CHAPTER 11 final.docx

CHAPTER 11 – TABLE OF CONTENTS

List of Tables

List of Figures

Figure 11.1: NIST Organizational Chart (partial) 28

Chapter 11 – Record of Revisions

11-

11 RADIATION PROTECTION AND WASTE MANAGEMENT

11.1 Radiation Protection

NIST has a structured radiation protection program that supports all aspects of NBSR operations. The health physics staff is equipped with sufficient radiation detection equipment to determine, control, and document all occupational radiation exposures. NIST also has established policies that employ the ALARA concept in all operations at NBSR. An environmental monitoring program is also in place to assure that potential radiation exposures in unrestricted areas surrounding the reactor facility are well within regulations and guidelines.

11.1.1 Radiation Sources

In this section, the sources of radiation that are monitored and controlled by the radiation protection and radioactive waste management programs are described. Radiation sources at NBSR can be classified into four general classes:

· Calibration & check sources

· Startup, and other sources used for instrumentation and nuclear support functions

· Airborne, liquid, and solid radiation sources from reactor operations

· Radiation sources produced within the experimental facilities.

The major radionuclide constituents of the radiation for which monitoring and control by the radiation protection program are or could become necessary are summarized below: Comment by Myers, Thomas J.: Changed after discussion with Health Physics. Including old radionuclide in the SAR does not have much value. I will delete such values from this chapter. DONE.

Major Sources of Radioactivity

Airborne

Liquid

Solid

41Ar, 3H

3H, 110mAg, 64Cu, 66Cu

60Co, 55Fe, 59Fe, 65Zn

Monitoring and/or control can depend upon exposure to the source, e.g. maintenance on a fluid system. Typically, a dose estimate would be included prior to commencement of an activity that could generate a non-trivial exposure to a source of radiation and that estimate could include a sample, e.g. water sample of thermal shield cooling water as a precaution for changing a fluid valve. Other sources of radioactivity that are found in various reactor and support systems, but are of negligible consequence to occupational or environmental doses are listed below.

NBSR Systems and Radiation Sources (Bolded nuclides are major components)

Primary coolant: 3H, 16N, 24Na, 41Ar, 51Cr, 54Mn, 56Mn, 60Co, 122Sb, 124Sb Primary pipe (internal contamination): 3H, 51Cr, 60Co, 65Zn Helium sweep: 41Ar, 85mKr, 87Kr, 88Kr, 131mXe, 133Xe, 135Xe, 135mXe, 138Xe, 138Cs Thermal Shield Cooling System: 64Cu, 66Cu, 65Zn, 110mAg, 16N, Comment by Myers, Thomas J.: Probably have changed. Confirm. DONE.

Reactor shield plug/refueling plug: Al and steel activation products, 14C Air: 3H, 38Cl, 41Ar, 82Br, 138Cs CO2 sweep gas: 35S, 38Cl, 41Ar, 82Br Storage pool: 3H, fuel piece cutting products from aluminum activation Fuel pieces (6061 aluminum, stainless steel): 54Mn, 55Fe, 60Co, 63Ni, 65Zn Resin beds: 60Co, 65Zn Neutron guides: 58Co, 59Ni, 65Zn Pneumatic system: 60Co, 65Zn, 110mAg

11.1.1.1 Calibration, Check, Startup, and Other Radiation Sources

The primary reactor startup source is the combination of irradiated fuel and D2O, which provide photoneutrons. The backup startup source is an AmBe neutron source with an activity of approximately 2.0 Curies (Ci) and it has a 4.5 MeV average fast neutron emission. It is stored in a shielded container.

Instrumentation check and calibration sources used to support reactor and radiation protection activities are maintained under the Byproduct Materials license (SNM-362) that was issued by the NRC. These include 60Co and 137Cs sealed sources of various strengths (μCi to kCi activities) as well as sealed sources of other radionuclides, and a variety of unsealed sources used primarily for the calibration of laboratory instrumentation (e.g., 137Cs, 60Co, 152Eu, 14C, 99Tc, 3H, and 90Sr). These sources are mostly in the nCi to μCi activity range, in both solid and liquid form. Comment by Myers, Thomas J.: Confirm that these are still in use. Confirmed. DONE. Comment by Myers, Thomas J.: See previous comment. DONE.

The Safety Evaluation Committee reviews fissile and fissionable materials used in experiments that are inserted into the reactor for compliance with the Technical Specifications. These sources are acquired and maintained under NRC License SNM-362. They consist mainly of fission chambers and foils that are used to monitor or calibrate neutron beams and fields. These sources are strictly controlled and periodically inventoried.

Sources produced by the reactor that are related to experimental programs range in activity from aCi (1 attoCurie = 1.0 x 10-18 Ci) to kCi. These sources may consist of any chemical element in any physical form. Access to these sources is controlled and they are subject to the radioactive material accountability program.

