, care must be taken in monitoring very small objects and the sensitivity of the monitoring protocol should be evaluated. 2.05.01 Define the terms "removable and fixed surface contamination," state the difference between them and list common methods used to measure each. DOE-HDBK-1122-2009 Module 2.05 Contamination Control Study Guide 2.05-3 ASSESSING CONTAMINATION HAZARDS In order to acquire the radiological information necessary for contamination control, there are several components to a radiological monitoring program. These are: • Constant monitoring • Area and equipment surveys • External personnel surveys • Personnel internal monitoring and bioassay Constant Monitoring There are various types of constant monitoring instruments throughout the facilities to warn personnel of radiation and contamination hazards. Some instruments are permanently installed, and some instruments are portable to allow movement from place to place as deemed appropriate by the radiological control staff.

Continuous air monitor (CAM). These instruments continuously sample the air for radioactive contamination in specific locations. The air being sampled is either drawn through a moving particulate filter which is then monitored by a detector system or through an internal detector to directly identify radioactive materials present. A CAM can give both a visual and audible alarm to warn personnel of the presence of airborne contamination. Process monitoring systems. Process monitoring systems monitor certain operations in various facilities to alert operators of abnormal conditions which might lead to the release of excessive amounts of radioactivity to the facility or environment. Area and Equipment Surveys Area and equipment surveys are conducted routinely throughout the facilities to locate sources of radiation and contamination and to detect potential changes in radiological conditions. Pre-job surveys are performed prior to work in radiological areas in order to evaluate the hazards and determine work limitations and physical safeguards. Direct instrument surveys. Various types of portable survey instrumentation are used to measure the presence of radioactive contamination on a floor or surface. This is the only method available to detect "fixed" surface contamination. It must 2.05.02 State the components of a radiological monitoring program for contamination control and common methods used to accomplish them. DOE-HDBK-1122-2009 Module 2.05 Contamination Control Study Guide 2.05-4 be remembered, however, that this method will detect removable as well as "fixed" surface contamination activity. As a result, a direct survey must be combined with a "smear" survey to determine if the surface contamination present is removable or fixed. Smear surveys. A disk smear is wiped over an area of 100 square centimeters and counted with proper instrumentation to determine the activity of the nuclides present. Contamination levels are specified in units of dpm/100 cm2 after applying applicable instrument correction factors. For objects less than 100 cm2, the units are reported as dpm/object area. Disk smears are small so they are usually used in an area of suspected contamination. Properly applied experience will dictate to the surveyor where contamination is most likely to occur and hence those areas that should be surveyed with disk smears. Disk smears are required if contamination levels are to be quantified. Many routine contamination surveys are taken in areas not suspected to be contaminated with a chemically treated cloth called a masslinn (paper towel, atomic swipe, etc). The cloth is lightly pushed over an area and scanned with an appropriate detector to detect the presence of contamination. If contamination is detected, a more thorough disk smear survey should be performed. These large area wipes are used only as an indication of removable surface contamination. External Personnel Surveys Personnel surveys are either performed by the individual (self-monitoring) using hand-held or automated instruments or by a radiological control technician. Self- monitoring is typically performed upon exiting a contaminated area at established boundary points. Personnel monitoring by a RCT is usually conducted whenever contamination of the body or clothing is suspected, or as required by exit monitoring when self-monitoring is not feasible (remote location) or not allowed. The types of hand-held or automated instruments used for self-monitoring are generically described below.

