RADIOLOGICAL CONTROLS

Transcription

1 OMEGA LLE RADIOLOGICAL CONTROLS MANUAL S-AA-M-16

2 S-AA-M-16 OMEGA LLE RADIOLOGICAL CONTROLS MANUAL

3 15 August 2011 LLE INSTRUCTION 6610F SUBJECT: LLE Radiological Controls Manual 1. Purpose: To promulgate the procedures for radiological controls in the Laboratory for Laser Energetics. 2. Promulgation: The attached LLE Radiological Controls Manual is hereby promulgated. 3. Approval: Walter T. Shmayda LLE Radiation Safety Officer

4 LLE NOTICE August 2011 LLE NOTICE 1000 SUBJECT: LLE INSTRUCTION 6610F 1. Purpose: To promulgate LLE Radiological Controls Manual (LLE INST 6610F). 2. Directions: Remove Pages All-6610C All-6610D All-6610E Replacement Pages All-6610D All-6610E All-6610F 3. Complete Record of Change page i by entering 6610F, Entered by, and Date of Entry columns. 4. Approval: Walter T. Shmayda LLE Radiation Safety Officer

5 RADCONMAN 15 August 2011 Record of Changes and List of Effective Pages Record of Changes Change/ Date of Change/ Date of Rev. No. Entered by Entry Rev. No. Entered by Entry E 5/9/08 F W. Shmayda 8/15/11 i

6 RADCONMAN 15 August 2011 List of Effective Pages Page No. Change/ Rev. No. Date Page No. Change/ Rev.No. Date i Rev. F 15 August 2011 II-25 Rev. F 15 August 2011 ii Rev. F 15 August 2011 II-26 Rev. F 15 August 2011 iii Rev. F 15 August 2011 III-1 Rev. F 15 August 2011 iv Rev. F 15 August 2011 III-2 Rev. F 15 August 2011 v Rev. F 15 August 2011 III-3 Rev. F 15 August 2011 vi Rev. F 15 August 2011 III-4 Rev. F 15 August 2011 I-1 Rev. F 15 August 2011 III-5 Rev. F 15 August 2011 I-2 Rev. F 15 August 2011 III-6 Rev. F 15 August 2011 I-3 Rev. F 15 August 2011 III-7 Rev. F 15 August 2011 I-4 Rev. F 15 August 2011 III-8 Rev. F 15 August 2011 I-5 Rev. F 15 August 2011 III-9 Rev. F 15 August 2011 I-6 Rev. F 15 August 2011 III-10 Rev. F 15 August 2011 I-7 Rev. F 15 August 2011 III-11 Rev. F 15 August 2011 I-8 Rev. F 15 August 2011 IV-1 Rev. F 15 August 2011 I-9 Rev. F 15 August 2011 IV-2 Rev. F 15 August 2011 I-10 Rev. F 15 August 2011 IV-3 Rev. F 15 August 2011 I-11 Rev. F 15 August 2011 IV-4 Rev. F 15 August 2011 II-1 Rev. F 15 August 2011 IV-5 Rev. F 15 August 2011 II-2 Rev. F 15 August 2011 IV-6 Rev. F 15 August 2011 II-3 Rev. F 15 August 2011 V-1 Rev. F 15 August 2011 II-4 Rev. F 15 August 2011 V-2 Rev. F 15 August 2011 II-5 Rev. F 15 August 2011 V-3 Rev. F 15 August 2011 II-6 Rev. F 15 August 2011 App Map A-1-i Rev. F 15 August 2011 II-7 Rev. F 15 August 2011 App Map A-1-ii Rev. F 15 August 2011 II-8 Rev. F 15 August 2011 App Map A-2a Rev. F 15 August 2011 II-9 Rev. F 15 August 2011 App Map A-2b Rev. F 15 August 2011 II-10 Rev. F 15 August 2011 App Map A-2c Rev. F 15 August 2011 II-11 Rev. F 15 August 2011 App Map A-3a Rev. F 15 August 2011 II-12 Rev. F 15 August 2011 App Map A-3b Rev. F 15 August 2011 II-13 Rev. F 15 August 2011 App Map A-3c Rev. F 15 August 2011 II-14 Rev. F 15 August 2011 App Map A-3d Rev. F 15 August 2011 II-15 Rev. F 15 August 2011 App Map A-4a Rev. F 15 August 2011 II-16 Rev. F 15 August 2011 App Map A-4b Rev. F 15 August 2011 II-17 Rev. F 15 August 2011 App Map A-4c Rev. F 15 August 2011 II-18 Rev. F 15 August 2011 App Map A-4d Rev. F 15 August 2011 II-19 Rev. F 15 August 2011 App Map A-4e Rev. F 15 August 2011 II-20 Rev. F 15 August 2011 App Map A-4f Rev. F 15 August 2011 II-21 Rev. F 15 August 2011 App Map A-4g Rev. F 15 August 2011 II-22 Rev. F 15 August 2011 App Map A-4h Rev. F 15 August 2011 II-23 Rev. F 15 August 2011 App Map A-4i Rev. F 15 August 2011 II-24 Rev. F 15 August 2011 App Map A-4j Rev. F 15 August 2011 ii

7 RADCONMAN 15 August 2011 List of Effective Pages Page No. Change/Rev. No. Date Page No. App Map A-4k Rev. F 15 August 2011 App Map A-4l Rev. F 15 August 2011 App Map A-4m Rev. F 15 August 2011 App Map A-4n Rev. F 15 August 2011 App Map A-4o Rev. F 15 August 2011 App Map A-4p Rev. F 15 August 2011 App Survey Rev. F 15 August 2011 Log A-1 App Survey Rev. F 15 August 2011 Log A-2 App Survey Rev. F 15 August 2011 Log A-3 App Survey Rev. F 15 August 2011 Log A-4 App Log A-4 Rev. F 15 August 2011 App Log A-5 Rev. F 15 August 2011 App Log A-6 Rev. F 15 August 2011 App Log A-7 Rev. F 15 August 2011 App Form A-1 Rev. F 15 August 2011 App Form A-2 Rev. F 15 August 2011 App Form A-3 Rev. F 15 August 2011 App B-i Rev. F 15 August 2011 App B-ii Rev. F 15 August 2011 App B-iii Rev. F 15 August 2011 App-B-iv Rev. F 15 August 2011 App-B-v Rev. F 15 August 2011 App-B-vi Rev. F 15 August 2011 App-C-i Rev. F 15 August 2011 App-C-ii Rev. F 15 August 2011 App-C-iii Rev. F 15 August 2011 Change/Rev. No. Date iii

8 Table of Contents Part I: Radiological Control Fundamentals 1000 Introduction... I OMEGA Neutron Radiation... I OMEGA EP Radiation... I Neutron Activation Gamma Radiation... I Tritium Beta Radiation... I Surface Contamination... I Airborne Radiation... I Biological Effects of Radiation... I Radiation Protection Principles... I Summary of Radiological Limits... I-10 Part II: Radiation Protection Systems 2000 OMEGA Shielding System... II OMEGA EP Shielding System... II Tritium Fill Station... II Tritium Transport... II Target Area Tritium Removal Systems... II Environmental Monitoring System... II Radiation Detection Instruments... II-22 Part III: Requirements and Procedures 3000 Shielding Effectiveness Monitoring... III Target Chamber Activation Surveys... III Target Bay General Radiation Surveys... III Airborne Radiation Surveys... III Surface Contamination Surveys... III Liquid Activity Surveys... III Anticontamination Clothing... III Establishing a Controlled Surface Contamination Area... III Target Chamber Entry... III Decontamination Procedures... III Internal Transfer of Tritium Targets... III Radioactive Material Accountability and Disposal... III LLE Monthly Tritium Inventory... III Scintillation Counting... III Personnel Monitoring... III-10 iv

9 3015 Material with Fixed Activity... III Estrablishing a Radioactive Materials Control Area... III-11 Part IV: Emergency Procedures 4000 Spill of Radioactive Material... IV High Airborne Activity... IV Acute Release of Tritium... IV UR Security Acute Release of Tritium Emergency Procedure (from LLE Tritium Fill Station)... IV-6 Part V: Maintenance Procedures 5000 Tritium Fill Station Radiological Requirements... V Experimental Operation Radiological Requirements... V-2 APPENDICES Appendix A: Survey Maps, Logs, and Forms Survey Map A-1 Radiation Shielding Monitoring Points Survey Map A-2a Target Bay Gamma Radiation Target Bay Ground Level Survey Map A-2b Target Bay Gamma Radiation TMS Platform Level 1 Survey Map A-2c Target Bay Gamma Radiation TMS Platform Level 2 Survey Map A-3a OMEGA EP Target Bay Gamma Radiation TAS Platform Level 1 Survey Map A-3b OMEGA EP Target Bay Gamma Radiation TAS Platform Level 2 Survey Map A-3c OMEGA EP Target Bay Gamma Radiation TAS Platform Level 3 Survey Map A-3d OMEGA EP Target Bay Gamma Radiation Laser Bay Floor Survey Map A-4a Surface Contamination Survey Log Survey Map A-4b Surface Contamination Survey Log (LaCave) Survey Map A-4c Surface Contamination Survey Log (LSC Area) Survey Map A-4d Surface Contamination Survey Log (Micro Assembly Lab, Rm. 2828) Survey Map A-4e Surface Contamination Survey Log (Tritium Facility) Survey Map A-4f Surface Contamination Survey Log (OMEGA Target Bay Ground Level) Survey Map A-4g Surface Contamination Survey Log (OMEGA Target Bay TMS Platform Level 1) Survey Map A-4h Surface Contamination Survey Log (OMEGA Target Bay TMS Platform Level 2) Survey Map A-4i Surface Contamination Survey Log (Target Positioner 2) Survey Map A-4j Surface Contamination Survey Log (LaCave Darkroom) Survey Map A-4k Surface Contamination Survey Log (LaCave) v

10 Appendix B: Definitions Survey Map A-4l Surface Contamination Survey Log (Decontamination Area Room 136) Survey Map A-4m Surface Contamination Survey Log (Cart Maintenance Room Room 150A) Survey Map A-4n Surface Contamination Survey Log (Pump House/TC-TRS Room 150B) Survey Map A-4o Surface Contamination Survey Log (Room 157 Hood) Survey Map A-4p Surface Contamination Survey Log (TC-TRS Condensate Transfer) Survey Log A-1 Surface Contamination Survey Log for Clean Items Survey Log A-2 Surface Contamination Survey Log Cryo and Tritium Facility Water Condensate Survey Log A-3 Liquid Activity Survey Log Survey Log A-4 Airborne Radiation Survey Log Log A-4 Target Chamber Personnel and Material Entry Log Log A-5 Radioactive Material Control Log Log A-6 Weekly Radiological Survey Cover Sheet Target Production Log A-7 Weekly Radiological Survey Cover Sheet Experimental Operations Group Form A-1 Target Chamber Access Authorization Form A-2 LLE TFS, CTHS, OMEGA Training Inventory Form A-3 Warm Tritium Target Inventory Appendix C: Acronyms & Abbreviations vi

11 Part I: Radiological Control Fundamentals Part I Radiological Control Fundamentals 1000 Introduction 1001 OMEGA Neutron Radiation 1002 OMEGA EP Radiation 1003 Neutron Activation Gamma Radiation 1004 Tritium Beta Radiation 1005 Surface Contamination 1006 Airborne Radiation 1007 Biological Effects of Radiation 1008 Radiation Protection Principles 1009 Summary of Radiological Limits 1000 Introduction Radiological controls include all aspects of ensuring the protection of personnel and the environment from the hazards associated with ionizing radiation and radioactive materials. As a U.S. Nuclear Regulatory Commission agreement state, New York State has accepted the responsibility to establish and enforce regulations governing the handling of radioactive material and radiation-producing devices. The NYS Department of Health governs the licensing of users of radioactive material and the regulations that are applicable to these users, while the NYS Department of Environmental Conservation governs the protection of the environment and general public at large. At the University of Rochester, a Radiation Safety Committee is responsible for establishing the procedures that implement the NYS requirements. These procedures are contained in a UR Radiation Safety Manual. The UR Radiation Safety Unit is responsible for enforcing compliance with these procedures by the users (Principal Investigators) of radioactive material and radiation-producing devices at the University. The Laboratory for Laser Energetics (LLE) is both a user of radioactive material (primarily tritium) and has a device, the OMEGA Laser System, that produces radiation (neutrons and x rays directly and beta-gamma radiation indirectly). Tritium stored in the LLE Tritium Fill Station is used to fill polymer capsules for inertial confinement fusion (ICF) implosion experiments. Fusion experiments yield high-energy prompt neutron radiation as a result of the fusion process, prompt gamma radiation from neutron capture, and delayed gamma radiation as a result of neutron activation of high-z structural material. This manual describes the radiation protection systems in the OMEGA Laser Facility and the radiological control procedures used at LLE to ensure personnel and environmental protection from sources of ionizing radiation. In effect, this manual describes how the requirements of the University of Rochester Radiation Safety Manual are fulfilled at LLE. In the event of a conflict between this manual and the UR Radiation Safety Manual, the UR Radiation Safety Manual takes precedence. See Appendix B for definitions of radiation safety terms. 15 August 2011 Page I-1

12 Part I: Radiological Control Fundamentals 1001 OMEGA Neutron Radiation The fusion of deuterium-tritium (DT) and pure deuterium (DD) produces energetic charged particles, x rays, and neutrons. The basic reactions are as follows: D 2 D 2 He " 2 + 0n 1 ] MeVg 50% D 2 D " 1T 3 + 1P 1 ] MeVg 50% T 3 1D 2 " 2He n 1 ] 14 MeVg 100% While the x rays and charged particles are easily shielded by the Target Chamber, the neutrons with energies up to 14 MeV are penetrating enough to escape the Target Chamber. Since the neutrons are emitted promptly when the fusion reaction occurs, they are a hazard only at the time of a target shot. However, because of the high numbers of neutrons produced, they are the most significant radiation hazard associated with the operation of the OMEGA Laser System. The fluence per unit dose equivalent for monoenergetic neutrons is as follows: Neutron Energy (MeV) Fluence per unit dose equivalent (neutrons/cm 2 /rem) # # # # # # 10 7 Considering that the Target Chamber OD is inches (330 cm), a target shot that produces the maximum credible yield of 3 # neutrons would result in a dose of 516 rem at the surface of the Target Chamber (the lethal dose for 100% of the population is considered to be 900 rem). Neutrons emanate from the target radially; thus, the neutron flux decreases due to spherical spreading (as a function of r 2 from Target Chamber Center). The concrete shielding around the Target Bay protects personnel external to the Target Bay from neutron radiation. Because of the beam holes in the shield wall, the shield does not protect personnel in the Laser Bay. Rather, excluding access to the Laser Bay during shots offers the necessary protection for personnel within LLE, and distance from the radiation source protects anyone external to the west end of the Laser Bay. Protection of personnel adjacent to the Laser Bay, i.e., in the Amplifier Test and Assembly areas, is offered by a combination of distance and the oblique angle in the straight line path from the target center to these areas. For a more detailed discussion of shielding see Section August 2011 Page I-2

13 Part I: Radiological Control Fundamentals Neutrons interact with matter by either scattering or absorption. In scattering, the neutron s kinetic energy is reduced by either elastic or inelastic collisions. The most effective energy transfer is by elastic collisions (as with a billiard ball that imparts all of its energy to another billiard ball when it hits it head on). Thus, the best shielding is low-z materials, in particular hydrogenous material since the mass of a hydrogen nucleus equals that of a neutron. The effectiveness of the concrete shield is primarily due to the included hydrated water. In inelastic scattering, only part of a neutron s kinetic energy is transferred to another atom, and the neutron scatters in another direction with slightly less energy (as when a billiard ball hits a bowling ball a glancing blow). In absorption, a neutron is absorbed by an atom s nucleus. The nucleus in turn becomes excited and releases its excess energy by either the prompt or delayed emission of gamma, beta, proton, or other radiation OMEGA EP Radiation The interaction of the high-intensity OMEGA EP laser beams with planar high-z targets produces fast electrons e that, in turn, produce hard x rays via bremsstrahlung (braking radiation). While the spreading of the lower-energy x rays, kt = 0.1 MeV, is uniformly spherical over a solid angle of 4r, the spreading of higher-energy x rays is directional. X rays of energy 1 MeV and 10 MeV are in approximately 90 and 45 cones, respectively, along the axis of the incoming laser beam (Fig. I-1). OMEGA EP was designed to direct the short-pulse backlighter beams north and the sidelighting beam west to minimize the amount of shielding required. Very energetic protons Target e, p g, e, p Ions, debris G6507 g, e Fig. I-1 The laser accelerates the electrons, and the charge separation sets the ions in motion. 15 August 2011 Page I-3