11.1.1.2 Airborne Radiation Sources

The principal airborne sources of radioactivity associated with the operation of the NBSR are 41Ar and tritium (3H). The only authorized release path for air from the various confinement building ventilation systems is via the building stack exhaust, which has a nominal flow rate of 30,000 cfm (850 m3/Min). Annual emissions of 41Ar typically ranges from 800 to1200 Ci and 3H release ranges from 400 to 800 Ci. This constitutes a dose of less than 2 mrem of exposure to the closest member of the public, which is less than 2% of the NRC dose limit to the public. This analysis was performed with the EPA COMPLY computer code using local wind rose data and computing the dose based on the closest resident in each wind sector, which constitutes conservative analytical boundary conditions.

Monitoring in both the stack and in the building ventilation systems utilizes both installed and periodic sampling. This provides redundant methods for assessing both occupational and public exposure. Occupational exposure is discussed below.

11.1.1.2.1 41Ar Sources

Argon (40Ar) is about a 0.93% natural constituent of air. Any air volume that is exposed to neutrons will produce 41Ar by the 40Ar(n,γ)41Ar reaction. 41Ar is a strong beta and gamma emitter with a half-life of 110 minutes. At NBSR, extensive engineering and procedural measures have been taken to minimize 41Ar production. These include:

· Maintaining all heavy water primary systems under positive helium pressure to minimize air in primary water, potentially exposing it to a neutron flux.

· Conducting all maintenance activities on primary systems in a way that minimizes air intrusion.

· Using cover gases like CO2 to exclude air wherever practicable in neutron irradiated volumes, such as the cavity around the reactor vessel, and using CO2 as the driving gas for pneumatic samples being irradiated.

· Sealing all penetrations and openings to the extent practicable to exclude air intrusion.

· Designing experiments to minimize neutron irradiated air volumes.

· Regularly assessing 41Ar production to verify the effectiveness of the existing reduction measures.

Production of 41Ar at the NBSR is primarily due to the presence of air in the cavity around the reactor vessel. Production associated with experiments is less than 0.1 % of the total because of the smaller irradiated air volumes and because of lower neutron fluences associated with most experiments. The external exposure rate from 41Ar is minimal because the concentrations of 41Ar in the building are less than 1 Derived Air Concentration (DAC) and the building volume represents a small fraction of a “semi-infinite” cloud. Actual dose rates to a person in the building from a uniform DAC cloud would be less than 0.2 mrem/hr. Personnel dose rates from typical 41Ar levels observed within the NBSR confinement building have been less than 0.004 mrem/hr. This low level, when combined with typical occupancy times and reactor operating frequency results in an annual personnel exposure from this source that is less than 2 mrem. Direct measurements have demonstrated that the calculated values are conservative.

11.1.1.2.2 Tritium

Tritium is produced by the 2H(n,γ)3H reaction in the heavy water moderator/coolant of the reactor. This produces a primary coolant tritium concentration of 0.3 Ci/liter/yr. As an ALARA measure NIST replaces the heavy water at intervals chosen to limit tritium exposure. With a maximum production concentration of 5 Ci/liter, the radioactivity concentration and exposures discussed below for NBSR would increase by no more than a factor of 5.

During normal operations, the primary release pathway for tritium results from helium leakage into the ventilation system. Since the helium is used to minimize air intrusion into the primary cooling system, it can become saturated with heavy water. Activation of the heavy water hydrogen produces tritium. Secondary pathways can include various activities, such as refueling or any maintenance activity that exposes heavy water to the air. Abnormal loss conditions, such as a seal failure or a primary coolant boundary failure would be quickly identified by the various monitoring or leak detection systems. The airborne tritium monitoring system at NBSR is capable of detecting a few milliliters of leakage that can occur by water evaporation.

Confinement building tritium levels at a nominal primary concentration of 1Ci/liter are typically less than 1.0% DAC. Since the operating staff is in the building less than 1500 hours per year, this represents an annual dose commitment of less than 40 mrem. Bioassay data of the operating staff confirms that most exposures are well below this value. All other personnel are in the confinement building a much smaller fraction of time, and their annual tritium exposures are much less than 1 mrem. Local airborne exposure to heavy water sources by reactor operators during certain activities, such as refueling, can increase their annual exposure from tritium sources, but normally not in excess of 100 mrem. Comment by Myers, Thomas J.: Confirm this. See above comment. Confirmed. DONE.

Abnormal or transient conditions would increase these airborne tritium levels. Previously, when the ventilation system for NBSR was shut down for remediation over a five-day period, the tritium levels slowly approached DAC values. Also, when an auxiliary cooling loop had excessive heavy water leakage, the local airborne tritium levels increased to 1 μCi/m3, which corresponds to 5% DAC.

From a public dose perspective, tritium represents about one-tenth of the dose from 41Ar, assuming equal release activities. Conducting operations in a way that minimizes 41Ar production, even if that results in some increased heavy water loss and minor increases in tritium exposure, results in minimized collective dose because the increased occupational dose to the limited number of operational staff is more than offset by the reduced collective dose to the public. Therefore, ALARA efforts to reduce tritium losses, particularly through ventilation system modifications, must be tempered by possible related increases in 41Ar emissions.