Personnel monitors. Portable instruments (friskers) with sensitive hand held detectors are used by personnel to identify contamination on themselves whenever contamination is suspected. These monitors are used whenever exiting contaminated areas, Radiological Buffer Areas, and in some cases Radiological Control Areas (RCA). Geiger-Mueller (GM) detectors are most often used for beta-gamma monitoring and scintillation detectors for alpha monitoring. Personnel Contamination Monitors (PCM). The PCM provides personnel with an external whole body monitoring system. The contamination detectors within the monitors are capable of performing a survey of the whole body in a period of a few seconds, dependent upon background radiation levels present in the area and the personnel contamination limit of concern. These automated systems DOE-HDBK-1122-2009 Module 2.05 Contamination Control Study Guide 2.05-5 typically provide a more reliable method of locating personnel contamination over hand-held instruments. Hand and Foot Monitors. Hand and foot monitors with detachable hand-held detectors provide another alternative to using hand-held instruments (friskers). These devices can monitor the hands and feet during a period of a few seconds, again, dependent upon background radiation levels present in the area and the personnel contamination limit of concern. After the hands and feet have been monitored, the detachable hand held detectors, which are typically of a larger detector size, can be used to monitor the remainder of the body in a shorter time period than most friskers. Portal monitors. The portal monitor is a "door frame" type device which provides a final monitoring point to ensure contamination is not spread outside the facility to other facilities or the general public. These types of monitors are typically used only for beta-gamma monitoring. Personnel surveys. Personnel surveys are performed whenever contamination of the body or clothing is suspected, or as required for exit monitoring, e.g., when friskers or automated monitoring instruments are not available. The whole body should be surveyed with special attention to areas which are more likely to become contaminated. Contamination of the feet (shoes) would indicate removable surface contamination on the floor just traversed. The hands are extremely prone to becoming contaminated when working directly with radioactive materials. Upon completion of work or prior to leaving the area for glovebox, laboratory fume hood, sample station, or localized benchtop operations, a minimum survey of hands, arms, and front portions of the body must be performed. Other body areas which are prone to contamination are the buttocks, knees, and elbows and head. The nose and mouth should be surveyed upon discovery of any level of facial skin contamination, or if airborne contamination was detected in the workplace, since contamination in this area might indicate the need for bioassay sampling. The nose can be swabbed with Q-tips and the swab counted in a smear counter to determine a potential deposition. Contamination of the nose or mouth may indicate airborne contamination. All open wounds must be monitored since contaminants can be readily absorbed into the body. DOE-HDBK-1122-2009 Module 2.05 Contamination Control Study Guide 2.05-6

In addition to these specific body areas, the surveyor should pay special attention to any area of the body and/or clothing which he or she suspects might be contaminated. Upon detecting personnel contamination, follow-up area and/or equipment surveys may be necessary to determine the source of contamination and the extent the contamination has spread, if any. Personnel Internal Monitoring A routine program of internal contamination monitoring is conducted as a final check on contamination control procedures. This program consists of external whole/partial body counting and/or urinalysis. In-vivo Bioassay: The individual is placed inside a array of very sensitive detectors to measure the activity and energies of gamma ray emissions from inside the body. This information can be used to determine the amount and identify the type of nuclides present. Examples include whole body, lung, or scanning bed counters. In-vitro Bioassay: Urine or feces samples are collected from an individual to determine the type and activity of the nuclides present in bodily waste. This information is used to approximate the amount of nuclides present in the body by their calculated rate of elimination. This method can be used to assess the presence of non-gamma emitting nuclides. BASIC GOAL OF CONTAMINATION CONTROL Once the presence of radioactive material has been located, the basic goal underlying any effective contamination control program is to minimize contaminated areas and maintain contamination levels as low as reasonably achievable. In some situations, this is not always possible due to: • Economical conditions: Cost of time and labor to decontaminate a location(s) out-weighs the hazards of the contamination present. 2.05.03 State the basic goal of a contamination control program and list actions that contribute to its success. DOE-HDBK-1122-2009 Module 2.05 Contamination Control Study Guide 2.05-7 • Radiological conditions: Radiation dose rates or other radiological conditions present hazards which far exceed the benefits of decontamination. • Operating conditions: Some areas, e.g., hot cells, will be contaminated due to normal operations. Other means of control must be initiated when decontamination is not possible. Engineering control (ventilation and containment), administrative procedures (RWPs), and personnel protective equipment are alternatives for the control of contamination. In Fixed Contamination Areas the contamination may be covered by paint, floor tiles, etc. when decontamination is not possible. "Good Housekeeping" is a prime factor in an effective contamination control program. It involves the interactions of all groups within the facility. Each individual must be dedicated to keeping "his house clean" to control the spread of contamination. Every possible effort should be made in all operations to confine the spread of radioactive materials to the smallest possible area. A sound preventive and corrective maintenance program can prevent many radioactive material releases. All material taken into or out of contaminated areas must be controlled. RCTs should always be alert for potential violations to the basic principles of contamination control. • Use of improper contamination control methods • Bad work practices • Basic rule or procedure violations • Radioactive material releases or liquid spills CONTAMINATION CONTROL MEASURES