14 Part I: Radiological Control Fundamentals are also produced as a result of the interaction. Protons are produced from residual water on or in the target as the electrons are stripped away. As the electrons leave the target, the protons follow normal to the target. The protons are effectively stopped by the Target Chamber. Neutrons are produced from a D-D fusion reaction (Section 1001) if deuterium targets are employed. These neutrons have an energy of 2.45 MeV and spread spherically. An additional source of high-energy neutrons is a (cn) reaction; however, this source is two to three orders of magnitude less than the neutrons from the D-D reaction. The cn reaction also produces radioactive activation products within the target chamber that decay with varying half-lives and are a source of c radation. This source must be considered when entering the OMEGA EP target chamber. Gamma radiation is produced from the decay of neutron activation products (Section 1003). This source is significantly less than that on OMEGA due to the lower neutron flux on OMEGA EP. From a radiation control perspective, the neutron radiation dominates the general area shielding and the hard x rays dominate the directional shielding. Radiation protection is provided by a combination of shielding and the use of closed access to the OMEGA and OMEGA EP facilities depending on the type of experiment. For a more detailed discussion of the OMEGA EP shielding see Section Neutron Activation Gamma Radiation The neutron radiation produced from fusion over time can activate the structural material in the Target Bay. The activated isotopes in turn decay and emit gamma radiation. The primary contribution to the activation radiation comes from the activation of the aluminum alloy (Al) Target Chamber. The decay of two nuclides formed by this activation dominates the radiation from the chamber. Na-24 (half-life = 15 h) emits most of the radiation shortly after a target shot, whereas Mn-54 (half-life = 313 days) contributes most of the long-term radiation. Of a secondary concern is activation of the iron in the Target Bay structures. The nuclide Fe-59 (half-life = 45 days) would be the primary activation product in the unlikely event that structure activation is observed. The primary activation and decay reactions are as follows: n 1 Na 23 Na 24* Mg " 11 " 12 + b] 1. 4 MeV + c] MeV + c] MeV g g g n 1 Mn 53 Mn 54* Fe " 25 " 26 + b] MeVg + c] MeVg. Figures I-2 and I-3 show the specific activation levels of the OMEGA Target Chamber by a 3 # neutron-yield shot after ten years with 12 shots per year at this level. With these activation levels the radiation level will be reduced to <1 mrem/h at the surface of the Target Chamber or <0.01 mrem/h at a distance of 5 m from the Target Chamber in less than two days after such a shot occurs. The integrated neutron yield over time will be used in conjunction with periodic 15 August 2011 Page I-4

15 Part I: Radiological Control Fundamentals Ci/cc G Time (h) Time (days) Fig. I-2 Maximum design target chamber specific activity. Fig. I-3 Maximum design target chamber specific activity. radiation surveys to determine if neutron activation will require personnel monitoring and stay times to restrict personnel exposure. Since gamma rays are not charged particles, interactions with matter are not a result of electrostatic forces as with alpha and beta interactions. Rather, gamma or x rays transfer energy by direct interaction with atoms. There are three types of interactions: Photoelectric effect The photon transfers all of its energy to an orbital electron, causing its ejection. This effect is most probable for low-energy photons and high-z materials. Compton scattering The photon transfers part of its energy in ejecting an electron. The remainder of the energy is emitted as a secondary photon, which might scatter in any direction. Pair production The photon (with a minimum energy of 1.02 MeV) interacts with the nucleus of an atom to produce an electron (b ) and a positron (b + ) pair and a photon of lower energy Tritium Beta Radiation Tritium is used as a fusion fuel in ICF target experiments. As an isotope of hydrogen it behaves chemically and physically like hydrogen. In either the gaseous (HT) or vapor (HTO) form, continuous isotopic exchange with the H in materials, plants, and animals occurs. Additionally, because of the small size of the hydrogen molecule, it easily penetrates the grain boundaries and interstices of materials. It is these properties of tritium that make its use and handling more difficult than other radioactive materials; e.g., tritium in either the HT or HTO form can enter the body through either the mouth, nose, or skin, and, because it is a gas and not a particulate, it cannot be filtered. Additionally, because tritium is absorbed in materials, it out-gases as an inverse function of the vapor pressure and undergoes isotopic exchange as a direct function of vapor pressure. Thus, because of out-gassing over time, a material that has been exposed to tritium is very difficult to certify as being decontaminated. See Section 3007 for a discussion 15 August 2011 Page I-5

16 Part I: Radiological Control Fundamentals of decontamination. Figure I-4 shows the outgassing rate of tritium from a surface as a function of the surface activity level. Tritium decays with a half-life of 12.3 years as follows: T 3 He 3 1 " 2 + b - _ kev maximum, 5. 7 kev averagei. Up to 15,000 Ci of tritium is stored in the uranium getters of the Tritium Fill Station (TFS). The tritium is used to fill warm CH capsules in the TFS or cryogenic capsules in the CTHS (Cryogenic Target Handling System). (See Volume IV CTHS Description for more information on the TFS and CTHS.) These targets are transported and inserted into the Target Chamber in a manner that ensures against release to the environment. After a tritium target is imploded, the residual tritium is absorbed into the Target Chamber and diagnostics and is subsequently removed via the Target Chamber vacuum systems. Most of the residual tritium is collected in the cryogenic vacuum pumps connected to the Target Chamber. When these pumps are regenerated, the tritium is collected in a scavenger bed composed of a zirconium-iron getter. Tritium is also released from the Target Chamber and diagnostics when they are evacuated by the main Target Chamber roughing pumps and the diagnostic roughing and turbo-backing vacuum pumps, respectively. These streams are processed by a tritium removal system prior to being discharged. Separate monolayer/h (each particle labelled with a T atom) 10 3 Outgassing rate (nci/h/100 cm 2 ) Note: outgassing at room temperature 10 1 Outgas rate (nci/h/100 cm 2 ) = 7 # 10 6 Surface activity 0.74 (DPM/100 cm 2 ) Surface activity (DPM/100 cm 2 ) G7137J1 Fig. I-4 Tritium outgassing rate as a function of surface activity. 15 August 2011 Page I-6

17 Part I: Radiological Control Fundamentals ventilation exhaust stacks from the TFS and CTHS room (Room 157) and the target vacuum systems are monitored to measure the integral amount of tritium released to the environment. As a charged particle, beta radiation causes ionizations as well as interactions as it passes through matter. Due to its smaller mass and lower charge, it is not as densely ionizing as an alpha particle, thus it may travel slightly farther through matter (e.g., an energetic beta can travel a few meters through air, but a tritium beta can travel only a few millimeters through air). The negatively charged beta particle can also interact with the positively charged nucleus causing the release of x rays as the beta deflects (bremsstrahlung radiation). The probability of bremsstrahlung radiation is proportional to an atom s Z number. Therefore, effective shielding that minimizes bremsstrahlung radiation includes low-z materials, such as plastic, wood, aluminum, etc. Low-energy beta radiation is easily shielded by a piece of paper; therefore it is only an internal radiation hazard. In the case of tritium, the beta particle is of such low energy it cannot be detected on film badges or thermoluminescent dosimeters (TLD s) used to monitor personnel exposure. Accordingly, film badges or TLD s are not required when working with or around tritium. Rather, urinalysis testing of personnel who work with >100 mci of tritium is required within one week of exposure; or, for continuous exposure, weekly testing is required Surface Contamination Surface contamination is loose radioactive material that can be removed from surfaces and may therefore be easily spread. Surface contamination is normally only an internal radiological hazard. Thus, personnel protection can be provided by ensuring that such contamination does not enter the body. The method of protection depends upon whether or not the surface contamination is filterable. Tritium in the form of free HT or HTO is not filterable. Thus, protection from tritium surface contamination is afforded by wearing rubber or plastic covering over parts of the body that are likely to contact tritiated components (normally, rubber gloves when handling tritiated components and full anticontaminiation clothing, including rubber gloves, filter masks, and shoe coverings, when entering the Target Chamber). Where the tritium is contained in target debris, such as inside the Target Chamber, it is filterable. The only other possible source of surface contamination at LLE is from the ablation of activated Target Chamber or diagnostic material. This is not expected to be a concern at LLE; however, any precautions to protect personnel from tritium will likewise protect them from this source of particulate material. Surface contamination is detected by taking wipe survey samples as described in Section Airborne Radiation Airborne radiation may be in the form of gaseous or particulate radioactive material. There are two possible sources of airborne radiation at LLE: tritium and neutron activation of air. Tritium comes from the TFS and tritium targets while neutron activation of air is given by the following reactions: 15 August 2011 Page I-7

18 Part I: Radiological Control Fundamentals * 0n 1 + Ar " Ar hr > K b^2. 49 max MeVh + c^1. 29 MeV * 0n 1 + N 15 7 " N sec > O b] max MeVg + c^6. 13 MeV * 0n 1 + O 16 8 " N p sec > O b] max MeVg + c^6. 13 MeV h h h For a maximum credible yield DT shot of 3 # neutrons, the maximum instantaneous airborne activity as a function of distance from the center of the Target Chamber is given as follows: Distance (cm) Activity (nci/m 3 ) Since both Ar 41 and N 16 are airborne gases that give off gamma radiation, they are primarily external radiation hazards. The applicable limit for submersion in a hemispherical semi-infinite cloud of Ar 41 is 3 nci/m 3. Since the average activity in a finite -size target bay is less than this limit and recognizing the short radiological half-life and the even shorter effective half-life considering ventilation, any exposure from Ar 41 is insignificant. The half-life of N 16 is so short that it is effectively gone within a minute after a shot and is therefore not a radiological hazard. All normal tritium-handling operations are conducted in a manner to ensure that atmospheric levels are below the applicable airborne limits. To ensure compliance with environmental limits, the two exhaust stacks that could possibly contain tritium are monitored by Environmental Protection Agency (EPA) compliance monitors Biological Effects of Radiation Radiation can effect living tissue by mutating, killing, or otherwise damaging cells. There is no substantiated threshold for genetic effects; however, it is known that the magnitude of these effects increases with exposure. The radio sensitivity of cells is directly proportional to their reproductive rate and inversely proportional to their degree of differentiation. Thus, blood-forming organs, digestive organs, reproductive organs, and embryos are more highly radiosensitive than nerve cells, muscle tissue, and the vascular system. The following are the effects of acute radiation doses: 15 August 2011 Page I-8

19 Part I: Radiological Control Fundamentals Dose (rads) Effect <5 No detectable clinical effects 5 25 Some changes noted in chromosomes, no physical effects Vomiting in 10% of exposed population 450 Lethal dose for 50% of the exposed population after 30 days Hematopoetic Syndrome: death in 10 to 15 days after exposure. Changes in lymphocytes causes increased susceptibility to infection ,000 Gastrointestinal Syndrome: most die in 3 to 5 days after exposure. The gastrointestinal tract loses its absorptive functions. Symptoms are nausea, vomiting, salivation, dehydration, and weight loss. >10,000 Central Nervous System Syndrome: death occurs in 1 to 2 days after exposure. Symptoms are hyperexcitability and incoordination in minutes followed by cardiovascular collapse. >100,000 Molecular Death: death occurs almost immediately due to inactivation of basic metabolic processes Radiation Protection Principles The aim of radiation protection is to reduce all radiation exposures to as low as reasonably achievable (ALARA). This is done by applying time, distance, and/or shielding. Time Minimize the exposure time to radiation sources to reduce the dose equivalent since Dose Equivalent = Dose Rate # Time. Since OMEGA is a pulsed source, the number of shots is equivalent to time. Distance Maximize the distance between a radiation source and the exposed person. This causes the dose rate to decrease by spreading. For point sources, spherical spreading applies and the dose rate decreases by a factor of 1/r 2 : 2 D1 D2 = `r2 r1j For OMEGA, closed access ensures that personnel are at a safe distance. Shielding Decrease the radiation dose rate by inserting an attenuating/absorbing material between the radiation source and the exposed person. The effectiveness of a shielding material depends upon the type and energy of the radiation. For example: Alpha particles easily shielded by a piece of paper or a few centimeters of air; thus they present no external radiation hazard. Beta particles easily shielded by 1/4 in. or less of a low-z material such as aluminum, plastic (Lucite or Plexiglass), or wood. Small thicknesses of a high-z material should be avoided since more-penetrating bremsstrahlung radiation (x rays) will be produced. 15 August 2011 Page I-9

20 Part I: Radiological Control Fundamentals Neutrons and gamma and x rays Since these types of radiation are more penetrating, their shielding is usually handled more rigorously. R = R 0 e nx where R = radiation rate n = linear absorption coefficient (a function of material and type and energy of radiation) x = shielding material thickness Often, the coefficient is not used directly; rather, the half-thickness (L 1/2 ) of the material is used, where n = ln 2/L 1/2. To shield for gamma or x rays, a high-z material such as lead is most effective, whereas for neutrons, a low-z hydrogenous material such as water, oil, or cement is most effective. Some typical shielding half-thicknesses are as follows: Type Radiation Energy/Nuclide Half-thickness gamma or x ray 1.27 MeV/Na cm of Pb gamma or x ray 0.66 MeV/Cs cm of Pb gamma or x ray 0.36 MeV/I cm of Pb neutrons 10 MeV 7.64 cm of H 2 O 1009 Summary of Radiological Limits A summary of the environmental and personnel protection radiological limits of concern to LLE is presented in this section. This summary is based on the environmental limits contained in NYS Department of Environmental Conservation publication 6 NYCRR Part 380 (Rules and Regulations for Prevention and Control of Environmental Pollution by Radioactive Materials) and US EPA compliance standards contained in the Code of Federal Regulations, 40 CFR part 61, subpart I, and the personnel protection limits contained in the NYS Department of Health Sanitary Code, Title 10, Chapter I, part 16 (Ionizing Radiation) that are implemented at the University of Rochester in the UR Radiation Safety Manual. For a detailed listing and explanation of these limits, these references should be consulted. 15 August 2011 Page I-10

21 Part I: Radiological Control Fundamentals Tritium Limits Surface contamination (based on controlling radioactive material) Airborne limit in radiological work areas for HTO (applicable to personnel qualified as Radiation Worker, based on exposure for 2000 hr/yr and a total dose equivalent of 5 rem/yr) Airborne limit for general population for HTO (applicable to general population based on continuous exposure and a total dose equivalant of 50 mrem/yr to allow exposure to all age groups) NYS DEC non-permit stack limit (10% of airborne limit for general population) Drinking water limit (based on a total dose equivalant of 50 mrem/yr) Water sewage disposal limit Annual limit on intake (ALI) (exposure of 5 rem) 1000 dpm/100 cm 2 20 nci/m nci/m nci/m 3 1 # 10 3 nci/ml 22,000 dpm/ml 80 mci General External Radiation Limits Occupational Radiation Worker Annual Limits Total effective dose equivalant to whole body Total effective dose equivalant to embryo/fetus Total effective dose equivalant to minors (<18 yr) or Sum of deep dose equivalent and committed dose equivalent to any individual organ or tissue other than the lens of the eye 1 Eye dose equivalent 1 Shallow dose equivalent to skin 1 Extremity dose equivalent 1 Non Radiation Worker Annual Limit Total effective dose equivalent Radiation Area (film badge monitoring and control required) High Radiation Area (no entry allowed) Very High Radiation Area 1 Limits to minors (<18 years old) are 10% of these limits. 5 rem/yr 500 mrem 500 mr/yr 50 rem/yr 15 rem/yr 50 rem/yr 50 rem/yr 100 mrem/yr >5 mrem/h at 30 cm from source >100 mrem/h at 30 cm from source >500 rem/h at 1 m from source 15 August 2011 Page I-11