11.1.1.2.3 Fission Products

Noble gas fission products can be detected in the helium sweep system that is maintained over the primary coolant. Those detected radionuclides include gases of Xenon, Krypton, and 138Cs (a daughter product of 138Xe). Using the typical make up rate for the helium system, it is calculated that less than 0.1 Ci of these radionuclides are released annually. These release concentrations are so low (less than 10-10 µCi/ml) that they represent a negligible contribution to the total gaseous emissions.

11.1.1.3 Liquid Radiation Sources

The dominant liquid radionuclides of the NBSR are tritium and 16N. Some other minor liquid sources are also discussed in subsequent sections.

11.1.1.3.1 Reactor Primary Coolant

The NBSR primary coolant consists of high purity heavy water. Its primary radionuclides come from the following reactions:

· 3H, produced via 2H(n,γ)3H, a low energy beta emitter

· 16N, produced via 16O (n,p) 16N, a high energy beta and gamma emitter

· 24Na, produced via 27Al(n,α)24Na, a high energy beta and gamma emitter

· 28Al, produced via 27Al(n,γ)28Al, a high energy beta and gamma emitter

· 60Co, produced via 59Co(n,γ)60Co, a low energy beta and high energy gamma emitter

· 51Cr, produced via 50Cr(n,γ)51Cr, a low energy gamma emitter

Other radionuclides present in the primary coolant include 65Zn, 56Mn, 99mTc, and 122Sb. These are suspended corrosion products that are activated by neutrons, but are a minor portion of the total liquid radiation source.

During reactor operations, 16N is the primary source of external radiation exposure from the water in the primary piping system. Its short half-life of 7 seconds means that exposure from this source diminishes very rapidly after the reactor is shutdown. At the NBSR, there are two areas where an exposure potential from this source exists, C006/C007(Process Room) and C03 ( Monitoring and Sample Room). The Process Room contains all of the primary water pumping and processing systems. Dose rates in the process room during 20 MW operation range from a few mrem/hr in relatively shielded or distant zones, to 100 rem/hr in close proximity to the primary reactor piping nearest the reactor vessel. Entries into the high radiation areas of this room are rare. If entry to those areas is necessary, the reactor power is reduced whenever possible. The 16N source is present in C03 only when the primary sampling system is in operation. General area dose rates when sampling is done at power are less than 100 mrem/hr.

Tritium is produced in the primary D2O at a rate of 0.3 Ci/kg per year based on 5000 MWd per year. This is equivalent to about 3 Ci/MWd for the 12,000 gallons (45,500 liters) in the primary system. A 1 Ci/liter concentration is used for a number of reference calculations in this document. All used heavy water is stored onsite until transferred to authorized processors for recycle or disposal.

As discussed above, tritium is a significant source of exposure from airborne contamination due to evaporation of tritiated heavy water. Exposure by direct contact, thru skin absorption, could result in significant exposures. At a nominal primary coolant concentration of 1 Ci/liter, this represents an exposure of 62 mrem per milliliter of heavy water absorbed. Therefore, any work involving potential exposure by this mechanism requires control measures such as containment, eye protection, gloves, and protective clothing, to minimize and prevent such an occurrence. Individuals that perform this work are required to periodically provide tritium bioassays. The other radionuclides present are at such low concentrations that they represent a negligible residual contamination problem.

24Na is present on the order of 0.1 mCi/liter. It represents a transient external exposure source term in the process room. Due to its short half-life (15 hours), as an ALARA measure, work in the process room typically is limited for the first day following shutdown.

51Cr represents the highest activity, longer-lived (half- life of 27.7 days) primary system contaminant other than tritium. It is present in the primary at a concentration approaching

0.001 mCi/liter. Since 51Cr emits a low energy gamma, it is almost totally self-shielded by the primary system components. As a contamination source, it is the dominant radionuclide, based on activity, by at least a factor of ten for freshly removed primary components. After several months, 65Zn and 60Co become the dominant residual sources of contamination due to decay of the 51Cr. These radionuclides present an exposure primarily in terms of local external dose due to system contamination. Local “hot spot” radiation sources, at valves, heat exchangers, filters, and resin beds range from a few mrem/hr to 500 mrem/hr. Components that have the higher dose rates, such as primary coolant filters and the resin beds, have local shielding to reduce the radiation levels to less than 5 mrem/hr. Exposures from other “hot spots” are controlled through local posting of the areas concerned. The general area dose rate in the process room due to the cumulative effect of these long-term internal contaminants ranges from a few mrem/hr. to about 20 mrem/hr. This room is routinely surveyed, and the survey data is made available for any work performed in this area Comment by Myers, Thomas J.: No record of this. N16 will produce 100 R areas near the main pumps, but the highest does measured during the April 2013 shutdown was 500 mR, near HE2. DONE.

11.1.1.3.2 Thermal Shield Cooling System

Cooling of the thermal shield uses purified, light water. The primary radionuclides present in the cooling system include the following:

· 16N, which presents a local dose rate at the ring header, during power operations at various tube locations around the header.