Controlling the spread of contamination is probably the most difficult and challenging task the Radiological Control Technician will encounter. To have a successful contamination control program, the radiological control staff must have considerable foresight, initiative, and experience. The radiological control staff will assist line management with the basic principles of contamination control. • Access/Administrative Controls • Engineering Controls • Personnel Protective Measures • Decontamination • Preventive Methods 2.05.04 State the basic principles of contamination control and list examples of implementation methods. DOE-HDBK-1122-2009 Module 2.05 Contamination Control Study Guide 2.05-8 Access/Administrative Controls Once contamination has been located and quantified and radiological areas have been determined, access control to these areas must be adequately established. Two basic access control points, primary and secondary, are used in contamination control. The primary access control point in a facility is the entry and exit portal between the clean area and the radiologically controlled area or Radiological Buffer Area. The success of a control program is based on controlling the movement of personnel and equipment between these areas to prevent release of contamination to a clean location. The secondary access control points (perhaps the most important) are set up within the Radiological Buffer Areas (RBAs) to control access between surface contaminated areas and non-contaminated areas. Yellow and magenta rope, chain, tape or similar barriers are used to identify the boundaries and provide a recognizable visual barrier to personnel. In areas of ongoing work activities, special requirements will always be established for entry and exit through these access control points. When the radiological conditions are severe, the access control point may be continuously manned by a Radiological Control Technician. It is not expected that Radiological Buffer Areas will be established around inactive or secured Contamination Areas. Step-off pads (SOPs) identify the entry and exit points to contaminated areas when possible. The use of SOPs creates a sharp line of distinction between the surface contamination area and the clean areas. Proper procedures must be established and observed for crossing the SOP to prevent the spread of contamination. All tools and/or equipment used in a surface contamination area which are unmonitored should be placed in clean plastic bags or securely wrapped in plastic before being removed from the area. All personnel and materials exiting the area should be monitored to ensure they are free of contamination. Radiological Buffer Areas should also be established in areas where there is a need to limit exposure to external radiation, such as Radiation, High Radiation, and Very High Radiation Areas. The boundary should be established to limit radiation dose to general employees to less than 100 mrem per year. RBAs need not be posted for external exposure control if other posted boundaries provide equivalent employee protection. Other administrative controls used for contamination control include the use of Radiological Work Permits, routine workplace surveys that are performed in order to detect trends in the potential buildup of workplace contamination, and DOE-HDBK-1122-2009

Module 2.05 Contamination Control Study Guide 2.05-9 review of operational and maintenance procedures to ensure radiological requirements are incorporated in the daily conduct of operations. ENGINEERING CONTROLS Ventilation. The design of permanent or temporary ventilation systems needs to be such that air flow is from clean areas to RCAs, to areas of moderate contamination, to areas of high contamination, and finally to an exhaust system capable of removing any contamination from the air. Slight negative pressure is typically maintained in buildings/rooms where potential contamination exists. As necessary, high efficiency particulate air (HEPA) filters are used to remove radioactive particles from the air. Containment. On jobs with very high contamination potential, a plastic tent (greenhouse or hut) can be built around the work area to confine all contamination to as small an area as possible. A portable ventilation exhaust system (such as HEPAs) may be used to control air flow in the work area and remove airborne contamination. Where possible, small containment devices, such as glove boxes, glove bags, or hoods can be used to contain the contamination depending on the nature and location of the work being performed. Drums or other approved containers are also utilized. Bagging. The most widely used method of containment is bagging or wrapping. Contaminated tools or equipment are placed in plastic bags, or securely wrapped in plastic, before being moved outside a contaminated area. When possible, wrapping tools or equipment prior to entry can help control contamination during use inside the contaminated area. Design and Control. Design of facilities should be such that efficiency of maintenance, operations, and decontamination is maximized. Components should be selected that minimize the buildup of radioactivity. Support facilities are to be included that provides for donning and doffing of protective clothing and for personnel monitoring. Personnel traffic should be routed away from contaminated areas. PERSONNEL PROTECTIVE MEASURES 2.05.05 List and describe the possible engineering control methods used for contamination control. 2.05.06 State the purpose of using protective clothing in contamination areas. DOE-HDBK-1122-2009 Module 2.05 Contamination Control Study Guide 2.05-10 If engineering control methods are not adequate, then personnel protective measures, such as protective clothing and respiratory equipment, will be used. The purpose of protective clothing is to keep contamination off the skin and clothing of the workers. Protective clothing allows personnel to work inside a contaminated area with removable contamination and to exit the area without spreading contamination to uncontrolled areas. The use of protective clothing alone will not guarantee complete elimination of personnel contamination and is not a substitute for implementing proper controls, but if used properly, protective clothing will afford a high degree of protection. All personnel entering contaminated areas with removable contamination will be required to wear certain items of protective clothing. The types of clothing required will vary depending upon the contamination levels and the nature of the work to be performed. Some additional factors for the selection of protective clothing include the type and form of contamination; potential for increased levels of contamination, area of the body at risk, and competing hazards, i.e., heat stress, asbestos, etc.