22 Part II: Radiation Protection Systems Part II Radiation Protection Systems 2000 OMEGA Shielding System 2001 OMEGA EP Shielding System 2002 Tritium Fill Station 2003 Tritium Transport System 2004 Target Area Tritium Removal System 2005 Environmental Monitoring System 2006 Radiation Detection Instruments 2000 OMEGA Shielding System A combination of shielding, distance, and shot cycles/yields is employed to limit radiation exposures in all accessible areas to <25% of the general population radiation exposure limit of 100 mrem/yr. The neutron shielding is conservatively designed to limit exposures on the basis of ten maximum credible yield shots of 3 # (14.1-MeV) neutrons per year. This maximum credible yield is used to provide conservative dose calculations and shield design since it represents the absolute maximum conceivable yield. In actuality the average expected neutron yields are from to and the maximum credible yield of 3 # neutrons may be approached only with cryogenic DT targets. The radiation shielding design is discussed in more detail in the Environmental Assessment for the OMEGA Upgrade Project. The radiation levels detailed in this section are based on three sources as modified by the actual shielding installed. Report Number 6, Radiation and Health Safety of December 16, 1975, by United Engineering and Constructors, Inc., documents the design criteria of the original OMEGA laser system. For the upgrade, computer simulations of the radiation levels were performed by the Corporate Research Center of the Grumman Corporation under contract (#US3411) with the U of R and by the Inertial Fusion Applications Group (Y division) at LLNL. The Grumman analysis identified the locations with the highest radiation levels, and the LLNL group performed specific calculations for these locations. Sky shine was not explicitly calculated but is estimated not to add more than 10% to the calculated radiation levels in accessible areas of concern. The neutron shield shown in Fig. II-1(a) and described below consists of the concrete floor of the Target Bay and a light-weight concrete wall constructed around the perimeter of the Target Bay. Floor: The concrete floor is 30 in. thick and is penetrated by a 7-ft-diam hole immediately below the Target Chamber and by several ventilation and service holes. This floor provides at least 30 in. of shielding in the path from the Target Chamber center to any area outside LaCave. Thus, all areas below the Target Bay and Laser Bay floors that are outside LaCave including the anteroom and darkroom of the north of LaCave are protected. Because of the penetrations in the Target Bay floor, access to LaCave except for the anteroom and the darkroom is precluded during any target shot that is expected to produce neutron yields > August 2011 Page II-1

23 Part II: Radiation Protection Systems Control Room ESO Fig. II-1(a) The neutron shield. Curtain shield Target chamber center Curtain shield G6509 Looking south Emergency fire exit G6510 Fig. II-1(b) South Wall: The south wall is 30 in. thick and extends to a height above the Target Chamber. It is penetrated by an emergency fire exit door on the east end. A 30-in. curtain shield constructed of large concrete blocks provides at least 30 in. of shielding between the Target Chamber center and any area of the door. 15 August 2011 Page II-2

24 Part II: Radiation Protection Systems Target Bay shield looking east OMEGA EP beam hole Target Bay door filled in G6511 Fig. II-1(c) East Wall: The east wall is 30 in. thick and extends to a height above the Target Chamber. The south portion is higher because it is part of the original building construction. The wall was extended during the upgrade building project completed in 1993 and is slightly lower to prevent placing additional unnecessary weight on the Target Bay foundation. This wall is penetrated by the large Target Bay door that was used during the construction period. After construction this door was filled in with blocks filled with concrete to provide 30 in. of shielding. A 4-ft # 4-ft hole was cut in the shield to provide a path for the OMEGA EP beams to OMEGA. Due to the shield penetration, closed access of the opposing facility is required under some circumstances (see Section 2001). Looking north Normal access LDL beam hole G6512 Fig. II-1(d) North Wall: The north wall is nominally 35.4 in. thick and extends to a height above the Target Chamber. To provide for the full shielding effectiveness for ten maximum credible yield shots, the addition of 1.82 in. of borated polyethylene may be required if yields and shot operations prove this to be necessary. This borated polyethylene was not added during upgrade building modifications in 1993 for cost considerations. If yields and the monitoring of shielding effectiveness dictate that this additional shielding is required, it will be added at a later date by attaching the borated polyethylene to the outside of the shield wall in the 15 August 2011 Page II-3

25 Part II: Radiation Protection Systems areas considered necessary. Because of space restrictions in the area of the Cluster 1 F-ASP and the stairwell leading to LaCave, the shielding thickness is staggered from 35.4 in. on the east end to 33 in. on the west end such that 35.4 in. of shielding is maintained on a line from the Target Chamber center. This wall is penetrated by two openings: the normal entrance doorway and a beam hole that allows entrance of a beam from the Laser Development Laboratory. Each of these openings is provided with a curtain shield of concrete blocks to provide shielding between the Target Chamber center and any accessible external area. Looking west Target chamber center G6513 Fig. II-1(e) West Wall: The west wall is 30 in. thick and extends nominally to a height above the Target Chamber. The wall is penetrated by 67 openings: sixty 18-in.-diam holes for the laser beams to pass from the Laser Bay to the Target Bay; six 3-in.-diam holes at the base of the wall for service lines, and the door opening to the Target Bay Viewing Gallery. Because the bottom of the door opening for the Viewing Gallery is above the Target Chamber center, protection is provided to all accessible areas outside the west end of the Laser Bay. The straight-line path from the Target Chamber center through the shielding in the vicinity of the 3-in. service openings is in excess of 30 in.; therefore protection is provided by the shielding. The impact of the sixty 18-in.-diam holes was the subject of a special study conducted by LLE and the Inertial Fusion Applications Group (Y division) at LLNL. Since the beam holes are to the north and south of the centerline of the west wall, protection to the west is provided by the shield wall. Protection is provided to the south of the Laser Bay by the fact that the Laser Bay floor is nominally 14 ft above ground level and accordingly access is restricted. Additionally, because of the angular path from the Target Chamber center through these holes, substantial (although less than 30 in.) shielding is provided. Thus, by a combination of shielding, distance, and accessibility, the area to the south meets the shielding design criteria. Since the Amplifier Test and Assembly areas are north of the 15 August 2011 Page II-4

26 Part II: Radiation Protection Systems Laser Bay and accessible during shot operations, these areas require special consideration. Figure II-1(f) details the geometry of the line-of-sight path of neutrons from the Target Chamber center to these areas and specifies the areas for which the radiation levels in Table II-1 were calculated. Since these areas do not meet the shielding design criteria as maximum credible neutron yields are approached, access to these areas may have to be restricted. The shielding effectiveness will be monitored in these areas as a function of neutron yields to determine if and when these areas must be included as closed access areas during shot operations. Control Room Gowning Room Zone 11 Zone 10 Zone 9 Zone 8 Zone 7 Zone 6 Zone 5 Zone 4 Zone 3 Zone 2 Zone 1 Laser Bay N Target Bay G6514 Fig. II-1(f) Laser Bay north zones. The expected radiation levels for a maximum credible yield target shot of 3 # neutrons are detailed in Table II-1 for areas outside the perimeter of the shield. These levels may be extrapolated linearly for yields less than the maximum credible. These calculations take into consideration the shielding of the 4-in.-thick aluminum Target Chamber, the thickness of the concrete shield, and the distance from the Target Chamber center to the listed locations. 15 August 2011 Page II-5

27 Part II: Radiation Protection Systems Location Table II-1: Calculated Radiation Levels Neutron Dose/ Shot of 3 # Neutrons Gamma Dose/ Shot of 3 # Neutrons Total Dose/ Shot of 3 # Neutrons Total Dose/ Ten Shots of 3 # Neutrons South of Target Bay 3 ft above ground level 1.54 mrem 0.17 mrem 1.71 mrem 17.1 mrem East of Target Bay 3 ft above ground level 1.54 mrem 0.17 mrem 1.71 mrem 17.1 mrem East of Target Bay at target level * mrem mrem mrem mrem Control Room wall as built without borated polyethylene Control Room at ESO station without borated polyethylene Control Room wall with borated polyethylene (not existing) Control Room at ESO station with borated polyethylene (not existing) Laser Bay east-west centerline at wall (no access allowed) mrem 0.30 mrem mrem mrem 6.80 mrem 0.20 mrem 7.00 mrem 70.0 mrem 2.50 mrem 0.30 mrem 2.80 mrem 28.0 mrem 1.75 mrem 0.20 mrem 1.95 mrem 19.5 mrem mrem 0.90 mrem mrem mrem Laser Bay west wall 0.91 mrem 0.05 mrem 0.96 mrem 9.6 mrem Laser Bay Zone 1 (Anteroom) mrem ~1.00 mrem mrem mrem Laser Bay Zone 2 (Control Conference Room) mrem ~1.50 mrem mrem mrem Laser Bay Zone 3 (Rod Amplifier Room) mrem ~2.00 mrem mrem mrem Laser Bay Zone 4 (Rod Amplifier Room) mrem ~1.20 mrem mrem mrem Laser Bay Zone 5 (SSA Cleaning Room) mrem ~1.90 mrem mrem mrem Laser Bay Zone 6 (SSA Assembly Room) mrem ~0.90 mrem mrem mrem * excluding the effect of the 12-in. OMEGA EP west shield wall. 15 August 2011 Page II-6

28 Part II: Radiation Protection Systems 2001 OMEGA EP Shielding System Like the OMEGA radiation protection system, OMEGA EP uses a combination of shielding, distance, and shot types/numbers to limit radiation exposure in all accessible areas to <25% of the general population radiation exposure limit of 100 mrem/yr. The shielding design is based on 100 shots per year that produce either the maximum x-ray and/or maximum neutron yields indicated below: X-ray maximum source function Spectrum I = I 0 e ho/kt with kt = 0.1 MeV, 1 MeV, 10 MeV Energy/shot 1000 J of x rays Solid angle 4r for kt = 0.1 MeV 90 cone for kt = 1 MeV 45 cone for kt = 10 MeV Neutron maximum source function Spectrum E = 2.45 MeV Number n < Solid angle 4r The shielding shown in Fig. II-2 and described below consists of a normal-density concrete floor and a shield wall that surrounds the entire OMEGA EP bay to a height of 6.6 ft above the target chamber. The shielding design is based on a study entitled Radiation Safety and Effects Analysis for OMEGA EP prepared by Advanced Energy Systems, Inc., for the University of Rochester. Floor: The concrete floor is 30 in. thick and is penetrated by a 3-ft-diam hole immediately below the Target Chamber and by several ventilation and service holes. This floor provides at least 30 in. of shielding in the path from the Target Chamber Center to any area outside the experimental diagnostic bays located below the Target Chamber area. This ensures that all areas below the floor and outside the experimental diagnostic areas remain at <25% of the general population s yearly limit. Because of the penetrations between the target area and the experimental diagnostic area, this diagnostic area must be in closed access during any target shot. North Shield Wall: Neutrons and high-energy x rays (1 and 10 MeV) control the design of the north shield wall. This shield wall is made of normal-density concrete that is 39.4 in. thick except for a 4-in.-thick, 20-ft # 20-ft cutout centered on Target Chamber Center. This cutout covers the 45 cone north of the Target Chamber that is expected to contain the directional 10 MeV x rays. This cutout allows the addition of 4 in. of lead, which is required for the shield to meet the <25 mrem/yr design criteria for 10-MeV x rays. The shield is 21 ft, 4 in. high, which is approximately 6.6 ft above Target Chamber Center. With 4 in. of lead inserted in the cutout, the shield meets its design criteria for all sources of radiation except for 1-MeV x rays. Since this radiation subtends a 90 cone north of Target Chamber Center, additional lead shielding varying from 0 to 2.75 in. thick outside the cutout area to a distance 41 ft on either side of the shield-target chamber center line would have to be installed to meet the design criteria. The necessity for this shielding will be determined during the activation program and either the shielding will be installed or the number of high x-ray 15 August 2011 Page II-7

29 Part II: Radiation Protection Systems 82 ft 39.4 in. Beam transport tube from OMEGA 21 ft 20.3 ft TC 4-in. indentation in shield to allow installation of 4-in. c lead (if necessary) 56 ft Both the OMEGA and OMEGA EP shields have a 4-ft 4-ft beam tube opening in. NOTES: 1. Shield material is normaldensity concrete. Formed except where 12-in. filled concrete blocks are shown. 2. Shield is 21 ft, 4 in. high 19 in. 16 in. 16 ft 16 ft 265 ft 12-in. filled concrete blocks 20 ft 12 in. N Curtain shield (this plus 2 others to south, not shown) Dry wall 12-in. filled concrete blocks G6515aJ1 Fig. II-2 OMEGA EP shield. 15 August 2011 Page II-8

30 Part II: Radiation Protection Systems yield shots will be restricted to remain within the radiation design criteria of accessible areas remaining <25 mrem per year. East Shield Wall: Neutrons control the design of the east shield wall. This shield wall is poured concrete for the first 108 ft from the north and has a tapered thickness from north to south: 31.5 in. for 56 ft, then 19 in. for 16 ft, then 16 in. for 16 ft, and then 12 in. for 20 ft. Thereafter, the shield is formed by 12-in.-thick concrete blocks filled with concrete. Where doorways enter the bays, curtain shields offset the entrance doors to provide shielding in the line of sight from the target chamber. The shield is 21 ft, 4 in. high, which is approximately 6.6 ft above Target Chamber Center. South Shield Wall: Neutrons control the design of the south shield. This shield is formed by 12-in.-thick concrete blocks filled with concrete. The shield is 21 ft, 4 in. high, which is approximately 6.6 ft above target chamber center. West Shield Wall: Neutrons control the design of the west shield under normal circumstances. If one short-pulse beam to sidelight a target is used, the shield performance is limited by 1- and 10-MeV x rays. To accommodate this, either OMEGA would have to be in closed access or additional lead shielding would be required. The shield, constructed as part of OMEGA EP, is formed by 12-in.-thick concrete blocks filled with concrete. The shield, in combination with the 30-in. OMEGA shield, provides 42 in. of shielding between OMEGA and OMEGA EP and vice versa. To allow the OMEGA EP beams to pass to the OMEGA target chamber, a 4-ft # 4-ft hole penetrates both the OMEGA east shield and the OMEGA EP west shield wall. Accordingly, for independent shots on OMEGA or OMEGA EP, closed access in designated areas in the opposite facility is required under the following circumstances: For OMEGA shots: Closed access on top of GCC, OMEGA EP target area structure (TAS) upper deck, GCC interior, and EP target chamber if >3 # neutrons on OMEGA (implemented as OMEGA Type 7b shots) Closed access to the entire EP Laser Bay if >3 # neutrons on OMEGA (implented as OMEGA Type 7c shots) For OMEGA EP shots: OMEGA access not restricted for UV shots, OMEGA EP neutron yield shots up to 1 # and OMEGA EP short-pulse backlighter shots with energies up to 2.1 kj on target OMEGA target bay must be in closed access for an OMEGA EP shot to the OMEGA target chamber OMEGA target bay must be in closed access for OMEGA EP sidelighter shots 2002 Tritium Fill Station As originally constructed the Tritium Fill Station (TFS) consisted of the receiving and inventory loop, pumping system, DT storage loop, capsule charging loop, piping and valving, glovebox, ancillary systems, glovebox purification system, electrical and instrumentation systems, 15 August 2011 Page II-9