· 66Cu and 64Cu, which are the primary short-term sources of external exposure following reactor shutdown. The ALARA measure for routine maintenance on this system is to delay the start of that work for 36 to 48 hours. 64Cu concentrations up to 1 mCi/liter have been observed.

· 65Zn and 110mAg, which are the long-term radionuclides in the system. Maintenance procedures are formulated to maximize containment of all fluid transfers to control this potential source of contamination.

While control of the thermal shield cooling water chemistry is used to minimize the concentration of the aforementioned radionuclides, the water and piping of the system still is the most significant source of routine external radiation exposure in C100. The supply and return headers for the system, located high on the biological shield, generated a dose rate of approximately 25 mrem/hr at 30 cm in February 2015. Local shadow shielding is employed, where practicable, to reduce exposures to experimenters working for extended periods at their desks/workstations in the room. Dose rates of 0.5 mrem/hr, or less, were recorded early in 2015 in these areas.

11.1.1.3.3 Fuel Storage Pool

Water in the spent fuel storage pool is contaminated due to the transfer of spent fuel elements to the pool and from cutting operations performed on aluminum of the spent fuel. The major radionuclide present is tritium, measured in February 2015 at a concentration of approximately 0.16 μCi/ml. Since the volume of the spent fuel pool is 33,000 gallons (125,000 liters), the tritium activity in the pool is approximately 20 curies.

Extensive drying of each transferred element is performed as an ALARA measure to limit the amount of transferred tritium. However, trapped or absorbed D2O evolves from each element over several months, a contribution to the tritium concentration in the pool which cannot reasonably be prevented. Liquid releases of less than allowable concentrations and which result in a total activity less than the annual release limit are performed annually. Comment by Myers, Thomas J.: Implies this is done routinely. Change. DONE.

Cutting of spent fuel elements releases various chips and small particles of aluminum to the pool water. These particles contain the normal activation constituents of aluminum, which are 51Cr, 60Co, and 65Zn. An aggressive spent fuel pool vacuuming program and spent fuel pool filtration maintain these radionuclides to levels of less than nanocuries per liter.

11.1.1.3.4 Other Systems

The H2O in the Thermal Column Tank Cooling System is subjected to a neutron flux, but potential exposures from the water in the piping are not considered significant. The cold sources comprise two cryostats, two liquid/gaseous hydrogen loops cooled by helium, and a single refrigerator. The liquid hydrogen is subjected to a high neutron flux. Consequently tritium is produced in this fluid resulting from the natural occurrence of deuterium in hydrogen. However, this is a closed system designed not to require opening for any kind of maintenance, except for removal of the cold source. Therefore, no operational radiological consequences associated with this system fluid exist.

The Helium Cooling System for the cold source has no exposure to neutrons or to any contaminated system. Therefore, this system has no radiological consequences.

The Liquid Waste System comprises selected drains in the laboratory wing and all light water drains from the confinement building. These drains are routed to the liquid waste collection facility. The dominant radionuclide in the collection system is tritium, and is discussed further in Section 11.2.6.

11.1.1.4 Solid Radiation Sources

Solid sources of radiation at NBSR result from reactor operations. The sources range from very low specific activity, such as used rubber gloves from handling potentially contaminated materials, to intermediate activity items such as activated foils from experiments, and to the high activity spent fuel from the reactor. These sources are described in the following subsections.

11.1.1.4.1 Fuel Elements

All operations involving movement of irradiated reactor fuel elements are performed underwater or in a shielded component, which provides the needed shielding. The non-fuel element portions of the spent fuel are removed by underwater cutting and are disposed of separately from the fueled portions of the fuel elements. The radioactivity, at the time of shutdown, in the pieces from a single fuel element that was used for eight operating cycles is shown in Table 11.1. Only the longer-lived radionuclides are tabulated because of the time delay to shipment. This delay is a minimum of approximately 280 days, but is more typically greater than two years. Table 11.1 shows that when shipments are normally made, the shipment total activity is dominated by 55Fe, 60Co, and 65Zn. For elements used for fewer operating cycles, these values would be reduced, because the neutron exposure time (production time) would be less. Personnel exposure when performing spent fuel handling operations is minimal because of the conditions described at the beginning of this paragraph and in the next paragraph.

The fission product inventory for one NBSR fuel element is listed in Table 11-2. Radiation dose rates from these elements are the primary issue for personnel protection. The room through which the elements are transferred is controlled as a Very High Radiation Area during these transfers. All handling of the fuel in the storage pool is monitored with area monitors or survey instruments as a precaution to ensure the fuel element being handled remains adequately shielded.

New NBSR fuel elements nominally contain 350 grams of 235U. Upon receipt they are surveyed for both radiation level and contamination. Prior to insertion into the reactor, each element undergoes a thorough quality assurance evaluation, but dose to operators when handling the new fuel is negligible, since there are no fission or activation products present.

11.1.1.4.2 Reactor Shims

Control shims are the only other high activity component routinely removed from the reactor. This occurs every 4 to 5 full-power years. This process is accomplished by procedure under the control of a radiation work permit with strict adherence to ALARA policy. All disassembly work is conducted from the reactor top area through access ports to the reactor vessel, using extended tools. Transfer operations are done remotely using cameras and remote crane controls from a shielded location when practicable. All unnecessary personnel are excluded from affected areas during this work.

After allowing for a minimum decay period, while stored under water in the spent fuel storage pool, the stainless steel hubs for the shims are separated from the Cd-Al shim arm and shipped with the other radioactive non-fuel element metal pieces. The Cd-Al shim arm is stored in the storage pool or in dedicated dry vertical or horizontal storage.

11.1.1.4.3 Other Radioactive Solids

Other radioactive solids that could contribute to personnel dose and waste volume include: Comment by Myers, Thomas J.: See last bullet. DONE.

· Reactor primary resins, which are replaced very infrequently on the order of once every 10 to 20 years

· Reactor primary filters, replaced as needed, usually once or twice a year

· Filters and resins from other systems

· Shielding plugs and related neutron beam shields

· Experiments, or experimental components removed from high neutron flux locations

· Activated experiment samples

· Miscellaneous contaminated materials, such as laboratory waste

· Emergency response.

The radioactivity in these items range from curie quantity material for items such as resins, to barely detectable levels in other items, which constitute the bulk of the waste volume. 60Co in the activated metals, resins, and much of the waste is the primary contributor to personnel external dose rate. This material is stored in restricted areas where access and area dose rates are controlled to limit personnel exposure. Local shielding is used as necessary to limit areas to less than Radiation Area conditions. Sometimes this material is stored in shielded casks or the dry horizontal or vertical plug storage located in the G-Wing of Building 235. Bulky items with low-level activation, typically experiment shields and components, may be stored in Building 418, which is adjacent to the reactor building. Both of these storage areas are maintained as restricted areas.

11.1.1.4.4 Solid Radioactive Waste Disposition

All radioactive waste is disposed of in accordance with 10 CFR 20, Subpart K. Solid waste is transferred to organizations specifically authorized or licensed to receive the material, such as the Department of Energy. Materials designated as radioactive waste are transferred to the H wing of the facility for characterization, packaging, and preparation for transfer to authorized recipients. All of this waste is 10 CFR 61.55 Class A waste, unless otherwise noted.

Routine waste collection and screening:

Any reactor support maintenance that requires a Radiation Work Permit and results in non-routine waste that is specific to that job is monitored by Health Physics and transferred to H100 as radioactive waste. Health Physics would respond to any special needs or requests for non-routine waste collection.

10 CFR 61.55 Class B waste is seldom generated and is less than 0.001 per cent of the volume of waste transferred to H100. Waste of this type, e.g. an activated valve component, is transferred from its location to H100 only after a review, the completion of which could require specific measures, such as use of a cask, to minimize personnel dose. Access to the H wing annex is strictly controlled. Annual radioactive waste volumes and activities are typically in the range of 126 to 423 ft3 (11 to 36 m3) and are less than 1 Ci. In years when unfueled element shipments occur, or major facility modifications performed, larger quantities of radioactive material will be generated. Based on past experience, these are infrequent occurrences on the order of once every 5 or more years.

All reactor support systems that flow either light water or heavy water circulate through cartridge-type particulate filters and through H-OH bead-type ion exchange resin beds. Reactor operations personnel periodically replace these filters and resins with new media as part of the routine maintenance of those systems. Each resin bed change produces approximately 5 ft3 of resin beads. These filters and resin are stored in a shielded area of H100 until shipment for disposal.

Particulate filters from some ventilation systems are checked annually and replaced as needed. Those filters that service areas with radioactive materials are collected and transferred to H100 as radioactive waste.

Radioactive liquid wastes are typically transferred to H100 in 55-gallon drums for disposal, if the water cannot be evaporated. If evaporated, the residue is collected and transferred to H100.

Waste solutions from laboratory analyses and separations are neutralized by the responsible researcher and then accepted by Health Physics for transfer to H100.

Discrete radioactive sources that have decayed or no longer are needed are only picked up by prior arrangement with Health Physics, as the specific activities usually require special handling and disposal procedures, and in many cases the specific activity results in Class B or Class C 10 CFR 61.55 classification. These are not generally transferred to the H100 annex, but are kept in secure storage until they can be disposed of as waste.

The redesign and upgrading of reactor experimental facilities often causes a reconfiguration of instrument shielding. The replaced shielding generally has slight long-lived activity that is not removable and does not contribute to personnel doses. The shielding is usually bulky and heavy and is generally kept in a designated storage area other than H100 until it can be disposed of as waste. Occasionally, some component or block of obsolete shielding will have a relatively high induced activity; this type of waste would be transferred directly to the H100 annex.

All vacuum cleaners with potentially contaminated content are emptied by Health Physics or similarly authorized personnel and transferred to H100.

11.1.1.5 Radiation Sources from Experimental Facilities

NBSR is primarily used for research purposes. The majority of this research involves the use of neutrons to study material constituents, processes, and structure. Therefore, radiation sources will be present in the experimental facilities supporting these activities. These sources are described in the subsections below.

11.1.1.5.1 Neutron Beams

Neutron beams at the NBSR typically range from a few square millimeters to 200 cm2. Beams with an in-beam dose rate in excess of 100 mrem/hr and accessible are designated as High Radiation Areas. Section 11.1.5.1 has a discussion of beam controls. A characteristic of neutron beams is that the radiation field outside of the beam is typically less than 5 mrem/hr. Sometimes experimental samples or equipment, such as collimators or filters, can result in Radiation Area or possibly High Radiation Area conditions in areas near the beams. These areas are controlled as required by applicable regulations and NCNR policies. Non-beam related and short-term experiments are shielded and controlled to minimize personnel exposures.

11.1.1.5.2 Thermal Column Facility

This facility was used to provide highly thermalized neutron beams, but has been placed in a temporary decommissioned status.

11.1.1.5.3 Pneumatic System and In-core Exposure Facilities

Experiments utilizing these facilities are highly variable, frequently producing multi-curie activity sources. All elements of the activity, facility usage, disposal, and potential personnel exposures are addressed by technical review and administrative authorization processes. Holding the source in a shielded configuration to allow sufficient decay prior to direct manipulation, processing, or analysis is the primary ALARA technique used in these situations.

11.1.1.5.4 Cold Neutron Experiments

The cold neutron guides are fully shielded to the point of neutron beam extraction, wherever possible. At the entry wall to the Guide Hall, the unshielded dose rate from a typical guide is 300 mrem/hr (neutron) and 100 mrem/hr (gamma) at one meter from the guide. The guides vary in size up to 14 in2 (90 cm2), and have a cold neutron flux of approximately 5x109 n/cm2-sec. The in-guide gamma and fast neutron flux rates decrease by the square of the guide length for straight guides. For filtered guides with either bulk or optical filters, the fast neutron and core gamma components of the flux are largely removed. All eleven guides have shutters that completely stop the radiation from the core after the shutter is closed. These shutters are key controlled, and have status indication (opened or closed). When closed, the design allows unrestricted disassembly and work on experiments for a particular guide.

11.1.2 Radiation Protection Program

The Radiation Protection Program comprises the “Health Physics Instructions”, “Health Physics Procedures”, “Health Physics Standard Operating Procedures”, Gaithersburg Radiation Safety Division procedures, “Radiation Safety Good Work Practices”, training for occupational workers and users, and other documentations, e.g. beamline policies and procedures.

11.1.2.1 Radiation Protection Program Staff

Administration of the radiation protection program is performed by the reactor health physics team. This structure is shown in Figure 11.1.

A senior health physicist serves as Chief, Reactor Health Physics (RHP), and is responsible for the implementation of the Radiation Protection Program for the NBSR facility. Since maintaining the reactor operating license is the responsibility of the director of the NCNR, the Chief, RHP reports to the Director, NCNR. The activities of Reactor Health Physics include: Comment by Myers, Thomas J.: Still the title? No, per P. Brand. DONE.

· Calibration of survey instrumentation

· Effluent and environmental monitoring

· Radiation and contamination surveys

· Personnel monitoring

· Review of proposed experiments and compliance reviews of operating experiments

· Radiological sample analysis

· Training of reactor staff, visiting researchers, and NIST support staff

· Safety Evaluation Committee membership

Reactor Health Physics is typically staffed with two to four Health Physicists and two to four Radiation Protection Technicians. All of the health physicists meet the qualification requirements of the Office of Personnel Management for health physicists, GS-7 or higher and typically have sufficient training and experience to meet the American Board of Health Physics requirements for comprehensive certification.

Radiation protection technicians meet the qualification requirements of the Office of Personnel Management for Physical Science Technicians, GS-5 or higher. They receive additional on-the-job training specific to the NBSR prior to becoming fully qualified as a Reactor Radiation Protection Technician.

11.1.2.2 Plans and Procedures

Plans and procedures for the implementation of the Radiation Protection Program relating to reactor activities may be written by either the operations or the health physics staff. Such plans and procedures, regardless of authorship would be reviewed by appropriate members of both staffs as a minimum, and usually by the Safety Evaluation Committee (SEC) as well. Obtaining final approval from the SEC and document control is assigned to an individual on the operating staff. That person retains the original, signed document as the master and ensures the appropriate distribution to the staff is made. Plans and procedures not directly related to reactor operations, such as instrument calibration, routine shipping and receiving of radioactive materials, are maintained under the NIST materials license and controlled through an analogous structure involving the NIST Radiation Protection Officer and the NIST Ionizing Radiation Review Committee. Plans and procedures that are needed under both programs are either dual approved or are maintained as separate but consistent procedures.

11.1.2.3 Safety Evaluation Committee and Safety Audit Committee Comment by Myers, Thomas J.: See NBSR 15. DONE.

The SEC provides the NCNR with a method for the independent review of the safety aspects of reactor facility operations and health physics, in accordance with Technical Specification 6.2.

The Safety Assessment Committee (SAC) provides the NCNR with a method for performing independent audits of various aspects of the reactor facility, in accordance with Technical Specification 6.2.

11.1.2.4 Interdiction Authority

Any licensed reactor operator, any member of the NCNR management, and any NBSR staff Health Physicist have the authority to interdict and terminate any activity related to the use of the reactor or the use of radioactive materials within the reactor facility that is judged unsafe or that could reasonably lead to an unsafe condition or violation of NRC regulations. Only the licensed operators have jurisdiction over the operation of the reactor itself. Comment by Myers, Thomas J.: Is this advisable? No change per 12/2/15 meeting. DONE.

11.1.2.5 Radiation Safety Training Program

To obtain unescorted access to Building 235, individuals must be trained in the following subject areas:

· Basic radiation science

· Meaning and proper response to radiation signs

· Proper use of assigned radiation dosimetry

· Proper response to emergency alarms

· NBSR procedures related to their duties

Individuals who have duties relating to the direct use of radioactive materials or reactor experiments are given additional training, which includes:

· Radiation science specific to their radioactive material usage

· Radiation protection techniques for their specific duties

· Proper use of the appropriate radiation survey instruments

· NBSR procedures and NRC regulations specific to the materials’ usage pertaining to usage limitations, ALARA requirements, and material control

Individuals requiring training include the NCNR staff who routinely work in Building 235, visiting researchers using NBSR experiment facilities, selected plant personnel, and security and fire protection personnel who have unescorted access within Building 235. All trained personnel receive refresher training every 24 months, not to exceed a 30-month interval. Reactor operating staff and the reactor Health Physics staff maintain their radiation safety skills through ongoing training (i.e., on the job training).

11.1.2.6 Records

The following records are retained for the life of the facility, as prescribed in 10 CFR 20 Sections 2101 through 2110.

· Personnel exposure records

· Radioactive emission determinations and related calculations

· Survey data in areas where radioactive materials are used and any contamination events related to personnel exposure in those areas

· Results of air sampling, surveys, and bioassays required by 10 CFR 20.1703(a)(3)

All other radiation protection documents (survey records, calibration records, work logs) are retained for at least 10 years. Records relating to 10 CFR 21 issues are retained for at least 5 years.

11.1.2.7 Part 21 Program

At NBSR, a senior staff member has been designated the responsible individual for receiving, reviewing, and reporting to the NRC any matter relating to requirements under 10 CFR 21. NBSR has promulgated procedures reflecting the 10 CFR 21.21 requirements. A notice is also posted in a prominent location outlining the reporting requirements and the staff responsibilities required under 10 CFR 21.6. Comment by Myers, Thomas J.: I do not believe this, esp the all staff part. Trying to review the training modules for content. Deleted two sentences per 12/2/15 meeting. DONE.

11.1.3 ALARA Program

The NBSR ALARA program, as required by 10 CFR 20.1101, addresses all aspects of NBSR operations. These activities include specific emphasis on proposed new experiments, planned activities involving significant potential personnel exposures, ambient radiation environments within NBSR, and retrospective reviews of occupational and public doses. The Reactor Senior Health Physicist and the SEC have primary responsibility for the prospective analyses, while the SEC and SAC have primary responsibility for retrospective ALARA reviews and audits. Comment by Myers, Thomas J.: Standalone document? Replace with link if there is one. NONE. DONE.

Activities involving the potential for exposures greater than 0.5 person-rem usually require a formal operating plan, a meeting of the involved personnel to discuss the plan, identification of methods for reducing exposures, and specific oversight by the Health Physics staff to ensure that recommended ALARA measures are implemented. Activities involving less potential for personnel exposure have less formal planning and a pre-operational review.

Examples of various ALARA activities implemented at NBSR are identified in Section 11.1.1 above. Engineering controls, such as shielding are utilized to the maximum extent practicable to minimize radiation levels in work areas. Through review of regular surveillance surveys, the Chief, Reactor Health Physics, will identify unusual radiation conditions or work practices and recommend improved methods, as well as “lessons learned” feedback to the staff involved.

Minimum ALARA goals are to improve on past performance for ongoing activities and to achieve the lowest exposures by thorough planning. Explicit numerical ALARA goals are rarely established due to the non-routine nature of much of the NBSR research environment. The typical monetary equivalent expended per person-rem avoided is greater than $100,000. The NBSR ALARA goal is to limit radiation doses in unrestricted areas to 10% of 10 CFR 20.1301(a)(1) using shielding and procedures. An explicit limit for public dose from gaseous effluents has been established at 10 mrem per year, pursuant to 10 CFR 20.1101(d).

11.1.4 Radiation Monitoring and Surveying

Health Physics supports the NCNR by maintaining portable and fixed monitoring instrumentation as well as laboratory radiological analysis instrumentation. Similar instrumentation located in the Radiation Physics Building provides on-site back up for most of the NBSR instrumentation requirements. Laboratory instrumentation available includes low background proportional counters for alpha-beta counting, liquid scintillation systems for tritium and other low energy beta counting, gross beta counting, and beta spectroscopy, and InGe gamma spectroscopy systems.

Health Physics instruments used for quantitative radiation measurements are calibrated and performance checked at specific frequencies for the radiation measured. Health Physics conducts internal quality control programs to assess the reliability and stability of the laboratory radiological analysis instrumentation. All calibration sources are traceable to either manufacturer supplied reference standards or to national reference standards, such as NIST Standard Reference Materials or NIST primary standards. Radiation instrumentation that provides engineered safety functions related to the operation of the reactor are described elsewhere.

11.1.4.1 Area Radiation and Contamination Monitoring

Contamination and radiation surveys are conducted weekly during operation. During extended shutdowns, alternative schedules are established, which are usually more frequent. Areas that are surveyed include the accessible areas of the Confinement Building and other radioactive material work areas, with emphasis on those areas that pose the greatest potential for changing conditions. These would include the reactor systems and experiments. Additional surveys are performed on an as-needed basis. Typically radiation surveys of active work locations are performed daily.

Spot contamination measurements are routinely performed with paper smears over a 100 cm2 area. Large area coverage contamination surveys are performed with floor contamination beta monitors on an as-needed basis.

A full range of portable beta, gamma, and neutron survey instruments are available at NBSR. These include G-M detectors, ion chambers, proportional detectors, plastic scintillation gamma detectors, NaI detectors, and BF3 and 6LiI moderated neutron instruments. Selected portable survey instruments are positioned at various locations around the facility and near experiments for ready use.

Fixed gamma area radiation monitors are positioned at ten selected locations in the confinement building. These ten locations are: three on the C200 level, which includes the control room, top of the reactor, and west wall of C200; four on the C100 level, which includes the experiment and neutron beam room; two in the process room, which contains primary cooling water systems; and one in the spent fuel storage pool area. Alarm set points are specified in NBSR procedures. Typical alarm settings are 5 mrem/hr and adjusted as needed for non-routine activities, generally with the objective of identifying unusual changes in radiation conditions.

Monitors in the spent fuel storage pool area are positioned to detect: increased radiation levels associated with handling of irradiated fuel elements; a loss of shielding from a loss of pool water; or criticality in the pool. The monitor installed in the fuel storage area serves as a criticality detector.

Fixed personnel contamination monitors are located at the entrances to the reactor confinement building, and elsewhere on an as-needed basis. Three types of monitors are available for this use. They are:

· Hand and foot monitors using G-M or proportional detectors

· Portal monitors using G-M or plastic scintillation detectors

· Half-body contamination monitors using sealed tube or gas-flow proportional detectors

11.1.4.2 Air Monitoring

Conditions requiring airborne radioactivity monitoring under 10 CFR 20.1502(b) are rarely present at the NBSR. The two primary airborne radionuclides present at the NBSR are 41Ar and 3H. For 41Ar, area radiation monitors are used to control personnel radiation exposures. Cary ion chambers or gas Marinelli chambers are used to determine airborne activity concentrations. An installed gas-flow ion chamber system takes samples from representative areas of the building and from the ventilation system for tritium detection. This system can detect 10% of DAC for tritium levels and is also sensitive to 41Ar. “Cold trap” sampling is also used to sample for tritium. These cold trap samples are analyzed using liquid scintillation.

Continuous air monitors are available for airborne particulate and iodine monitoring on an as-needed basis. One continuous air monitor is typically positioned in the spent fuel storage pool area. Filter and charcoal cartridge samplers are also available for iodine and particulate sampling. These filter and cartridge samples are analyzed in the radioanalysis laboratory.

11.1.4.3 Effluent Monitors

41Ar effluent at NBSR is monitored with a G-M detector located in the stack. This system is calibrated by comparison to a grab sample that is analyzed in the radioanalysis laboratory.

The NBSR tritium effluent out the stack is continuously monitored by the building tritium monitoring system. Monthly grab samples from the stack are also collected and analyzed for verification purposes. More frequent sampling or additional continuous monitoring is implemented when unusual or non-routine activities involving the potential for added tritium release are performed. Effluent sampling can also be performed with a particulate filter and charcoal cartridge, and analyzed on an as-required basis.

11.1.4.4 Environmental Monitors

Thermoluminescent dosimeters (TLDs) are used for environmental ambient gamma monitoring. Also used for selected monitoring situations are:

· A pressurized tissue-equivalent ion chamber system.

· Environmental G-M monitors with data logging

· A gain stabilized NaI system for monitoring 41Ar or other specific gamma emitters.

11.1.4.5 Personnel Dosimeters

Personal radiation dosimeters, for both gamma and neutron dose measurement capabilities are provided by a NVLAP certified supplier. Occupational doses can also be determined by pocket ion chambers (PICs) or electronic dosimeters, or by area radiation surveys combined with stay-times. Extremity dosimeters, such as finger TLDs, wrist TLD badges, and wrist PICs are also used when needed.

11.1.5 Radiation Exposure Control and Dosimetry

This section describes how radiation exposure is controlled within the facility and how uncontrolled radioactivity is prevented from entering work areas or the environment.