Some type of respiratory protective equipment will be required for work in areas where very high contamination levels exist or airborne contamination is present. Decontamination Line management is responsible for ensuring prompt decontamination, where practical, of facilities, tools, material, and equipment so that contamination can be minimized in the workplace. Reasonable efforts should be directed toward the decontamination and unconditional release of these items rather than their disposal as radioactive waste. Only items that are extremely contaminated with risks during decontamination that out-weighs the benefit to be gained for reuse should be considered for disposal. Preventive Methods The following are practical methods used for the prevention/control of contamination: • Identify and repair leaks before they become a serious problem. • Establish adequate work controls before starting jobs. • While conducting pre-job briefs, discuss measures that will help reduce or prevent contamination spread. • Change out gloves or protective gear as necessary to prevent cross- contamination of equipment. • Pre-stage areas to prevent contamination spread from work activities. DOE-HDBK-1122-2009 Module 2.05 Contamination Control Study Guide 2.05-11 • Cover piping/equipment below a work area to prevent dripping contamination onto less contaminated areas. • Cover/tape tools or equipment used during the job to minimize decontamination after the job. • Follow good work practices such as good housekeeping and cleaning up after jobs. • Confine the spread of radioactive material releases by a sound preventive maintenance program. • Control and minimize all material taken into or out of contaminated areas. BASIS FOR ESTABLISHING PROTECTIVE CLOTHING REQUIREMENTS In order to prevent radioactive contamination from getting on or into the body, protective clothing requirements must be established where the potential exists. There are several basic factors which determine the type and extent of protective clothing required: • type and form of contamination • levels of contamination • type of work being performed Some additional factors to consider include the potential for increased levels of contamination, the area of the body at risk, and competing hazards, i.e., heat stress, asbestos, etc. Once the types of protection needed are established, the most efficient protective clothing must be selected from the different articles of protective clothing available for use. A discussion of the controls/clothing types for specific areas of the body follows. 2.05.07 List the basic factors which determine protective clothing requirements for personnel protection. DOE-HDBK-1122-2009 Module 2.05 Contamination Control Study Guide 2.05-12 Whole body protection A lab coat provides protection from low levels of contamination and is only applicable when the potential for upper body contact with contaminated surfaces is very low. In general, lab coats are worn for hands-off tours and inspections in areas with removable contamination at levels 1 to 10 times the values in Table 2-2 of the Radiological Control Standard, or during benchtop, laboratory fume hood, sample station, and glovebox operations.

Coveralls provide protection from low to moderate levels of DRY contamination protection. Protection is low when body contact with contaminated surfaces is prolonged (since contamination can be ground into or through the cloth) and when the surface is wet. The degree of protection can be increased by use of more than one pair at a time to protect the body. Cloth coveralls are permeable, and so are not effective against radionuclides with high permeability properties (gases, tritium, etc.). Plastics coveralls provide protection from high levels of dry contamination and wet contamination. They provide limited protection from tritium and other highly permeating radionuclides (which may be transported through coveralls to the skin surface). Disposable coveralls, e.g., tyvek suits, provides moderate protection from radioactive contamination and are used for work involving mixed hazards, i.e., asbestos, PCBs, etc., where reuse is not desirable. Disposable coveralls can be fairly easily torn. It should be noted that at a minimum, outer personal clothing should not be worn under protective clothing for entry into High Contamination Areas or during work conditions requiring a double set of protective clothing. Sites may choose to be more restrictive as necessary to minimize potential skin/clothing contamination. Hand protection Surgical gloves are a minimal requirement normally used in only light contamination work areas which require a high degree of dexterity. Surgical gloves are fairly easily torn or punctured. Rubber gloves are l