31 Part II: Radiation Protection Systems glovebox atmosphere monitors, control system, and the supporting facility. In 1995 the TFS was modified as follows: addition of a glovebox annex to allow for the handling and coating of filled targets, modification of the process loop to allow for the addition of cryogenic filling capabilities, replacement of the contact tube to allow filling of mounted targets, and installation of a cryogenic cooler. The original system is fully described in the Tritium Filling Station Environmental Assessment and the original Tritium Filling Station Design Manual, and the 1995 modifications are included on the updated as-built drawings. In 1998 the facilities containing the TFS were expanded to accommodate the cryogenic target handling system (CTHS), which was installed in For further information on this system see the CTHS Volume IV System Description (S-AA-M-31), Chap. 5. In 2003, the TFS was moved to a new addition to Room 157 and a new glovebox (connected to the previously added annex) was added for tritium experiments. Additional upgrades to the TFS included modification and simplification of the glovebox cleanup system; upgrading the programmable logical controller (PLC) used to control the process; the addition of a new graphical user interface employing a touch panel; the removal of the Overhoff tritium sampling system; and the addition of new moisture analyzers, a flowmeter, and a pressure sensor in the cleanup loop. In 2004 new instrumentation and valves were added to the TFS process equipment. In addition, a new double-contained process line interconnecting the TFS with the DT High-Pressure System (DTHPS) was completed. This line will serve as the means to transfer tritium between the TFS and the DTHPS. In 2005 a new computer was added to the control system, enabling the operation via a new graphical user interface, which includes a dynamic process schematic. Overview of Room-Temperature Fill The maximum permissible tritium inventory at LLE is 1.5 grams of tritium (15,000 Ci). This tritium is reversibly stored in one of two uranium beds. At room temperature, the U-beds absorb deuterium/tritium (by converting DT to uranium hydride), and the tritium is released when the U-bed is heated to 425 C. When the system is not in use, the entire tritium inventory is stored in the U-beds in the form of uranium hydride (each of the two beds has a 26,400-Ci capacity for a 50/50 D/T mixture). Under normal operation the TFS can complete one loading cycle every three days. To improve the efficiency of DT transfer, it is condensed in the condensation cell to minimize back pressure in the U-bed. The DT is transferred from the U-bed to the condensing cell and then to the assay volume. The assay volume is a 7.4-liter volume that has two pressure gauges and temperature gauges that are used to assay the contents. A portion of the DT is then recondensed in the condensation cell. The DT that is not condensed is returned to the second U-bed (referred to as the cold bed). The condensation cell temperature is then raised to evaporate the DT and pressurize the contact cell where the targets reside. After the fill has been completed, the balance of the DT is transferred to the cold U-bed. The residual DT left in the process line is removed by circulating the residual DT over the cold U-bed using the Normetex pump. The system is then evacuated into a singlepass, dedicated clean-up train (process loop getters) before being expelled into the glovebox cleaning getters. The actual tritium content of each polymer capsule will be variable dependent 15 August 2011 Page II-10

32 Part II: Radiation Protection Systems upon the experimental program. A reference value of 12.5 mci is assumed for gaseous targets, and a reference value of 0.3 Ci is assumed for cryogenic targets. Overview of Cryogenic, High-Pressure Fill The TFS supplies DT for the cryogenic, high-pressure fills. The DT is first transferred to the condensation cell by heating the U-bed and condensing the DT in the condensation cell (as discussed in the previous section). The pressure in the system peaks at 1 atm during condensation but reaches low pressure (50 torr) at the end of the condensation cycle because nearly all of the DT gas is condensed. This allows all of the DT to be driven off the hot U-bed in one heating. The DT is then expanded to the assay volume, and an assay is taken. Subsequently, the DT is transferred to the DTHPS (by condensing in the DTHPS condensation tube). The DTHPS will return portions of the inventory during the high-pressure fill cycle. Each time DT is returned, the DT is assayed and then transferred to the cold U-bed. The residual DT is then circulated over the cold U-bed and the TFS is evacuated. The high-pressure fill cycle is described in more detail in the CTHS Volume IV System Description. Under normal operating conditions, the TFS was designed to achieve practically zero leakage of tritium from system components located in the glovebox. The glovebox serves as a second barrier to prevent tritium leakage into the atmosphere. Any tritium that enters the glovebox will be collected in the glovebox purification system. Therefore, by design the contribution of the TFS to the chronic ambient tritium concentration is negligible. As a third barrier to people within the UR/LLE building, the room housing the TFS is kept at a negative pressure to ensure zero leakage to adjacent spaces. Tritium levels are monitored continuously inside the glovebox by a tritium monitor integral to the TFS. If this monitor detects excessive tritium, the TFS is automatically shutdown and the tritium is returned to the uranium beds. To ensure safety, accidents that could affect either the general environment and population or individual UR/LLE employees were analyzed. The analysis concluded that there was no accident that would put either the general population or environment at risk. An accident that releases the entire inventory of tritium to the TFS room and thence to the environment is the worst possible accident. The TFS Radiological Safety Assessment evaluates the probability of this accident to be very remote since it would require dual failures (i.e., failure of the TFS system resulting in release to the glovebox and the simultaneous failure of the glovebox). This accident s analyses and the consequences of an acute release of 15,000 Ci of tritium to the TFS room and thence to the environment are detailed below. To calculate the potential radiation dose, the following assumptions were made: Instantaneous release of 15,000 Ci of elemental tritium to the TFS room while a UR/LLE staff member was inside. * *The 1-h assumption is based on a ventilation half-life of 11.5 min. Considering that it takes approximately five half-lives for complete elimination, it would take approximately 1 h to exhaust all of the tritium from the TFS room to the environment. 15 August 2011 Page II-11

33 Part II: Radiation Protection Systems Following the failure, it would take less than 1 min for the tritium alarm to annunciate and and no more than 2 min for the UR/LLE staff member to evacuate the room. Negligible oxidation of tritium occurs within the glovebox or the TFS room. Based on the above assumptions, the effective whole body dose equivalent resulting from a 2-min inhalation exposure to elemental tritium is less than 6 mrem to the UR/LLE staff member. To calculate the potential dose to the general public, the following assumptions were made: Release of 15,000 Ci of elemental tritium to the environment via the TFS room exhaust over a period of 1 h. Adverse weather conditions: summer, no rain, stable conditions (Pasquill Stability Class F) with wind speed of 2 m/s, and a low capping inversion height. The individual exposed remains downwind on the plume centerline for a period of 7 days. Exposure is from elemental tritium. Based on the above assumptions, the radiation doses from exposure to elemental tritium and conclusions are as follows: Maximum individual doses (200 m downwind) Average general population (adult): 12 nrem Average general population (child): 10 nrem Dose at 2 km falls below 2 nrem. Collective dose is unlikely to exceed 1 man-rem. To ensure that the TFS is operated safely, all operators must be qualified as Radiation Safety Workers and system operators in accordance with the LFORM. To protect against the inadvertent release of tritium to either the building or the environment, the facility must be locked when no one is in attendance, and all operations must be performed strictly in accordance with the CTHS Volume V operating procedures Tritium Transport Transport of Warm Tritum Targets: The tritium targets are transported in dedicated transport bottles from the Tritium Fill Station to the Target Positioner (TP) following procedure SS-5-4, Part 4 (OMEGA System Operations Manual, Vol. II). After a target is placed in the transport bottle from within the TFS glovebox, the bottle is swiped (and decontaminated if necessary), and then transported to the TP. When the bottle is uncapped at the TP, it is held at arms length and an Overhoff portable tritium monitor is used to detect tritium. Additional exhaust ventilation at the TP and the handling procedures outlined in procedure SS-5-4, Part 4 (OMEGA System Operations Manual, Vol. II) ensure that exposure to the operator is as low as reasonably achievable. Weekly urinalysis testing confirms that exposures are insignificant. At all times during this operation, the target, target support, and target transfer containers are under the direct and continuous supervision of a Radiation Worker qualified technician. 15 August 2011 Page II-12

34 Part II: Radiation Protection Systems Transport of Cryogenic Tritium Targets: Cryogenic tritium targets are transported in a Moving Cryostat Transport Cart in full containment. See CSO-9 for cryogenic target transfer from the FTS to an MCTC and SO-7 for transfer from the MCTC to the OMEGA Target Chamber Target Area Tritium Removal Systems Tritium is introduced into the Target Chamber from the implosion of targets containing tritium. Since most of the unburned tritium will collect on the cryogenic vacuum pumps, the pumps are exhausted to a Tritium Scrubber System. A small fraction of the tritium will make its way to the Turbo Backing and Auxiliary Roughing vacuum pumps during normal operations and to the Main Roughing vacuum pumps during Target Chamber pump-down from atmospheric pressure. To support cryogenic DT operations, the removal of tritium from the Main Roughing, Turbo Backing, Auxiliary Roughing and Lower Pylon pump exhaust streams, and the MCTC Maintenance Room 150A is accomplished by a TC-TRS located in the Pump House Room 150B. The integral target chamber vacuum and tritium removal system is shown in Fig. II-3. The tritium removal system is composed of the following major components (exhaust monitoring systems are discussed in Section 2004). See the OMEGA System Description, Volume I, Chap. 12 for a detailed description of the TC-TRS. Tritium Scrubber: The Tritium Scrubber System collects tritium in its various forms when the cryogenic pumps are regenerated. The Tritium Scrubber consists of a molecular sieve bed to collect water, an input ionization chamber, a pre-heater chamber, a nickel bed to crack hydrocarbons and chemically absorb oxygen, a zirconium iron alloy bed to collect HT, and an output ionization chamber. These components are housed in an integral unit located in LaCave. Based on regenerating the cryogenic pumps on a monthly basis, the beds are sized to receive 12 regenerations and still effectively reduce the tritium levels to below environmental limits. Data from the ion chambers is recorded and analyzed to determine the capture efficiency and to determine when the beds will be regenerated or replaced. The exhaust from the tritrium scrubber is directed to the TC-TRS to ensure environmental exhaust limits are met during regeneration. Target Chamber Tritium Recovery System (TC-TRS): The TC-TRS removes trace levels of tritium from air purge streams and vacuum system exhaust streams associated with the OMEGA TC and the Cart Maintenance Room (Room 150A). Figure II-3 illustrates how the TC-TRS connects to the existing OMEGA system. Figure II-4 shows a process flow diagram for the TC- TRS. The trace levels of tritium are first catalytically oxidized into tritiated water; then the water vapor is subsequently removed from the air stream by adsorption on molecular sieve beds. The gas streams are processed in a continuous fashion. For a detailed description of the TC-TRS see the CTHS System Description, Chap. 12. The system components are described below. The TC-TRS is composed of two skids: the Reactor Skid and the Molecular Sieve Skid. The two main functions of the Reactor Skid are to control flow into the TC-TRS and to catalytically oxidize tritium. Figure II-4 shows a process flow diagram for the TC-TRS. The TC-TRS draws gas from the OMEGA system through an intake manifold into a regenerative side-channel blower. The TC-TRS automatically compensates for different levels of flow into the manifold. This is achieved by a system employing a recycle valve that controls the gas pressure 15 August 2011 Page II-13

35 Part II: Radiation Protection Systems Fig. II-3 Present system. Target positioner Diagnostics Target chamber Purge bypass Purge bypass Purge bypass Aux. rough pump Turbo backing pump Main roughing pump Lowerpylon roughing pump TRS header Exhaust stack G6516J1 T Cryopumps Tritium scrubber T Cart Maintenance Room TC-TRS Low-pressure receiver T Recycle valve T Catalytic reactor 60-CFM flow (Bypass to stack) T Atmospheric low-level tritium Vacuum low-level tritium flow Atmospheric high-level tritium flow Vacuum high-level tritium flow Decontaminated flow Ion chamber tritium detector HTO recovery tank Tritium air sampler H 2 O (liq.) T Mole sieve bed Exhaust stack Hot N 2 15 August 2011 Page II-14

36 Part II: Radiation Protection Systems Exhaust stack Gas from OMEGA and cryopump scrubber, and cart maintenance room T Flow Blower Heat Reactor Temp Lowpressure receiver Reactor heat exchanger Temp Preheater Temp Chilled water supply/return FV-8500 Recycle valve Heat Aftercooler Reactor skid T Temp H 2 O Condenser Condensate recovery Mole sieve skid Chilled water supply/return Level Jacketed condensate receiver T Flow Chilled water supply/return H 2 O Press Heat Mole sieve bed Temp Heat Mole sieve bed Temp Heat Mole sieve bed Temp Temp Preheater Heat Flow Nitrogen Exhaust stack G6517J1 Fig. II-4 Target Chamber Tritium Recovery System. 15 August 2011 Page II-15

37 Part II: Radiation Protection Systems in the manifold. The blower continuously operates at a constant flow rate of 60 scfm. When there is no gas load on the system, 100% of the gas that leaves the blower is recycled back via the recycle valve. The recycle loop is comprised of the blower, the reactor, a regenerative heat exchanger, a preheater, an aftercooler, and a large tank the Low-Pressure Receiver (LPR). The blower takes suction from the LPR as well as from the intake manifold. Gas can flow either back to the LPR or onto the Molecular Sieve Skid. When large loads of gas are pumped into the intake manifold, recycle is choked and all the gas that exits the blower is passed to the rest of the system. The pressure in the intake manifold is monitored and used to control the amount of recycle flow. The reactor converts elemental tritium and trace tritiated organic compounds into HTO and CO 2. The reactor employs 300 lb of a palladium-based catalyst (a recombiner ) operating at 400 C. The reactor maintains this temperature very closely since there is a constant flow of gas through it, regardless of the flow into the intake manifold of the TC-TRS. Once gas leaves the Reactor Skid, it flows to the Molecular Sieve Skid where molecular sieve beds (300 lb each) adsorb the HTO and water. Three beds are employed in such a manner that while one is operating as a scrubber, a second can be regenerated. The third bed is used to polish the effluent from the regeneration process. During regeneration, hot nitrogen gas flows through the bed to be regenerated, releasing water and saturating the exhaust stream. The regeneration process employs a watercooled condenser to remove approximately 85% of the water from the saturated stream. The system automatically measures the rate of recovery of condensate. The condensed water is sampled for tritium: if it is below the limit for disposal 10 pci/l (1 # 10 2 nci/ml), it is disposed via the sewer; if >10 2 nci/ml but <0.5 nci/ml, it is disposed of as radiation waste, and if >0.5 nci/ml, it is sent to a waste processor for recovery. Room 157 Tritium Removal System (TRS): This TRS removes tritium from helium and air gas streams that originate from the OMEGA Cryogenic Target Handling System (CTHS). The TRS was designed and built to support the use of tritium in the CTHS. For a detailed description of the Room 157 TRS see the CTHS System Description, Chap. 12. The main objectives of the TRS are as follows: Remove sufficient tritium from the exhaust streams of the CTHS vacuum pumps so that exhaust concentrations do not exceed 1mCi/m 3 and that the total activity of tritium released is kept under 1 Ci per year. Remove sufficient tritium from the glovebox atmospheres of the CTHS equipment so that background levels are maintained at or below 1mCi/m 3. Maintain low levels of tritium in the DTHPS secondary containment spaces. Provide high leak integrity and safe operation to limit operator exposure to tritium. 15 August 2011 Page II-16

38 Part II: Radiation Protection Systems Table II-2 summarizes the CTHS gas streams that enter the TRS for processing. There are two general types of feed streams: glovebox atmospheres and vacuum exhaust from CTHS vacuum pumps. Table II-2: Summary of gas streams entering the TRS. Gas Stream Gas Source Gas-Generating Operation Cleaning TRS Subsystem Vacuum Pump Exhaust High DT DTHPS FTS Target filling Target transfer to MCTC Tritium Cleanup Subsystem Low DT MCTC Transport of targets between equipment Air Clean helium FTS target passthrough FTS interspace MCTC Characterization FTS base FTS dome FTS cooling module Inserting/removing target racks Connecting MCTC with FTS Pumping down MCTC from air Pumping down characterization chamber from air Continuous Continuous Continuous Glovebox Cleanup DTHPS glovebox DTHPS glovebox Continuous scrubbing of glovebox atmosphere FTS glovebox FTS glovebox Continuous scrubbing of glovebox atmosphere TFS glovebox TFS glovebox Continuous scrubbing of glovebox atmosphere Other DTHPS secondary DTHPS Continuous cleaning/monitoring of containment high-pressure equipment Tritium Cleanup Subsystem Air Vacuum Cleanup Subsystem Stack to Exhaust or Tritium Cleanup Subsystem Glovebox Cleanup Subsystem Glovebox Cleanup Subsystem Glovebox Cleanup Subsystem DTHPS Secondary Containment Cleanup Subsystem 15 August 2011 Page II-17

39 Part II: Radiation Protection Systems Exhaust gases from vacuum pumps originate from the CTHS system. A vacuum pump system helps to maintain the CTHS at cryogenic conditions by maintaining necessary vacuum. The effluent from the vacuum pumps is sent to the TRS for removal of tritium (detritiation) and other species including water and tritiated organics. In addition, there are several gloveboxes that surround and contain the CTHS equipment. The gas in these glovebox atmospheres become tritiated as a result of the tritium operations occurring within. Therefore, the gas in these gloveboxes needs to be constantly detritiated via the TRS. In general, the TRS will process two types of gas: air and helium. Helium is used as the inert gas in the CTHS gloveboxes. Helium is also used as a process gas during CTHS operations. Additionally, there are operations where CTHS equipment needs to be opened to air and then pumped down again to vacuum. The effluent from this evacuation must also be treated. Therefore, both air and helium need to be accommodated by the TRS, which significantly affects the methods used for tritium removal. As a result, different processing techniques are used depending on whether the carrier gas is air or helium. Three specific types of vacuum exhaust streams originate from the CTHS: Normally clean helium (from equipment that is normally not exposed to tritium) Air (from equipment pump-outs) A combined stream of highly tritiated helium ( high D-T ) and moderately tritiated helium ( low D-T ) from equipment used to process tritium targets. Another gas stream that the TRS processes is the containment volumes located within the DTHPS glovebox. The DTHPS system handles pure DT under high pressure and as a result has an additional containment around key equipment to contain any breach that might occur without release to the glovebox. Helium is used as the inert gas in these containment volumes. Tritium leakage is expected from the equipment in these volumes. Accordingly, these spaces must be continuously monitored and processed by the TRS to minimize tritium migration into the glovebox atmosphere. There are four distinct cleanup subsystems in the TRS: Air Vac Cleanup Subsystem (ACU) Tritium Cleanup Subsystem (TCU) Glovebox Cleanup Subsystem (GCU) DTHPS Secondary Containment Cleanup Subsystem(SECCON) The ACU cleans up air streams originating from the CTHS Air Vac Manifold. The TCU cleans gas originating from the CTHS High and Low Tritium Vac manifold. The GCU handles all glovebox gas cleanup. The SECCON is a separate, dedicated cleanup subsystem that is physically located outside of the TRS and inside the DTHPS glovebox. 15 August 2011 Page II-18

40 Part II: Radiation Protection Systems Figure II-5 is an overview of these various subsystems and the gas streams that they process. TRS Feed Streams Overview of Room 157 TRS TRS enclosure CTHS Vacuum Pump Exhaust CTHS Gloveboxes Normally tritiated helium volumes Normally clean helium volumes Air Volumes Fill and transfer station (FTS) glovebox Tritium fill station (TFS) glovebox DT High-Pressure fill (DTHPS) glovebox Y? N Tritium? N Y? Tritium? Tritium cleanup subsystem Tritium HTO Air vacuum cleanup subsystem HTO Tritium Glovebox cleanup subsystem Getter HTO Clean gas Clean gas Clean gas Clean gas Condensate for removal Clean gas TRS enclosure Exhaust header To stack exhaust Discard to sanitary drain or ship out as rad waste Return to gloveboxes G6518 DTHPS secondary containment DTHPS sec. contain. cleanup subsystem Clean gas Return to DTHPS secondary containment Fig. II-5 Graphical overview of the TRS Environmental Monitoring System Small amounts of tritium are exhausted to the environment via two separate exhaust stacks that rise above the Target Bay roof. One exhaust stack, located on the north side of the Target Bay, receives the exhaust from the Tritium Fill Station and the Room 157 Tritium Removal System at 2200 scfm. The other exhaust stack, located on the south side of the Target Bay, receives the exhaust from the Target Chamber vacuum systems and/or the Target Chamber Tritrium Removal System at 1500 scfm. Both stacks are continuously monitored to verify that the DEC discharge permit limit is not exceeded. Both stacks are sampled by an Overhoff Technology Corporation (OTC) EPA tritium-in-air sampler. Samples from this system are taken on a batch basis and read by a liquid scintillation counter. 15 August 2011 Page II-19

41 Part II: Radiation Protection Systems Overhoff Tritium Air Sampler: This sampler collects both elemental tritium and tritium oxide samples from the exhaust stack air stream to verify compliance with the NYS DEC discharge permit. A pump controlled by an adjustable constant flow controller draws a representative sample from the exhaust line and passes it through two sets of 50-ml vials containing liquid scintillation fluid separated by a high-efficiency (typically 99%) catalytic oxidizer. The oxide form of tritium is collected by the first two vials in the process stream, and elemental tritium is collected by the second two vials after it is converted to the oxide form by the catalytic oxidizer. Each vial has a >95% collection efficiency. The vials are located on the front panel of the sampler for ready detachment and measureing the fluid activity by liquid scintillation counting. The system is calibrated to determine the percentage of tritium collected by the vials to allow accurate measurements of tritium. Typically, a collection time of 8 h will yield detection thresholds to <1 nci/m 3 (10 9 nci/ml). Increasing the collection time can result in even lower detection thresholds. G6519 Overhoff Tritium Air Sampler (EPA Compliance Monitor) 15 August 2011 Page II-20

42 Part II: Radiation Protection Systems Femto-TECH Tritium Monitor U24-D: This system is composed of an ion chamber for detection and a U24-D Control Unit for system control and display functions. This system measures tritium activity directly using an 1800-cc, active-volume ion chamber capable of resolving tritium to 0.1 nci/m 3 with an accuracy of!5%. Flow is provided by a 1-scfm diaphragm pump with flow measured by a rotameter. The measurement system is compensated up to 200 nci/m 3 for uniform gamma backgrounds using an alignment on the front panel. The system has high and low alarms set at 10 and 20 nci/m 3 per cubic meter, respectively. G6520 Femto-TECH Tritium Monitor U24-D 15 August 2011 Page II-21

43 Part II: Radiation Protection Systems 2006 Radiation Detection Instruments Overhoff Model 700 BAcC Portable Tritium Monitor: This monitor is used to conduct routine surveys upon opening potentially contaminated systems to ensure that airborne levels remain below the limits applicable to radiation workers. The instrument is a small, high-sensitivity, handheld, battery-operated, gamma-compensated airborne survey meter that can detect tritium in its elemental or oxide form. It uses four identical ionization chambers, two for tritium measurement and two for gamma compensation. A sample is drawn through the ionization chambers by means of a small rotary vane pump. An analog meter is used for measurement display using switch-selectable ranges of 0 to 10, 100, 1000, or 10,000 nci/m 3. The instrument exhibits a basic sensitivity of 1 nci/m 3, which allows detection below the 20 nci/m 3 limit applicable to radiation workers. Power is supplied by two 9-V batteries. The onset of battery depletion is signaled by the illumination of an LED located next to the meter face. An alarmlevel setting potentiometer adjustable over the full scale is located on the front panel. A steady tone is emitted by an acoustic signaler if the alarm point is exceeded. An intermittent tone is heard if the sample air flow has been interrupted. A more detailed description of this instrument is contained in the Operating Instructions provided by the manufacturer. G6521 Overhoff Model 700 BAcC Portable Tritium Monitor 15 August 2011 Page II-22

44 Part II: Radiation Protection Systems Scintrix Limited Model 209 L Portable Tritium-in-Air Monitor: This monintor is used as a backup to the Overhoff Portable Tritium Monitor and, if necessary, during an emergency condition. Because it is not as sensitive as the Overhoff meter, it should not normally be used for routine operations. The instrumentation is a small, handheld, battery-operated, gamma-compensated airborne survey meter that can detect tritium in its elemental or oxide form. It uses two identical ionization chambers, one for tritium measurement and one for gamma compensation. A sample is drawn through the ionization chambers by means of a small rotary vane pump. A digital LCD readout is used for measurement display over a range of 0 to 20,000 nci/m 3 with a resolution of 10 nci/m 3. Power is supplied by five C-cell batteries. The onset of battery depletion is signaled by a blinking decimal point on the display. An electrometer output is used to trigger an alarm. The alarm may be set at either 100, 500, or 1000 nci/m 3 ; it is set at the factory to alarm at 500 nci/m 3. A more detailed description of this instrument is contained in the Technical Manual provided by the manufacturer. G6522 Scintrex Limited Model 209 L Portable Tritium-in-Air Monitor 15 August 2011 Page II-23

45 Part II: Radiation Protection Systems Perkin Elmer Tri-Carb 2910TR Liquid Scintillation Counter: This instrument is used to assay tritium swipe and liquid samples. The 2910TR is a semiautomatic, self-calibrating, liquid scintillation counter that assays racks of 7-ml sample vials. The 2910TR includes a detector unit, lead shielding, counting electronics, keyboard and display, and output printer and is fitted with a 19 nci Ba 133 external source that is used to gather Compton scattering data from each sample for quench compensation. Different samples types are identified automatically by barcode labels on the sample rack. Sample rack slot #1 will contain a background vial filled with 4 ml of a liquid scintillation cocktail and a clean swipe from the package used for the survey. The sample rack final sample slot will contain a cocktail-only background sample filled with only 4 ml of liquid scintillaion cocktail. Quench correction, background subtraction, and specific calibration curves are employed along with digital spectrum analysis techniques to directly determine the DPM value for the sample. A detailed description of this instrument is contained in the Instrument Manual provided by the manufacturer. G9368J1 Perkin Elmer Tri-Carb 2910 TR Liquid Scintillation Counter 15 August 2011 Page II-24

46 Part II: Radiation Protection Systems Wallac Model 1409 Liquid Scintillation Counter: This instrument is used to assay tritium swipe and liquid samples. The 1409 is a semiautomatic, self-calibrating, liquid scintillation counter that assays racks of 20-ml sample vials. The 1409 includes a detector unit, lead shielding, counting electronics, keyboard and display, and output printer and is fitted with a 12 nci Eu 152 external source that is used to gather Compton scattering data from each sample for quench compensation. Different samples types are identified automatically by barcode labels on the sample rack. Sample rack slot #1 will contain a background vial filled with 15 ml of a liquid scintillation cocktail and a clean swipe from the package used for the survey. The sample rack final sample slot will contain a cocktail-only background sample filled with only 15 ml of liquid scintillaion cocktail. Quench correction, background subtraction, and specific calibration curves are employed along with digital spectrum analysis techniques to directly determine the DPM value for the sample. Caution: Liquid activities that exceel 500,000 CPM will saturate the counter and must be diluted and recounted in order to obtain accurate results. A detailed description of this instrument is contained in the Instrument Manual provided by the manufacturer. G6524 Wallac Model 1409 Liquid Scintillation Counter 15 August 2011 Page II-25

47 Part II: Radiation Protection Systems Ludlum Model 3 Portable Survey Meter with a Model 44-9 Detector: This instrument is used to perform periodic and contact radiation surveys in the Target Bay and Target Chamber to measure the radiation levels caused by neutron activation of structural materials. The Model 3 is a portable GM radiation survey instrument with four linear ranges used in combination with dose rate (mr/h) or CPM meter dials. It is powered by two standard D-cell batteries that are tested when the selector switch is positioned to the battery test position. A speaker mounted on the instrument gives an audible indication of the relative dose rate being measured. A more detailed description of this instrument and its operation is contained in the Instruction Manual provided by the manufacturer. Note: This model is calibrated for CPM. G6525 Ludlum Model 3 Portable Survey Meter with a Model 44-9 Detector 15 August 2011 Page II-26

48 Part II: Radiation Protection Systems Ludlum Model 9 Ion Chamber: This instrument is used to perform area c surveys in the OMEGA Target Bay to measure the radiation levels caused by neutron activation of structural materials. The Ludlum Model 9 is a portable ion chamber instrument calibrated for exposure rates in the 0 to 5000 mrem/h range. Is is powered by two standard D-cell batteries. Warm-up time is 2 min until stabilized. Calibration change is less than 5% battery dependent when batteries are wihtin battery-check limits on the meter. Response time is 3 to 5 sec, depending on the range selected. A more detailed description of this instrument and its operation is contained in the instruction manual provided by the manufacturer. Note: This model is calibrated for dose in mrem/h. Bottom view of Model 9 Ludlum Model 9 Ionization Chamber 15 August 2011 Page II-27

49 Part III: Requirements and Procedures Part III Requirements and Procedures 3000 Shielding Effectiveness Monitoring 3001 Target Chamber Activation Surveys 3002 Target Bay General Radiation Surveys 3002 Airborne Radiation Surveys 3003 Airborne Radiation Surveys 3004 Surface Contamination Surveys 3005 Liquid Activity Surveys 3006 Anticontamination Clothing 3007 Establishing Anticontamination Controlled Area 3008 Target Chamber Entry 3009 Decontamination Procedures 3010 Internal Transfer of Tritium Targets 3011 Radioactive Material Accountability and Disposal 3012 LLE Monthly Tritium Inventory 3013 Scintillation Counting 3014 Personnel Monitoring 3015 Material with Fixed Activity 3000 Shielding Effectiveness Monitoring To monitor that the OMEGA and OMEGA EP Target Bay shielding is performing to design specifications and to ensure that exposure limits applicable to the general population are not being exceeded, the radiation dose on the outside of the shield is monitored by using thermoluminescent dosimeters (TLD) with CR39 obtained from the University s Radiation Safety Unit. Specific monitored sites are detailed on Survey Map A-1 in Appendix A. These sites were selected to monitor the shield in general and the penetrations detailed in Sections 2000 and 2001, in particular, to ensure shielding effectiveness in limiting exposure to less than allowable limits. The TLD s with CR39 are exchanged for new ones quarterly or whenever the integrated neutron yield exceeds neutrons since the last exchange, whichever occurs earlier. The TLD s are delivered to the Radiation Safety Unit for delivery to Global Dosimetry Solutions for processing. The dosimetry report is reviewed by the LLE Radiation Safety Officer to ascertain shielding effectiveness and cumulative doses in occupied spaces adjacent to the shielding. If required, access will be restricted to limiting areas, e.g., the Amplifier and Assembly Areas, during shots that produce high neutron yields. If required to limit general dose rates, the shot schedule will be modified to limit integrated neutron production over time. The dosimetry report together with any analysis or restrictions imposed is filed as a permanent record Target Chamber Activation Surveys To monitor for neutron activation of the OMEGA Target Chamber and the OMEGA EP Target Chamber, a contact gamma radiation measurement at specific control points is performed semiannually. The control points to monitor are detailed in Survey Maps A-2a, A-2b, and A-2c for OMEGA and on Survey Maps A-3a, A-3b, A-3c, and A-3d for OMEGA EP in Appendix A. 15 August 2011 Page III-1

50 Part III: Requirements and Procedures These maps are also used to record and trend the data and are filed when completed. The survey is performed with the Ludlum Model 3 Portable Survey Meter with a Model 44-9 Detector by holding the probe in contact with the control point Target Bay General Radiation Surveys Target Bay general area gamma radiation surveys are taken periodically to determine the trend of neutron activation of structural components, whether or not the Target Bay requires posting as a Radiation Area, and to determine entrance requirements and stay times should radiation levels exceed 5 mrem/h. To determine the trend of neutron activation, a gamma radiation survey is conducted every six months or after the integral production of neutrons since the last survey, whichever is more frequent. If general area radiation levels of $1 mrem/h are detected during a periodic survey, a gamma survey must be conducted concurrent with any Target Bay entry after a shot that produces $10 15 neutrons. Additionally, any personnel entering the Target Bay prior to the completion of the survey must be monitored by a TLD. Following any high-yield shot, the radiation level will be measured at 30 cm from any structure and 30 cm from the TC at location G5 indicated on Map A-2b. If the survey indicates $5 mrem/h, the Radiation Area lights are illuminated, and anyone entering the Target Bay must be monitored by a TLD. Under these circumstances, the LLE Radiation Safety Officer will be notified. Access will be controlled and stay times employed to limit personnel exposures to within limits. The results of these gamma radiation surveys are logged on Survey Maps A-2a, A-2b, and A-2c for OMEGA and on Survey Maps A-3a, A-3b, A-3c, and A-3d for OMEGA EP in Appendix A for OMEGA EP and are filed when completed. For the points labeled G, a general area radiation reading is taken by holding the detector probe at waist level and turning 360. The average reading obtained is recorded. For the points labeled C, a contact radiation reading is taken by holding the detector probe in contact with the designated point Airborne Radiation Surveys Tritium airborne radiation surveys are taken: (1) when opening a component that has been exposed to tritium e.g., the Target Chamber, diagnostics, and the Lower Pylon when removing the MCTC after a cryogenic DT shot, and vacuum pump systems that communicate with the Target Chamber, (2) periodically while conducting decontamination procedures, and (3) after an accidental release of tritium. The Overhoff Model 700 BAcC Portable Tritium Monitor is used to verify that tritium is not being released above the limit applicable to radiation workers 20 nci/m 3 when initially opening diagnostic systems (e.g., the Target Positioner and 10-in. manipulators) during the conduct of DT cryogenic shot operations. If levels >20 nci/m 3 are detected, the diagnostic will be closed and purged to reduce the level of contaminiation. Procedures for entering the Target Chamber are found in Section 3008 of this manual and in the LFORM Sec The Overhoff Model 700 BAcC Portable Tritium Monitor is used to verify that the tritium concentration is below 20 nci/m 3 prior to entry. 15 August 2011 Page III-2

51 Part III: Requirements and Procedures (Not applicable for OMEGA EP provided the surface contamination is below 1000 DPM/100 cm 2 in the GCC.) During general decontamination operations of material exceeding 100,000 dpm/100 cm 2 conducted in LaCave or Room 136 (or other locations approved by the LLE Radiation Safety Officer), a tritium airborne survey must be conducted every 8 h using the Femto-TECH Tritium Monitor U24-D or equivalent to ensure that airborne levels are below 0.1 nci/m 3, the limit applicable to the general public. The results of these surveys will be recorded on Survey Log A-4 in Appendix A. In the event of an accidental release of tritium or the spill of tritiated material or water, an airborne survey will be taken with the Overhoff Model 700 BAc C Portable Tritium Monitor to determine if areas must be evacuated and/or breathing apparatus donned. The results of these surveys will be recorded on Survey Log A-4 in Appendix A. (See Part IV for emergency procedures.) 3004 Surface Contamination Surveys Area surface contamination surveys are conducted periodically in accessible areas adjacent to where tritium is handled, e.g., the Tritium Fill Station room, prior to allowing components that have been exposed to tritium to be removed from the Target Bay, TFS, TRS, or CTHS glovebox; prior to Target Chamber entry; during and after decontamination procedures; and after accidental spilling or release of tritium. Additionally, surface contamination surveys will be conducted during cryogenic DT operations, while performing maintenance on potentially contaminated equipment, on sinks used to discharge liquid waste via the sewer (see Section 3011), and prior to and during maintenance on drains used to discharge liquid waste. Survey results >1000 dpm/100 cm 2 in accessible, uncontrolled areas are to be reported immediately to the LLE Radiation Safety Officer. All accessible areas must be maintained at levels <1000 dpm/100 cm 2 ; inaccessible controlled surface contamination areas should be maintained at levels <10 5 dpm/100 cm 2. At LLE, the inside of the TFS, FTS, and DTHPS gloveboxes, the inside of diagnostics and ten-inch diagnostic manipulators (TIM s), and decontamination areas, when established, are designated controlled surface contamination areas. TIM s will be maintained <10 4 dpm/100 cm 2 on surfaces that can be reached with the door open. Other TIM surfaces closer to the Target Chamber that cannot be reached will be maintained <50,000 dpm/cm 2. The inside of the Target Chamber and CTHS components are highly contaminated and will not be maintained at any particular level. Area surface contamination surveys are conducted by dragging a wipe over approximately 100 cm 2 of the surface being surveyed and then counting the wipe in the Wallac Model 1409 or the Packard Tri-Carb Liquid Scintillation Counters. If the area is too small to swipe over 100 cm 2, due to small component parts or irregular surfaces, as much of the area as possible will be wiped and the results will be recorded per wipe rather than per 100 cm 2. Area surface contamination surveys will be conducted at the following frequencies and recorded on Survey Map A-4 in Appendix A: OMEGA Target Chamber each entry, semi-annually Room 157, LaCave, and Room 136 weekly Pump House/TC-TRS Room 150B when condensate is transferred Cart Maintenance Room 150A weekly, during DT MCTC maintenance Controlled contamination area during decontamination/maintenance operations every 8 hours 15 August 2011 Page III-3

52 Part III: Requirements and Procedures All materials removed from systems after being exposed to tritium, e.g., blast window assemblies, diagnostics, materials from the TFS glovebox, etc., must be controlled and handled as radioactive material until it is proven otherwise. To verify material to be uncontaminated, two consecutive surveys taken at least 8 h apart after completion of decontamination are required. Both consecutive survey results must be <1000 dpm/ One survey is not sufficient since tritium is known to diffuse to the surface over time. Only personnel who are trained as Radiation Workers will handle potentially contaminated materials, and controlled contamination procedures must be followed. When removing potentially contaminated material, personnel must wear rubber gloves, and the material must be immediately surveyed and placed in suitable containment such as a pet G container or plastic bag. All survey results will be logged on one of the Survey Map A-4 series in Appendix A. If the survey results indicate contamination >1000 dpm/100 cm 2, the item must be tagged and subsequently decontaminated as described in the paragraph above before it is released to uncontrolled areas. Whenever transferring potentially contaminated material outside of controlled surface contamination areas, it must be contained in suitable containment and transported by Radiation Worker qualified personnel along a predetermined route. All contaminated material must be controlled by attaching a numbered tag indicating the type and level of contamination, and by logging the tag number, contamination levels, and location in a Radioactive Materials log. (See Section 3011 for the applicable Radioactive Material handling procedures.) Prior to handling potentially contaminated material or entering controlled surface contamination areas, suitable anticontamination clothing must be worn (see Section 3006). Subsequent to removing anticontamination clothing, and if a risk of personal contamination exists (e.g., after contacting contamination levels >10 5 dpm/100 cm 2 ), personnel shall smear their hands and any other portion of their body subject to contamination with a glycerol-wetted smear paper as described in the following paragraph. In the case of exiting the Target Chamber or a controlled surface contamination area, the hands and exposed skin on the face shall be wiped with a glycerol-wetted smear. For other areas, if only rubber gloves were worn and no potential exists for other parts of the body to have become contaminated, then only the hands must be smeared. The results of these personnel smearing surveys need not be recorded as long as all levels are <1000 dpm/100 cm 2. If, however, any reading is >1000 dpm/100 cm 2, then all readings must be recorded on Survey Map A-4a in Appendix A and a urine sample must be taken and sent to the UR Radiation Safety Unit for analysis. A glycerol wipe is performed by squirting 0.25 ml of glycerine (glycerol) on a smear paper, rubbing two smear papers together to distribute the glycerine evenly on both, smearing the subject area, and counting the collected activity using the Liquid Scintillation Counter (LSC). After the smear is taken, the applicable body part will be washed with warm water and soap and rinsed with warm water Liquid Activity Surveys Liquid activity surveys are conducted periodically to monitor the activity levels of water used in ultrasonic sinks used for decontamination, waste from blast window decontamination, dedicated TRS chilled water systems, and condensate from molecular sieve drier regeneration. Highly contaminated samples from regeneration must be taken with a micro-pipette rather than 1 ml to prevent overranging the Wallac. Action will be taken to investigate and correct the cause of 15 August 2011 Page III-4

53 Part III: Requirements and Procedures increasing levels of activity. The UR Radiation Safety Unit has authorized LLE to discharge liquid waste via the sewer as long as LLE has verified the activity as averaged over a month is <1 # 10 2 nci/ml (see Section 3011). Ultrasonic sinks used for decontamination are sampled after each use for decontamination and at least weekly. The samples are counted using the Liquid Scintillation Counter and logged on the Sample Activity Log A-3 in Appendix A. Additionally, a log will be maintained at the ultrasonic sink giving time, date, person using, RLM # of the item decontaminated, and the sample results. The LLE Radiation Safety Officer will be informed of any sample that reads >100,000 dpm/ml Anticontamination Clothing Anticontamination clothing is worn to prevent contamination of personal clothing or skin. Full anticontamination clothing consists of full coveralls, hood, shoe covers, gloves, and a filtered mask or respirator. For particulate contamination, cloth clothing and a filtered mask are adequate; however, to protect against tritium, plastic clothing and an air-fed breathing apparatus may be required (see Section 1004). Partial contamination clothing consists of lab coat and gloves or lab coat, gloves, and shoe covers. At LLE the following are the normal anticontamination clothing requirements: Target Chamber entry <10,000 dpm/100 cm 2 rubber gloves, clean-room clothing with shoe covers, and face mask >10,000 dpm/100 cm 2 double clean-room clothing with shoe covers, rubber gloves, and face mask (double clothing is required to allow removal of one set when exiting the anticontamination area while still maintaining a set of clean-room clothing; additional hoods and shoe covers should be available at the control point to don when others are removed) Decontamination area <1000 dpm/100 cm 2 lab coat, shoe covers, and rubber gloves >1000 dpm/100 cm 2 full clean-room clothing (coveralls) with shoe covers and Material handling rubber gloves rubber gloves when only the hands come into contact with potentially contaminated material Special procedures must be followed when exiting a controlled surface contamination area, e.g., the Target Chamber or a decontamination area. First, the shoe covers must be removed in a manner that allows one to step down on a step-off area adjacent to the controlled surface contamination area immediately after removing the shoe covers. Then the hood and coveralls are removed, followed by the gloves. When removing anticontamination clothing, it should be turned outside-in to avoid spreading any contamination. After removal, all anticontamination clothing must be placed in segregated radioactive material containers for disposal Establishing a Controlled Surface Contamination Area To prevent the spread of contamination, a controlled surface contamination area shall be established whenever an area has or may have surface contamination in excess of established limits (>1000 dpm/100 cm 2 ). A barrier must be established around the perimeter of controlled surface contamination areas to prevent inadvertent entry, and the barrier will be posted with yellow and magenta signs stating Controlled Surface Contamination Area Do Not Enter. 15 August 2011 Page III-5

54 Part III: Requirements and Procedures Normally, entrance to and exit from a controlled surface contamination area shall be limited to one point, the control point. The control point will have a step-off area made of a disposable material such as an absorbent mat on the floor immediately outside of the controlled surface contamination area. This step-off area shall be maintained contamination free by frequently surveying and changing as required. The control point shall also have two radioactive material collection bags/packages, one for material to be laundered and one for material to be processed for disposal. When full, the UR Radiation Safety Unit will be called to process material for disposal. A squirt bottle containing glycerol will be provided to allow personnel smearing in accordance with Section 3004 upon exiting from the area. The control point established for Target Chamber entry shall be continuously manned whenever the Target Chamber is open and personnel are in the chamber. This Control Point Watch is responsible for ensuring both contamination control and for ensuring the safety of personnel. The Blast Window Assembly (BWA) and LaCave experimental operations decontamination areas must be maintained <10 4 dpm/100 cm 2 during decontamination work. At the end of the work period or the end of a shift, the area should be cleaned and surveyed to a level <1000 dpm/100 cm 2. If contamination levels >10,000 dpm/cm 2 are encountered, anticontamination clothing consisting of coveralls, gloves, and shoe covers must be worn and discarded or laundered after use in decontamination. The step-off area should ALWAYS be maintained below 1000 dpm/100 cm 2. If the step-off area is surveyed >1000 dpm/100 cm 2, follow the procedures of Section 4000 and notify the Group Leader and the OMAN OMEGA Operations supervisor immediately Target Chamber Entry Target Chamber (TC) entry requires both the Shot Director s and the Radiological Safety Officer s approval. TC entry shall be performed in compliance with approved procedures that fulfill the requirements of Sections 3006, 3007, and Note When the Target Chamber access port is open, a Control Point Watch must be stationed at all times to control personnel entry, communicate to Control in the event of an Emergency, and effect radiological controls. The appropriate procedure checklist, Target Chamber Access Authorization form, and a Personnel and Material Entry Log MUST be at the Control Point and maintained current throughout the TC entry. All personnel, tools, and equipment MUST be logged in and out of the TC. A Controlled Surface Contamination Area in accordance with Section 3007 shall be established in the area adjacent to the access port prior to opening the port. The Control Point watchstander must be a qualified Radiation Worker. The Target Chamber shall be brought to atmospheric pressure and purged in accordance with OMEGA operating procedures. The Target Chamber airborne tritium levels must be <20 nci/m 3 in accordance with Section (Not applicable for OMEGA EP as long as tritium surface contamination is below 1000 dpm/100 cm 2 in the GCC.) 15 August 2011 Page III-6

55 Part III: Requirements and Procedures A surface contamination survey shall be conducted in accordance with Section This shall include the access port and the adjacent area inside the Target Chamber (that area reachable from the access port without extending more than the head, trunk of the body, and arms inside the chamber). (Not applicable for OMEGA EP provided the surface contamination is below 1000 dpm/100 cm 2 in the GCC.) For OMEGA EP target chamber only, a c survey using the Ludlum Model 9 ion chamber shall be conducted upon opening the access hatch. If radiation levels $5 mrem/h are detected, the LLE Radiation Safety Officer must be notified and control procedures required for a Radiation Area as outlined in Section 3002, paragraph 4 must be used. Only personnel who are qualified as Radiation Workers and have been fitted for respirators are allowed entry into the Target Chamber. Target Chamber Close-Out The Target Chamber Access Authorization Form will be used to close out the Target Chamber Decontamination Procedures Due to its high mobility, tritium readily absorbs into the bulk of many materials as well as being present on the surface; subsequently it can outgas and readily spread. To release tritiated components from controlled contamination areas, surface contamination levels must be <1000 dpm/100 cm 2. To control occupational exposure, the ICRP recommends that surface contamination levels in occupied areas, e.g., the Target Chamber when open for access, be maintained <2 # 10 7 dpm/100 cm 2. Tritium-decontamination techniques include washing, vacuuming, purging, thermal desorption, isotopic exchange, chemical or electrochemical etching, plasma discharge, and melting. Experience with the decontamination of the original OMEGA indicates that vacuuming and washing are sufficient for the contamination levels expected to be encountered. Should these methods fail to reduce contamination levels sufficiently, additional procedures will be implemented with the approval of the LLE Radiation Safety Officer. The following decontamination procedures will be used at LLE: Decontamination must be performed within a controlled surface contamination area established in accordance with Section Component surface contamination surveys in accordance with Section 3004 must be performed before, during, and subsequent to decontamination. Before a component can be considered decontaminated, it must have levels of <1000 dpm/100 cm 2 on two consecutive surveys conducted at least 8 h apart. Surveys of the controlled surface contamination area must be performed in accordance with Section Appropriate anticontamination clothing in accordance with Section 3006 must be worn. Where particulate material is present, such as target debris inside of the Target Chamber, vacuuming with an approved radiological vacuum cleaner shall be performed. The vacuum filter must be collected and transferred to the UR Radiation Safety Unit for disposal. 15 August 2011 Page III-7

56 Part III: Requirements and Procedures Washing with soap (10% Brulin 8156D) and water. Paper wipes, Q-tips, etc., will be used as required to reach cracks and crevices. All materials used must be collected and transferred to the UR Radiation Safety Unit for disposal. Ultrasonic cleaning using a solution of DC13 and water. The water in the ultrasonic sink will be sampled in accordance with Section Components will be dried by using either absorbent wipes or natural convection air drying. Residue from decontamination must be transferred to the UR Radiation Safety Unit. For the Target Chamber, diagnostic inserters, the MCTC s, etc., purging with the moist room air to the TC-TRS will be used to reduce the levels of contamination. Surface contamination surveys will be logged on one of the Survey Map A-4 series in Appendix A. Before any component is released from a controlled area, it must have been shown to have <1000 dpm/100 cm 2 surface contamination on two consecutive surveys conducted at least 8 h apart; this result must be logged on one of the Survey Map A 4 series. Any component not successfully decontaminated to <1000 dpm/100 cm 2 must be controlled as radioactive material using the procedures of Section Internal Transfer of Tritium Targets To prevent the uncontrolled release of tritium into the environment, tritium targets must be continuously contained and controlled during transfer from the Tritium Fill Station (TFS) to the Target Positioner. The Tritium Transport procedures described in Section 2002 and outlined in Procedure SS-5-4, Part 4 (OMEGA System Operations Manual, Vol. II) are designed to provide this continuous containment. To ensure proper operation, only Technicians qualified as both Experimental Systems Technicians and Radiation Workers will insert targets containing tritium into the Target Positioner. The transfer route to the Target Bay will be via the LaCave to Target Bay stairwell access. Targets shall be transferred to the Target Positioner in a dedicated target bottle and inserted manually, while monitoring with the Overhoff Model 700 BAc C Portable Tritium Monitor. Spent target support and container materials will be handled as contaminated material and will be under the same continuous control during the return to the TFS for disposal Radioactive Material Accountability and Disposal To prevent the unauthorized or uncontrolled release of radioactive material, strict accountability procedures are required. For radioactive sources, the control and inventory procedures of the UR Radiation Safety Manual apply. For the control of generated radioactive material, this procedure supplements those of the UR Radiation Safety Manual. These procedures apply to all equipment and material that has the potential of being transported (since the Target Chamber, Tritium Fill Station, and CTHS are fixed pieces of equipment, they are exempted from the requirements of this section; however, component parts removed from these pieces of equipment must be handled in accordance with these procedures). At LLE, the most likely source of radioactive material is that generated through exposure to tritium. Secondary sources of radioactive material are structures activated by neutrons in OMEGA and components activated by high-energy cn reactions on OMEGA EP. Since the only source of transportable radioactive material at LLE is tritium, radioactive material requiring accountability under this section is that material that has surface contamination or suspected internal contamination of >1000 dpm/100 cm August 2011 Page III-8

57 Part III: Requirements and Procedures Equipment and components (other than fixed equipment, but including fixed equipment when it is removed) that are found to have surface contamination levels >1000 dpm/100 cm 2 must be controlled as follows: Attach a tag (shown below) that includes a control number, equipment or material description, isotope, level of activity, and date tagged. Radioactive Material Tag Tag Number: Material Nomenclature: Surface Contamination: Date: Log the radioactive material tag information into the Radioactive Material Control Log, A-5 in Appendix A; in addition include the current location of the tagged item, the ultimate disposition (decontaminated or transferred to UR Radiation Safety Unit), and the signature of the person verifying the ultimate disposition. There will be two such logs maintained, one by the Cryogenic and Tritium Facility Group and one by the Experimental Operations Group. Radioactive material shall be stored in authorized locations only. These locations must be labeled and locked. Approved storage locations in LLE are as follows: Room 157 (under the TFS glovebox and under the fume hood) An area located on the south side of the Target Bay A decontamination area located in La Cave A source locker located in Room 150B The Blast Window Decontamination area, Room 136 Moving Cryostat Transfer Cart Maintenance Area, Room 150A An inventory and audit of all radioactive material including reconciliation of the Radioactive Material Log shall be performed quarterly by the Cryogenic & Tritium Facility and Experimental Operations Group Leaders. The results of these audits must be reported in writing to the LLE Radiation Safety Officer. Radioactive material that is no longer required will be transferred to the UR Radiation Safety Unit. Only the UR Radiation Safety Unit is authorized to dispose of radioactive material; under no circumstances will anyone from LLE dispose of radioactive material. The single exception to this is that LLE has been authorized (by RSO letter dated 13 August 1998) to discharge liquid wastes subject to compliance with the following paragraph: Tritiated liquid having an activity <1 # 10 2 nci/ml (22,000 dpm/ml) may be disposed of via approved sewage drains. This includes liquid from ultrasonic sinks used for decontamination and liquid collected from blast window decontamination. The drains used to dispose of liquids must be marked as potentially contaminated to ensure proper radiological control precautions are taken prior to performing maintenance. For all liquids disposed of via the sewage drains, the sample activity and volume must be recorded and reported to the LLE Radiological Safety Officer with a copy to the UR Radiation Safety Officer. This report will include a copy of the scintillation counter printout annotated to identify samples which represent liquid discharges. Additionally, a cumulative record of the volume and activity of liquid will be maintained for 15 August 2011 Page III-9

58 Part III: Requirements and Procedures each drain used to discharge liquid wastes. Subsequent to such liquid discharges, a surface contamination survey of sinks used for the discharge must be performed in accordance with Section LLE Monthly Tritium Inventory Running inventories shall be maintained for the tritium that exists within the Cryogenic Target Handling System (CTHS) and associated equipment and for tritium contained in warm targets. The Cryogenic and Tritium Facility Manager shall maintain the CTHS, associated equipment, and the warm target inventory. The CTHS inventory shall make an accounting of all the tritium within the CTHS and the following equipment: Tritium Fill Station (TFS) and the Tritium Removal Systems (OMEGA scrubber, TFS glovebox, Rm157 TRS, and TC-TRS). The inventory shall be kept using an electronic log in the format of Form A-2. The warm-target inventory shall be kept in the format of Form A-3. The inventories shall be submitted to the LLE Radiation Safety Officer monthly. For the CTHS inventory, if the amount of tritium contained in the system outside of the uranium beds exceeds 4% of the system total, the Cryogenic and Tritium Facility manager shall investigate the problem and notify the LLE RSO Scintillation Counting Scintillation counting is used to count contamination survey wipe samples and liquid samples such as water. A Wallac Model 1409 or Tri-Carb 2910TR Liquid Scintillation Counter is used for all scintillation counting performed at LLE. The procedures for operating the counter are contained in the applicable Instrument Manual. Both units automatically determine quench, chemiluminescence, and counting efficiency and subtract background. A background sample must be prepared and placed in position in one of the correct (labeled) racks. For water samples, the result in nci/ml of counting a 1-ml sample is obtained by dividing the dpm of the sample by 2.22 # 10 6 dpm/nci. Caution: Liquid activities that exceel 500,000 CPM will saturate the counter and must be diluted and recounted in order to obtain accurate results Personnel Monitoring As long as the shield is verified by the procedures of Section 3000 to be performing to design criteria, TLD with CR 39 monitoring of personnel outside of the primary shielding is not required. Personnel TLD monitoring is required only in the x-ray laboratory, in the Target Bay should neutron activation increase the general area radiation levels there to >5 mrem/h, or when handling activated materials. Since tritium cannot be detected by a film badge or TLD, exposure to tritium is monitored by bioassay (urine analysis). Anyone using more than 100 mci of tritium is required to have a bioassay performed within one week following a single operation and at weekly intervals for continuous operations. At LLE, the following personnel monitoring is required: X-ray laboratory personnel: Monitored by TLD counted quarterly via the UR Radiation Safety Unit. 15 August 2011 Page III-10

59 Part III: Requirements and Procedures Exp. Operations personnel: Monitored by TLD counted quarterly via the UR Radiation Safety Unit; weekly urine analysis by the UR Radiation Safety Unit anytime gaseous DT targets are handled or the Target Chamber is entered during a particular week. TFS operators: Weekly urine analysis by the UR Radiation Safety Unit when targets are filled, the TFS or CTHS gloveboxes are accessed, or maintenance is performed on highly contaminated CTHS components during a particular week. Radiation-area entrants: Anyone who enters the Target Bay when it is a radiation area is monitored by thermal luminescent dosimeter (TLD) 760 Badge counted monthly Material with Fixed Activity Gamma-radiation surveys will be taken whenever neutron or high-energy c-n activation of material is suspected. The level of fixed activation will be measured by bringing the Ludlum Model 9 portable ion chamber in contact with the activated material. Material with radiation levels exceeding 0.1 mrem/h will be designated as Radioactive Material. The activated material must be controlled by attaching a numbered tag indicating the level of activation, the date, and by logging the tag number, contamination level and location in the Radiation Materials log. Tagged material will be resurveyed in conjunction with the quarterly audit of the Radioactive Material Log. If the level of activity has decayed to <0.1 mrem/h, the material may be cleared from the log. Any work with this material must be carried out in designated radioactive areas by qualified radiation workers wearing film badges and a finger dosimeter. Work on activated material with radiation levels exceeding 5 mrem/h on contact requires LLE Radiation Safety Officer approval. Section 3011 describes the material accountability and disposal procedures Establishing a Radioactive Materials Control Area Work on activated material with radiation levels exceeding 5 mrem/h on contact requires LLE Radiation Safety Officer approval. The work will be performed by a qualified radiation worker wearing a TLD and a finger dosimeter. A barrier must be established to prevent inadvertant entry at a distance that ensures the radiation levels do not exceed 0.1 mrem/h. The barrier will be posted with yellow and magenta signs stating Radioactive Materials Control Area Do Not Enter. Entrance to and exit from the radioactive materials control area shall be limited to one point, the control point. 15 August 2011 Page III-11

60 Part IV: Emergency Procedures Part IV Emergency Procedures 4000 Spill of Radioactive Material 4001 High Airborne Activity 4002 Acute Release of Tritium (from Tritium Fill Station) 4003 UR Security Acute Release of Tritium Emergency Procedure 15 August 2011 Page IV-1

61 Part IV: Emergency Procedures 4000 Spill of Radioactive Material Note Since the only potential spill at LLE is tritium, this procedure is tritium specific. It implements the procedural requirements found in Sec. 17 of the UR Radiation Safety Manual. a. Immediate Actions 1. Stop the spill if this can be accomplished without risk of personal contamination. Cover the spill with absorbent paper. 2. Warn others in the area and notify your supervisor (the supervisor should notify the LLE Radiation Safety Officer who, in turn, will notify the UR Radiation Safety Unit). 3. Isolate the affected areas by closing doors, putting up barriers, and/or guarding the entrances to the area. 4. Minimize your exposure to radiation and/or radioactive materials. 5. Secure local fans if radioactive material could be spread by the fans. b. Supplementary Actions 1. Clean up the spill as directed by the UR Radiation Safety Unit. 2. Complete surface contamination surveys before, during, and after decontamination of the affected area. 3. Smear personnel who may have been contaminated. 4. Obtain urine samples from personnel who may have been exposed above allowable limits and submit samples to the UR Radiation Safety Unit for analysis. 15 August 2011 Page IV-2

62 Part IV: Emergency Procedures 4001 High Airborne Activity a. Indications 1. Sustained airborne activity >20 nci/m 3 in controlled areas. 2. Sustained airborne activity >5 nci/m 3 in uncontrolled areas. 3. Tritium Monitor alarm. 4. Tritium Fill Station or CTHS glovebox tritium alarm. b. Immediate Actions 1. If acute release of tritium is suspected, refer to Section Notify the LLE Radiation Safety Officer. 3. Secure operations that may have caused the high airborne activity. 4. Rig temporary ventilation under controlled conditions to reduce airborne levels to <1 nci/m 3 to the TC-TRS. 5. Obtain an airborne reading using both the Overhoff Model 700 BAc C Portable Tritium Monitor and the more-sensitive Femto-TECH Tritium Monitor U24-D. 6. If airborne levels are >5 nci/m 3 in uncontrolled areas, evacuate non-radiation Worker personnel to unaffected areas. 7. If airborne levels are >20 nci/m 3, limit stay times to the following: Airborne Level nci/m3 Stay Time <20 Unlimited 20 to h 100 to h 1000 to 10,000 1 h c. Supplementary Actions 1. Obtain urine samples from personnel who may have been exposed above allowable limits and submit samples to the UR Radiation Safety Unit for analysis. 2. Notify the UR Radiation Safety Unit at Ext Submit an estimate of the total activity released. 15 August 2011 Page IV-3

63 Part IV: Emergency Procedures 4002 Acute Release of Tritium a. Indications 1. TFS or CTHS glovebox tritium alarm 2. Room 157 tritium monitor alarms (TMT-1001 and/or TMT-1002) are >2000 nci/m 3. Sustained airborne activity >5 nci/m 3 in uncontrollable areas. b. Immediate Actions 1. Evacuate Room 157 to the adjacent corridor, ensuring that both the inner room and outer room doors are shut and the exhaust fan is running. 2. Pass the word on the general announcing system Tritium alarm, all non-emergency personnel move from the effected area to adjacent areas. Remain indoors. Shut all doors and windows. 3. Contact Facility Support (John Sawyer at ) to shut the northside outside air intakes to the building ventilation system. 4. The LLE Radiation Safety Officer or, in his absence, the senior qualified LLE Radiation Worker from the Cryogenic & Tritium Facility will take charge of the emergency response until UR Radiation Safety Unit personnel arrive. 5. Establish a control point in the corridor outside the sliding doors of Room 157. The personnel who evacuated Room 157 should remain in this area to prevent the spread of possible contamination. 6. LLE Radiation Safety personnel will conduct tritium airborne surveys in the occupied areas of the LLE complex to verify that the atmosphere is environmentally safe and that the release of tritium has been limited to Room 157. c. Additional Immediate Actions 1. Notify University Radiation Safety Unit (5-3781). Security personnel will proceed to East River Road and Murlin Drive to clear any of the public that may be south of East River Road and within 200 yards of the LLE complex. This should be accomplished by posting patrol personnel at the following locations: 300 yards east of Murlin Drive on East River Road, at the intersection of East River Road and Murlin Drive, 300 yards south on Murlin Drive toward Whipple Park, and at the entrance to River Road Laboratory on East River Road. Security personnel should not loiter in areas toward LLE from these perimeter points until Radiation Safety personnel have determined that it is environmentally safe to do so. University Radiation Safety Unit personnel will proceed to the scene to assume command of the response actions and to take perimeter air samples. d. Supplementary Actions 1. UR Radiation Safety will immediately notify and report to regulatory authorities as required by NYS DEC Publication of 6 NYCRR, Part 380, Sec Perform personnel monitoring and decontamination of personnel who were in the TFS room at the time of the emergency or are suspected to have been exposed. Perform 15 August 2011 Page IV-4

64 Part IV: Emergency Procedures testing for tritium in urine of all potentially exposed personnel. These samples will be submitted no later than 6 hours after the incident and no later than 24 hours following the release. 3. Perform airborne and wipe surveys and decontamination of Room 157. Note: Personnel shall don appropriate anticontamination suits prior to entry, and stay times based on airborne levels should be established as follows: Airborne Level nci/m3 Stay Time <20 Unlimited 20 to h 100 to h 1000 to 10,000 1 h 4. Secure the TFS and CTHS and ascertain the damage and cause. 5. Document the incident including all survey results and incorporate these records into the TFS Decommissioning Record. 15 August 2011 Page IV-5

65 Part IV: Emergency Procedures 4003 UR Security Acute Release of Tritium Emergency Procedure (from LLE Tritium Fill Station) a. During on-duty hours or when the TFS is operating and occupied 1. University Security and UR Radiation Safety will take the actions specified in the attached procedure. b. During off-duty hours or when the TFS is not occupied 1. University Security will (a) Immediately notify all of the following LLE Personnel (starting from the top of list) Office Cell Roger Janezic Michael Koch Salvatore Scarantino Patrick Regan Walter Shmayda Samuel Morse (b) Immediately notify the Duty Radiation Safety on-call person by calling or digital pager If no one is present, the prerecorded message will give the duty person s home phone and page number. (c) Take the actions specified for Security in Section LLE Personnel contacted will immediately report to LLE, attempt to contact other LLE personnel on the call list, and take actions specified in Section University Radiation Safety personnel will immediately report to LLE and take actions specified in the attached procedure. In the event that UR Radiation Safety personnel are the first on the scene, they will take charge and take all actions specified in Section 4002 until relieved by qualified LLE personnel. 15 August 2011 Page IV-6

66 Part V: Maintenance Procedures Part V Maintenance Procedures 5000 Tritium Fill Station Radioalogical Requirements 5001 Experimental Operation Radiological Requirements 5000 Tritium Fill Station Radiological Requirements (see also the preventive maintenance requirements for the CTHS System Description, Volume IV) Radiological Requirement Reference Periodicity Radiation Detection Instruments Femtotech U-24D Area Monitor Verify flow rate via rotameter Tech. Man. ( 6.1) M Device test Tech. Man. ( 6.1.2) W Vendor calibration Tech. Man. ( 4.5.3) A Overhoff Tritium Air Samplers (2) (EPA Monitors) Replace/verify performance of catalytic oxidizer Inst. Manual A Calibrate constant flow controller Inst. Manual A Calibrate rotameter ASTM Procedure A Overhoff Model 700 BAcC Portable Tritium Monitor Replace batteries Inst. Manual R, Q Vendor calibration Inst. Manual A Ludlum Model 3 Portable Survey Meter with a Model 44-9 Detector Check batteries Inst. Manual R Replace batteries Inst. Manual R, Q Calibrate Inst. Manual A Airborne Radiation Surveys Opening of glovebox Sec R During decontamination procedures Sec R (8 h) During DT piping/system maintenance Sec R Room 150A during DT cart maintenance Sec W Room 150B during condensate transfer Sec R LaCave Lower Pylon (MCTC connect and Sec R disconnect) 15 August 2011 Page V-1

67 Part V: Maintenance Procedures Surface Contamination Surveys Room 157 and anteroom Sec W Room 2828 Sec W Room 136 Decontamination area Sec W Room 150A (Cart Maintenance Room) Sec R Room 150B (Pump Room) Sec R Upon removal of potentially tritiated components Sec R Controlled surface contamination area Sec R (4 h) Before, during, and after decontamination Sec & 3009 R, R (8 h after) Sinks used for liquid disposal Sec R Liquid Activity Surveys Ultrasonic sinks during decontamination Sec D TC-TRS condensate Sec R Room 157 TRS condensate Sec R TC-TRS chilled water Sec R Room 157 TRS chilled water Sec R CTHS DTHPS syringe pump oil Sec R Radioactive Material Accountability Inventory and audit Sec Q Liquid disposal report Sec R Personnel Monitoring Urinalysis testing Sec R, W Smear testing Sec & Experimental Operation Radiological Requirements (see also the preventive maintenance procedures in OMEGA Operations Manual, Volume III) Radiological Requirement Reference Periodicity Radiation Detection Instruments Overhoff Model 700 BAcC Portable Tritium Monitor Replace batteries Inst. Manual R, Q Calibrate Inst. Manual A Ludlum Model 3 Portable Survey Meter with a Model 44-9 Detector Check batteries Inst. Manual R Replace batteries Inst. Manual R, Q Calibrate Inst. Manual A Ludlum Model 9 Ion Chamber Check/replace batteries Inst. Manual R Check/replace dessicant Inst. Manual Q Calibrate Inst. Manual A 15 August 2011 Page V-2

68 Part V: Maintenance Procedures Wallac Model 1409 Liquid Scintillation Counter Count standard source using tritium protocol Transfer data files Back up system disk Manufacturer s service Perkin Elmer Tri-Carb 2910TR Count standard source using tritium protocol Transfer data files Manufacturer s service Shielding Shielding effectiveness monitoring Sec Q, R (10 16 total n) Target Bay Structure Activation Target Chamber activation Sec S Target Bay general radiation Sec S, R (10 16 total n) R (10 15 n/shot) Airborne Radiation Surveys Target Chamber entry Sec & 3003 R Diagnostics that communicate with Target Chamber Sec R Opening of other tritiated components Sec R During decontamination procedures Sec R (8 h) Surface Contamination Surveys Target Chamber Entry Sec & 3004 R Target Chamber Sec S Target Bay Sec S LaCave Sec S Room 147 decontamination area Sec W Target positioner and target storage area Sec R Upon removal of potentially tritiated components Sec R Controlled surface contamination area Sec R (4 h) Potentially tritiated diagnostics Sec M Before, during, and after decontamination D W M A D D A Sec & 3009 R, R (8 h after) Target Chamber alpha swipe (send to UR Q Radiological Safety) Liquid Activity Surveys Ultrasonic sinks during decontamination Sec D Radioactive Material Accountability Inventory and audit Sec Q Personnel Monitoring Urinalysis testing Sec W, R 15 August 2011 Page V-3

69 Appendix A Survey Maps, Logs, and Forms

70 15 August 2011 Radiation Shielding Monitoring Points Survey Map A-1 OMEGA Shield Point # Location 1 Experimental System Operation Station (Control Room) 2 Target/Laser Bay anteroom, south wall 3 Stairway opposite from Target Bay normal entrance door 4 Target Bay emergency exit, at foot of stairs on wall adjacent to LDL 5 Retired 6 Retired 7 Retired 8 Retired 9 Shaped-pulse damage testing facility, south wall/ceiling 10 LaCave Darkroom, east wall 11 Passageway between LaCave and Capacitor Bays on Capacitor Bay wall 12 Control Room, southwest corner 13 Rod Amplifier Room, east wall 14 Amplifier Test and Assembly Area, behind ultrasonic sink 15 Amplifier Test and Assembly Area, behind spare SSA storage area 16 Laser Bay, north wall (center) 17 Laser Bay, north wall (west) 18 Laser Bay, west wall (center near ceiling) 19 Retired 20 Laser Bay, south wall (east and low) OMEGA Shield Point # Location 21 LDL target area, south wall (east of target chamber) 22 LDL target area, south wall (middle of target chamber) 23 LDL target area, south wall (west of target chamber) 24 LDL target area, west wall 25 LDL target area, north wall 26 LDL target area, ceiling east side of chamber 27 Anteroom to target bay stairway south wall 28 OMEGA control room, under Shot Director s desk A-1 i

71 15 August 2011 OMEGA EP Shield Point # Location 29 Adjacent to 4-ft 4-ft beam hole in shield on OMEGA EP side (Prior to OMEGA EP activation this will give OMEGA background after the OMEGA/OMEGA EP shield; after activation it will determine the necessity for OMEGA closed access due to OMEGA EP hard x rays.) 30 On the first landing of the OMEGA EP Viewing Gallery located at the approximate centerline of the OMEGA EP Target Chamber (In particular this will monitor for the 10-MeV x rays expected to be in a 45 cone and will determine the necessity to install up to 10 cm of lead shielding.) 31 Video Conference Room west end service corridor above electric panel near ceiling above the fire alarm enunciator (This will check hard-x-ray levels outside the 45 cone.) 32 EXOps Film Processing Room west wall near ceiling above fire alarm enunciator 33 EXOps Diagnostic Work Area in SW corner near ceiling on west wall by the door 34 Laser Bay entrance near Control Room on the back side of the curtain shield 35 Amplifier Assembly Area door to Laser Bay on the back side of the curtain shield 36 Optical Assembly Area door to Laser Bay on the back side of the curtain shield (in NW corner of Rm. 6120) 37 South end on Laser Bay centerline on inside of shield wall 38 Diagnostics Bay below Target Chamber, center bay near ceiling on the north wall 39 North center Capacitor Bay on the north wall near the ceiling 40 Pump Room vestibule (fire exit from OMEGA Target Bay) on the east wall high on the sealed door 41 OMEGA EP Mechanical Room east wall between AHU 2 and 5 high up near the ceiling A-1 ii

72 15 August 2011 Target Bay Gamma Radiation Survey Map A-2a Target Bay Ground Level Date: Surveyed by: All readings in mrem/h N C1: G1: C2: C5: G2: C3: G3: C4: G6526 Notes: (1) All readings in mrem/h (2) All G (general area) measurements are taken 30 cm from any structure and averaged over a 360 sweep. A-2 i

73 15 August 2011 Target Bay Gamma Radiation Survey Map A-2b TMS Platform Level 1 Date: Surveyed by: All readings in mrem/h N C6: C7: G4: G5: C8: C9: G6527 G5: 30 cm from equatorial plane of TC and 30 cm from any structure at that location. A-2 ii

74 15 August 2011 Target Bay Gamma Radiation Survey Map A-2c TMS Platform Level 2 Date: Surveyed by: All readings in mrem/h N C11: C10: G6: G7: C12: C13: G6528 A-2 iii

75 OMEGA EP Target Bay Gamma Radiation Survey Map A-3a TAS Platform Level 1 15 August 2011 Date: Surveyed by: G13: C29: C16: G11: C17: C15: G10: G9: G8: C14: G8306J1 Notes: (1) All readings in mrem/h (2) All G (general area) measurements are taken 30 cm from any structure and averaged over a 360 sweep. A-3 i

76 OMEGA EP Target Bay Gamma Radiation Survey Map A-3b TAS Platform Level 2 15 August 2011 Date: Surveyed by: G16: C20: C19: G17: C21: C23: C22: G15: C24: G14: C18: G8307J1 Notes: (1) All readings in mrem/h (2) All G (general area) measurements are taken 30 cm from any structure and averaged over a 360 sweep. A-3 ii

77 OMEGA EP Target Bay Gamma Radiation Survey Map A-3c TAS Platform Level 3 15 August 2011 Date: Surveyed by: G20: C27: C26: G19: C28: G21: C25: G18: G8308J1 Notes: (1) All readings in mrem/h (2) All G (general area) measurements are taken 30 cm from any structure and averaged over a 360 sweep. A-3 iii

78 OMEGA EP Target Bay Gamma Radiation Survey Map A-3d Laser Bay Floor 15 August 2011 Date: Surveyed by: G22: G23: TGATS G8309J1 Notes: (1) All readings in mrem/h (2) All G (general area) measurements are taken 30 cm from any structure and averaged over a 360 sweep. A-3 iv

79 Surface Contamination Survey Log (Wallac & Tri-Carb) Survey Map A-4a 15 August 2011 Item/Area Surveyed: RMCL Number (if applicable): Rack Position Number S W I P E # First Final S S S Intermediate Post Decon W W W I I I DPM / 100 cm 2 P DPM / 100 cm 2 P DPM / 100 cm 2 P DPM / 100 cm 2 E E E # # # 1 BG BG BG BG First Survey FILE Performed BY Date Intermediate FILE Performed BY Date Post Decon FILE Performed BY Date Final Survey FILE Performed BY Date

80 Item/Area Surveyed: Decontamination Area (LaCave) Surface Contamination Survey Log Survey Map A-4b 15 August 2011 RMCL Number (If applicable): Wipe No. Rack Position dpm/ 100 cm 2 2 Cap Bays Table To elevator 1 Floor 3 4 Contamination area (LaCave) Cabinets To LaCave 5 XOPS storage shelf Background G6529 Cocktail BKG File ID: Initial: In Process: Intermediate: Final: Survey Performed by: Date:

81 Surface Contamination Survey Log Survey Map A-4c 15 August 2011 Item/Area Surveyed: LSC Area RMCL Number (If applicable): Wipe No. Rack Position dpm/ 100 cm 2 G6530a 1 To Target Bay 3 5 Tri-carb Floor 4 Table 2 Darkroom LaCave Background Cocktail BKG File ID: Initial: In Process: Intermediate: Final: Survey Performed by: Date:

82 Item/Area Surveyed: Micro Assembly Lab, Room 2828 Surface Contamination Survey Log Survey Map A-4d 15 August 2011 RMCL Number (If applicable): G6531 Wipe No. Rack Position Background Cocktail BKG dpm/ 100 cm 2 File ID: Initial: In Process: Intermediate: Final: Survey Performed by: Date:

83 Surface Contamination Survey Log Survey Map A-4e 15 August 2011 Item/Area Surveyed: Tritium Facility RMCL Number (If applicable): Key: F = Floor G = Glove S = Surface O = Other (Specify) Hood G6532J1 TFS glovebox TGCCharacterization station No. 2 FTS No. 1 Pump Characterization station No. 1 FTS No. 2 FTS-1 desk FTS-2 desk TRS desk TRS Wipe No. Rack Position Background dpm/ 100 cm 2 Cocktail BKG File ID: Initial: In Process: Intermediate: Final: Survey Performed by: Date:

84 Item/Area Surveyed: OMEGA Target Bay Ground Level Surface Contamination Survey Log Survey Map A-4f 15 August 2011 RMCL Number (If applicable): Wipe No. Rack Position dpm/ 100 cm 2 G6533 #3 #7 #1 #2 #6 N #4 #5 Background Cocktail BKG File ID: Initial: In Process: Intermediate: Final: Survey Performed by: Date:

85 Item/Area Surveyed: OMEGA Target Bay TMS Platform Level 1 Surface Contamination Survey Log Survey Map A-4g 15 August 2011 RMCL Number (If applicable): Wipe No. Rack Position dpm/ 100 cm 2 G6534 #10 (H7) #8 (H14) #9 N Background Cocktail BKG File ID: Initial: In Process: Intermediate: Final: Survey Performed by: Date:

86 Item/Area Surveyed: OMEGA Target Bay TMS Platform Level 2 Surface Contamination Survey Log Survey Map A-4h 15 August 2011 RMCL Number (If applicable): Wipe No. Rack Position dpm/ 100 cm 2 G6535 #11 (P1 Stairs) #13 #12 Console shelf N Background Cocktail BKG File ID: Initial: In Process: Intermediate: Final: Survey Performed by: Date:

87 Surface Contamination Survey Log Survey Map A-4i 15 August 2011 Item/Area Surveyed: Target Positioner 2 RMCL Number (If applicable): Wipe No. Rack Position dpm/ 100 cm Touchscreen 2. North side shelf 3. North side of web 4. TPS-2 gate valve 5. TPS-2 handles 6. Bottom of web 7. TPS-2 door 8. South side of web 9. South side shelf 10. Target storage box 11. Target bottle storage rack Please note any additional swipe locations if needed. 11 Background G6536 Cocktail BKG File ID: Initial: In Process: Intermediate: Final: Survey Performed by: Date:

88 Surface Contamination Survey Log Survey Map A-4j 15 August 2011 Item/Area Surveyed: LaCave Darkroom RMCL Number (If applicable): Wipe No. Rack Position dpm/ 100 cm 2 Jobo 3 Jobo 2 Jobo 1 4 Film drier 5 1. Darkroom floor 2. Film drop-off shelf 3. Film pickup shelf 4. Dry darkroom work table 5. Film-labeling counter 6. Door handle Background G6537 Cocktail BKG File ID: Initial: In Process: Intermediate: Final: Survey Performed by: Date:

89 Surface Contamination Survey Log Survey Map A-4k 15 August 2011 Item/Area Surveyed: LaCave RMCL Number (If applicable): Wipe No. Rack Position dpm/ 100 cm 2 G6538bJ2 Consult LaCave Contamination Area Survey Map Cryostat doorway Beneath spools 2. In front of MCTC parking lot 3. MCTC staging area 1 4 Spool 5 N 5 Background Cocktail BKG File ID: Initial: In Process: Intermediate: Final: Survey Performed by: Date:

90 Item/Area Surveyed: Decontamination Area Room 136 Surface Contamination Survey Log Survey Map A-4l 15 August 2011 RMCL Number (If applicable): Wipe No. Rack Position dpm/ 100 cm 2 Counter Shelves Entry Cabinet Sink Counter Background G6539 Cocktail BKG File ID: Initial: In Process: Intermediate: Final: Survey Performed by: Date:

91 Item/Area Surveyed: Cart Maintenance Room (Room 150A) Surface Contamination Survey Log Survey Map A-4m 15 August 2011 RMCL Number (If applicable): G6540 Wipe No. Rack Position Background dpm/ 100 cm 2 Cocktail BKG File ID: Initial: In Process: Intermediate: Final: Survey Performed by: Date:

92 Item/Area Surveyed: Pump House/TC-TRS (Room 150B) Surface Contamination Survey Log Survey Map A-4n 15 August 2011 RMCL Number (If applicable): G6541 Wipe No. Rack Position Background dpm/ 100 cm 2 Cocktail BKG File ID: Initial: In Process: Intermediate: Final: Survey Performed by: Date:

93 Surface Contamination Survey Log Survey Map A-4o 15 August 2011 Item/Area Surveyed: Room 157 Hood RMCL Number (If applicable): Wipe No. Rack Position dpm/ 100 cm 2 G7548J1 Sill Wall Floor Backsplash Glass Wall Background Cocktail BKG File ID: Initial: In Process: Intermediate: Final: Survey Performed by: Date:

94 Surface Contamination Survey Log Survey Map A-4p 15 August 2011 Item/Area Surveyed: TC-TRS Condensate Transfer RMCL Number (If applicable): Wipe No. Rack Position dpm/ 100 cm 2 HV Floor in front of T-8701 G7587J1 T-8701 HV HV HV Sampling port HV P /4 wrench Screw driver 5/8 wrench Pipette Receiving container Background Cocktail BKG File ID: Initial: In Process: Intermediate: Final: Survey Performed by: Date: