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SU-E-T-107: An Investigation Into Attenuation Effects On the Energy Spectrum for {sup 137}Cs Using Monte Carlo Techniques

Purpose: The calibration of radiation protection instrumentation including ionization chambers, scintillators, and Geiger Mueller (GM) counters used as survey meters are often done using {sup 137}Cs irradiators. During calibration, irradiators use a combination of attenuators with various thicknesses to modulate the beam to a known air-kerma rate. The variations in energy spectra as a result of these attenuators are not accounted for and may play a role in the energy-dependent response of survey meters. This study uses an experimentally validated irradiator geometry modeled in the MCNP5 transport code to characterize the effects of attenuation on the energy spectrum. Methods: A Hopewell Designs G-10 {sup 137}Cs irradiator with lead attenuators of thicknesses of 0.635, 1.22, 2.22, and 4.32 cm, was used in this study. The irradiator geometry was modeled in MCNP5 and validated by comparing measured and simulated percent depth dose (PDD) and cross-field profiles. Variations in MCNP5 simulated spectra with increasing amounts of attenuation were characterized using the relative intensity of the 662 keV peak and the average energy. Results: Simulated and measured PDDs and profiles agreed within the associated uncertainty. The relative intensity of the 662 keV peak for simulated spectra normalized to the intensity of the unattenuated spectra ranged from 0.16% to 100% based on attenuation thickness. The average energy for simulated spectra for attenuators ranged from 582 keV with no attenuation to 653 keV with 5.54 cm of attenuation. Statistical uncertainty for MCNP5 simulations ranged from 0.11% to 3.69%. Conclusion: This study successfully used MCNP5 to validate a {sup 137}Cs irradiator geometry and characterize variations in energy spectra between different amounts of attenuation. Variations in the average energy of up to 12% were determined through simulations, and future work will aim to determine the effects of these differences on survey meter response and calibration.

07 ISOTOPE AND RADIATION SOURCES↗

Perspective of Dry Interim Storage of spent nuclear fuel (SNF) in Germany - 16504

According to article 6 of the German Atomic Energy Act, storage of nuclear fuel requires a license granted by the competent authority in this field, which is the Federal Office of Radiation Protection (BfS). In 1995 a license to store spent nuclear fuel (SNF) and vitrified high level waste (HAW) for the interim storage facility in Gorleben was granted. Dry Interim storage at the NPP sites in its current form started in 2002 in the interim storage facility next to the NPP Lingen. Since this time each NPP erected its own storage facility and three central storage facilities have been built. All of these facilities use dry storage in metallic dual purpose casks (DPCs). The actual storage licenses for SNF and the HAW (which is stored in Gorleben and Rubenow) are limited to 40 years after closing the transport and storage cask. The first license to expire will be the storage license for the interim storage facility in Gorleben in the year 2035. During the last two years in Germany preparatory consultations to restart the process towards finding a repository took place, organised by a commission. This commission consists out of members of the scientific community, religious parties, social communities and environmental organizations as well as members of federal states governments and members of the German parliament (Bundestag). Recently the commission decided to extend its working period until summer 2016. Taking this background into account the BfS plans to intensify its research into extended interim storage with respect to potential material degradation leading to interference with safety functions (e.g. neutron shielding material). This includes topics like ageing of gasket materials which are already studied in ongoing experiments (e.g. at the Federal Institute for Materials Research and Testing) as well as studies of long term relaxation processes. Because of the need for transportation towards a repository another important field is the behaviour of the inventory under mechanical loads. For SNF, first calculations showed that because of the lower decay heat load after 60 years due to hydride reorientation the cladding might become much less ductile. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Transition of Site Programs and Implementation of Decommissioning Agent Oversight - 18653

San Onofre Nuclear Generating Station (SONGS), acting as the Decommissioning Agent (DA), retained its NRC License and contracted a Decommissioning General Contractor (DGC) to manage plant programs and execute the Demolition and Dismantlement (D and D) of the site. To support this effort, over twenty-one programs transitioned to the DGC, including Occupational Health and Safety, Radiation Protection, Maintenance, and Engineering. In addition, SONGS effected a cultural shift from an operating station performing work to one of overseeing it. The framework and road map for this kind of D and D model is new to the industry. This presentation shares the SONGS' experience and lessons learned associated with this transition. SONGS and the DGC entered an Agreement that established the contractual requirements and responsibilities of the parties, the Scope of Work, and the mechanisms to ensure effective transition of DA programs. As a part of defining the Scope of Work, the contract established functional areas and divisions of responsibilities divided into the roles of Governance, Oversight, Support, and Performance (GOSP). SONGS, as the site licensee, retained the Oversight Role, with exception of those areas that, as the licensee, would be considered 'non-transferrable'. Examples of those programs were Quality Assurance, Regulatory Affairs and the Employee Concerns Program. While the Company retained responsibility for those programs the DGC developed their own 'parallel' programs that the Company provided Oversight for. In 2017, SONGS completed program transition within one-year following the notice-to-proceed. The overall strategy was driven by three high-level principles: - Allow the DGC the most unfettered execution of the Scope of Work with limited interference and maximum self-governance. With the overriding principle that transitioned programs have a direct link to a basis in regulation or law. - Identify safety and compliance risks, and then bridge the transition of programs from the SONGS to the DGC such that identified risks are minimized and/or mitigated. - Allow the SONGS organization to transition from a Performance role to a primarily Oversight role as the DA. Working collaboratively with the DGC, programs were strategically grouped and sequenced for turnover. Central to the success of transition was fostering a collaborative environment with the DGC, facilitating knowledge transfer, applying schedule rigor, and keeping the team horizontally and vertically aligned to management expectations. SONGS achieved this alignment by embedding the formal process with management expectations for review and acceptance of proposed programs. SONGS' transitioned core programs to a general contractor and transition from a 'perform' organization to an 'oversight' organization. Lessons learned from this experience may benefit other nuclear generating station owners currently evaluating options for site D and D. (author)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Twenty years of dry interim storage of spent nuclear fuel (SNF) in Germany- History and developments - 15575

According to article 6 of the German Atomic Energy Act, storage of nuclear fuel requires a license granted by the competent authority in this field, which is the Federal Office of Radiation Protection. In 1983 a license to store spent nuclear fuel (SNF) and later expanded for vitrified high level waste (HAW) for the interim storage facility in Gorleben was granted. Dry Interim storage at the NPP sites in its actual form started in 2002 in the interim storage facility next to the NPP Lingen. Since this time each NPP erected its own storage facility and three central storage facilities have been built. All of these facilities use dry storage in metallic casks. The actual storage licenses for SNF and the HAW (which is stored in Gorleben and Rubenow) are limited to 40 years starting with the first cask emplacement. As no opening of the cask is intended during the storage period the condition of the inventory and the cask before loading has to be known in detail. This requires for instance detailed knowledge of the history of the SNF loaded as well as supervision of the manufacturing process of the cask. Within the last twenty years several mishaps influenced the regulatory point of view. During a loading campaign in a boiling water reactor traces of pool water remained in the gasket slot and during the drying procedure crystalline boron salts prevented the required tightness. After that event a long-term experiment was started and is still going on which studies the effects of water and minor damages to the gasket. As well changes in the loading procedure have been introduced. The pressure switch used to monitor the pressure in the space between primary and secondary lid has a self surveillance function. This indicated a malfunction 23 times in the last twenty years. Examinations of all events indicated no systematic failure, but a weak point seems to be the plug-in connection. The manufacturer introduced some modifications to improve the system. Introduction of FEM methods in safety analysis to amend the existing methods of testing allowed a new view on and in the cask safety analysis. Progress in this field is closely connected to the available computing capacity and stipulates experts to ongoing improvement. After expiring of the storage license which will be the case for the first cask loaded in 1992 and stored in Ahaus 2032 an extension or a new license can only be granted if the at this time relevant requirements are met. This will require detailed knowledge about the happenings and accidents during the previous storage period. All experience of the periodic safety assessment (PSA) which started in 2013 and accompanying experiments conducted till now and during the licensed storage time will be needed. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Major Achievements of the Springfields Decommissioning Program - 16512

Since the 1940's, Springfields has manufactured fuel products for the UK's nuclear power stations and for international customers. The early operations were in support of the atomic weapons program, closely followed by fuel manufacture for electricity generation via the Magnox power stations commencing in the late 1950's. In 2005, responsibility for the assets and liabilities of Springfields was transferred to the Nuclear Decommissioning Authority (NDA). A new company, Springfields Fuels Limited, was created to run the site, which was managed and operated by Westinghouse Electric UK Ltd on the NDA's behalf. On April 1, 2010, Westinghouse entered into an agreement with the NDA for a long-term lease of the Springfields site, which transferred responsibility for the commercial fuel manufacturing business and Springfields Fuels Limited to Westinghouse. Today, the main Springfields activities include: - Oxide fuels for Advanced Gas-cooled and Light Water Reactors, as well as intermediate fuel products, such as powders, granules and pellets, - Manufacture of Uranium Hexafluoride (now in Care and Maintenance), - Processing and recovery of Uranium residues, - Decommissioning and demolition of redundant plants and buildings. Some 120 plant buildings and support facilities have been successfully decommissioned and demolished in the period 1996 to date. These range from 1940's buildings, subsequently converted for nuclear fuel manufacture, together with some of the formerly sealed chemical agent manufacture and storage facilities. Decommissioning of the now redundant Magnox manufacturing facilities is ongoing at Springfields without compromising the other site operations. Westinghouse is managing all activities related to the whole lifecycle, planning, design, safety case, radiation protection, Post Operational Clean Out (POCO), service diversions, characterization, decommissioning all phases through to ground remediation, waste disposal and agreed end states. Decommissioning of process plants starts with the removal of the majority of radiological activity or hazard contained within the plant. This is known as Post Operational Clean Out. Once this is complete, the plant may be dismantled. Work is scoped and defined within packages (at plant or cell level) to aid control and commercial monitoring as the decommissioning is progressed. This package approach also enables clean or lightly contaminated items to be removed separately to aid potential re-cycling and minimize decontamination and waste disposal costs. Westinghouse is further optimising work package delivery by planning parallel POCO and Decommissioning work phases within plant at a cell level to systematically remove redundant plant and reduce hazard earlier, whilst optimising use of both internal and external resources to reduce total project cost. Waste and Residues from all phases are disposed of via authorized disposal routes and in accordance with our environmental authorizations. On completion of decommissioning and removal of plant and equipment the building is re-assessed radiologically, categorized and appropriate remediation work progressed to render it safe for conventional demolition. The major elements of the success achieved to date can be attributed to the following elements: - Overall programme management with extensive experience of planning, cost, and risk management. Integration of the programme within a major commercial site without compromising safety or operations. - Complete site EHS management working within a complex regulatory framework including the Nuclear Site Licenses, Environmental Authorisations and Security requirements. - Avoiding safety complacency by use of rigorous and systematic process to emphasise safety over competing goals. Including structured use of Human Performance and Corrective Action and Learning processes. - Unique safety experiences attained through complex hazard reduction across a broad range of uranium manufacturing and chemical hazard/agent facilities. - Project leadership and excellence utilizing a core site team to deploy repeatable processes and safe practices to deliver complex scope within budget and schedule. - Optimised use of internal and external resources, with extensive use of Operating Experience to drive continuous improvement and problem resolution. - Management of Contractors from initial selection, pre-qualification, tender assessment and Performance using varied contract forms across delivery phases. - Use of Westinghouse expertise for safety case, waste management, uranium recovery, health physics and design. - Implementation of a work package to ensure gated discreet manageable packages. Underpinned by package specific method statements, with appropriate hold points to maintain management and control of the works. - Continuous local safety focus via safety planning, induction, plant specific training, proactive risk assessment, and a defined approach to decoupling operating plant safety cases enabling focussed and usable documentation. This paper will provide an overview of the latest decommissioning activities achieved on the Springfields site and will focus on best practices that allowed Westinghouse to maintain the program on time and on budget. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Alpha Spectrometry Results for Groundwater Samples Collected in Northern Iraq and a Summary of the Environmental Setting of the Adaya Burial Site

The Radiation Protection Center (RPC) of the Iraqi Ministry of Environment continues to evaluate the potential health impacts associated with the Adaya Burial Site, which is located 33 kilometers (20.5 miles) southwest of Mosul. This report documents the radiological analyses of 16 groundwater samples collected from wells located in the vicinity of the Adaya Burial Site and at other sites in northern Iraq. The Adaya Burial Site is a high-risk dump site because a large volume of radioactive material and contaminated soil is located on an unsecure hillside above the village of Tall ar Ragrag. The uranium activities for the 16 water samples in northern Iraq are considered to be naturally occurring and do not indicate artificial (man-made) contamination. With one exception, the alpha spectrometry results for the 16 wells that were sampled in 2019 indicate that the water quality concerning the three uranium isotopes (Uranium-233/234, Uranium-235/236, and Uranium-238) was acceptable for potable purposes (drinking and cooking). However, Well 7 in Mosul had a Uranium-233/234 activity concentration that slightly exceeded the World Health Organization guidance level. Eight of the 16 wells are located in the villages of Tall ar Ragrag and Adaya and had naturally occurring uranium concentrations. Wells in the villages of Tall ar Ragrag and Adaya are located near the Adaya Burial Site and should be sampled on an annual schedule. The list of groundwater analytes should include metals, total uranium, isotopic uranium, gross alpha/beta, gamma spectroscopy, organic compounds, and standard water quality parameters. Our current understanding of the hydrogeologic setting in the vicinity of the Adaya Burial Site is solely based on villager's domestic wells, topographic maps, and satellite imagery. To better understand the hydrogeologic setting, a Groundwater Monitoring Program needs to be developed and should include the installation of twelve groundwater monitoring wells in the vicinity of Tall ar Ragrag and the Adaya Burial Site. Characterization of the limestone aquifer and overlying alluvium is needed. RPC should continue to support health assessments for the villagers in Tall ar Ragrag and Adaya. Collecting samples for surface water (storm water), airborne dust, vegetation, and washway sediment should be conducted on a routine basis. Human access to the Adaya Burial Site needs to be strictly limited. Livestock access on or near the burial site needs to be eliminated. The surface-water exposure pathway is likely a greater threat than the groundwater exposure pathway. Installation of a surface-water diversion or collection system is recommended in order to reduce the potential for humans and livestock to come in contact with contaminated water and sediment. To reduce exposure to villagers, groundwater treatment should be considered if elevated uranium or other contaminants are detected in drinking water. Installing water-treatment systems would likely be quicker to accomplish than remediation and excavation of the Adaya Burial Site. The known potential for human exposure to uranium and metals (such as arsenic, chromium, selenium, and strontium) at the Adaya Burial Site is serious. Additional characterization , mitigation, and remediation efforts should be given a high priority.

54 ENVIRONMENTAL SCIENCES↗

Gamma-ray dose from an overhead plume

Standard plume models can underestimate the gamma-ray dose when most of the radioactive material is above the heads of the receptors. Typically, a model is used to calculate the air concentration at the height of the receptor, and the dose is calculated by multiplying the air concentration by a concentration-to-dose conversion factor. Models indicate that if the plume is emitted from a stack during stable atmospheric conditions, the lower edges of the plume may not reach the ground, in which case both the ground-level concentration and the dose are usually reported as zero. However, in such cases, the dose from overhead gamma-emitting radionuclides may be substantial. Such underestimates could impact decision making in emergency situations. The Monte Carlo N-Particle code, MCNP, was used to calculate the overhead shine dose and to compare with standard plume models. At long distances and during unstable atmospheric conditions, the MCNP results agree with the standard models. As a result, at short distances, where many models calculate zero, the true dose (as modeled by MCNP) can be estimated with simple equations.

61 RADIATION PROTECTION AND DOSIMETRY↗

Dose Assessment of LANL-Derived Residual Radionuclides in Soils within C Tracts (C-2, C-3, and C-4) for Land Transfer Decisions

In 2015, soil sampling for radiological materials was conducted within Tract A-16-b specifically for land conveyance decisions. Measurements of radionuclides in soil samples were evaluated against a recreational use scenario, and all measurements were below screening action levels for each radionuclide. The total estimated dose was less than 1 mrem/y (< 10 μSv/y) for a hypothetical recreational user (compared to an action level of 15 mrem/y (150 μSv/y). Dose estimates were based on the 95% upper confidence limits for radionuclide concentrations within the tracts. Additionally, dose estimates less than 3 mrem/y are considered to be As Low As Reasonably Achievable, so no follow-up analysis was conducted. Release of this property is consistent with the requirements of DOE Order 458.1 and Los Alamos National Laboratory Policy 412.

54 ENVIRONMENTAL SCIENCES↗

Conceptual design of a fleet of autonomous regolith throwing devices for radiation shielding of lunar habitats

The National Aeronautics and Space Administration (NASA) in conjunction with Universities Space Research Association (USRA) has requested that the feasibility of a fleet of regolith tossing devices designed to cover a lunar habitat for radiation protection be demonstrated. The regolith, or lunar soil, protects the lunar habitat and its inhabitants from radiation. Ideally, the device will operate autonomously in the lunar environment. To prove the feasibility of throwing regolith on the Moon, throwing solutions were compared to traditional, Earth-based methods for moving soil. Various throwing configurations were investigated. A linear throwing motion combined with a spring and motor energizing system proved a superior solution. Three different overall configurations for the lunar device are presented. A single configuration is chosen and critical parameters such as operating procedure, system volume, mass, and power are developed. The report is divided into seven main sections. First, the Introduction section gives background information, defines the project requirements and the design criteria, and presents the methodology used for the completion of this design. Next, the Preliminary Analysis section presents background information on characteristics of lunar habitats and the lunar environment. Then, the Alternate Designs section presents alternate solutions to each of the critical functions of the device. Fourth, a detailed analysis of throwing the regolith is done to demonstrate its feasibility. Then, the three overall design configurations are presented. Next, a configuration is selected and the conceptual design is expanded to include system performance characteristics, size, and mass. Finally, the Conclusions and Recommendations for Future Work section evaluates the design, outlines the next step to be taken in the design process, and suggests possible goals for future design work.

Armstrong, Karem↗

Interpretation of TEPC Measurements in Space Flights for Radiation Monitoring

For the proper interpretation of radiation data measured in space, the results of integrated radiation transport models were compared with the tissue equivalent proportional counter (TEPC) measurements. TEPC is a simple, time-dependent approach to radiation monitoring for astronauts on board the International Space Station. Another and a newer approach to microdosimetry is the use of silicon-on-insulator (SOI) technology launched on the MidSTAR-1 mission in low Earth orbit (LEO). In the radiation protection practice, the average quality factor of a radiation field is defined as a function of linear energy transfer (LET), Qave(LET). However, TEPC measures the average quality factor as a function of the lineal energy y, Qave(y), defined as the average energy deposition in a volume divided by the average chord length of the volume. The deviation of y from LET is caused by energy straggling, delta-ray escape or entry, and nuclear fragments produced in the detector volume. The response distribution functions of the wall-less and walled TEPCs were calculated from Monte-Carlo track simulations. Using an integrated space radiation model (which includes the transport codes HZETRN and BRYNTRN, and the quantum nuclear interaction model QMSFRG) and the resultant response distribution functions from Monte-Carlo track simulations, we compared model calculations with the walled-TEPC measurements from NASA missions in LEO and made predictions for the lunar and the Mars missions. Good agreement was found for Qave(y) between the model and measured spectra from past NASA missions. The Qave(y) values for the trapped or the solar protons ranged from 1.9-2.5. This over-estimates the Qave(LET) values which ranged from 1.4-1.6. Both quantities increase with shield thickness due to nuclear fragmentation. The Qave(LET) for the complete GCR spectra was found to be 3.5-4.5, while flight TEPCs measured 2.9-3.4 for Qave(y). The GCR values are decreasing with the shield thickness. Our analysis of the measurements of TEPCs can be used for a proper interpretation of observed data of monitoring the space radiation environment.

Kim, Myung-Hee Y.↗

Comparison of Integrated Radiation Transport Models with TEPC Measurements for the Average Quality Factors in Spaceflights

The purpose of this work is to test our theoretical model for the interpretation of radiation data measured in space. During the space missions astronauts are exposed to the complex field of radiation type and kinetic energies from galactic cosmic rays (GCR), trapped protons, and sometimes solar particle events (SPEs). The tissue equivalent proportional counter (TEPC) is a simple time-dependent approach for radiation monitoring for astronauts on board the International Space Station. Another and a newer approach to Microdosimetry is the use of silicon-on-insulator (SOI) technology launched on the MidSTAR-1 mission in low Earth orbit (LEO). In the radiation protection practice, the average quality factor of a radiation field is defined as a function of linear energy transfer (LET), Q(sub ave)(LET). However, TEPC measures the average quality factor as a function of the lineal energy y, Q(sub ave)(y), defined as the average energy deposition in a volume divided by the average chord length of the volume. Lineal energy, y, deviates from LET due to energy straggling, delta-ray escape or entry, and nuclear fragments produced in the detector volume. Monte Carlo track structure simulation was employed to obtain the response of a TEPC irradiated with charged particle for an equivalent site diameter of 1 micron of wall-less counter. The calculated data of the energy absorption in the wall-less counter were compiled for various y values for several ion types at various discrete projectile energy levels. For the simulation of TEPC response from the mixed radiation environments inside a spacecraft, such as, Space Shuttle and International Space Station, the complete microdosimetric TEPC response, f( y, E, Z), were calculated with the Monte Carlo theoretical results by using the first order Lagrangian interpolation for a monovariate function at a given y value (y = 0.1 keV/micron 5000 keV/micron) at any projectile energy level (E = 0.01 MeV/u to 50,000 MeV/u) of each specific radiation type (Z = 1 to 28). Because the anomalous response has been observed at large event sizes in the experiment due to the escape of energy out of sensitive volume by delta-rays and the entry of delta-rays from the high-density wall into the low-density gas-volume cavity, Monte Carlo simulation was also made for the response of a walled-TEPC with wall thickness 2 mm and density 1 g/cm(exp 3). The radius of cavity was set to 6.35 mm and a gas density 7.874 x 10(exp -5) g/cm(exp 3). The response of the walled- and the wall-less counters were compared. The average quality factor Q(sub ave)(y) for trapped protons on STS-89 demonstrated the good agreement between the model calculations and flight TEPC data as shown. Using an integrated space radiation model (this includes the transport codes HZETRN and BRYNTRN, the quantum nuclear interaction model QMSFRG) and the resultant response distribution functions of walled-TEPC from Monte-Carlo track simulations, we compared model calculations with walled-TEPC measurements from NASA missions in LEO and made predictions for the lunar and the Mars missions. The Q(sub ave)(y) values for the trapped or the solar protons ranged from 1.9-2.5. This over-estimates the Qave(LET) values which ranged from 1.4-1.6. Both quantities increase with shield thickness due to nuclear fragmentation. The Q(sub ave)(LET) for the complete GCR spectra was found to be 3.5-4.5, while flight TEPCs measured 2.9-3.4 for Q(sub ave)(y). The GCR values are decreasing with the shield thickness. Our analysis for a proper interpretation of data supports the use of TEPCs for monitoring space radiation environment.

Kim, Myung-Hee Y.↗

Recent Low-Earth Orbit Radiation Measurements Using Passive Dosimeters during ISS Expeditions 13, 14 and 15

Current passive radiation measurements in low-Earth orbit (LEO) involve using a combination of thermally/optically stimulated luminescence detectors (TLDs/OSLDs) for the low-LET component (LET<10 keV/micron water) of the space radiation field and plastic nuclear track detectors (PNTDs) for the high-LET region (LET>10 keV/micron water), as per National Council on Radiation Protection recommendation (NCRP 132, 2004). This combination of radiation detectors has been successfully implemented at NASA Johnson Space Center for routine crewmember and ISS area monitoring using LiF:Mg,Ti, CaF2:Tm, Al2O3:C and CR-39 material. This paper will present in detail the measurement and data processing techniques employed for passive space radiation measurements by the Space Radiation Analysis Group (SRAG) at JSC. The paper will also summarize the absorbed dose, dose equivalent and quality factor results for the area monitoring during ISS Expeditions 13, 14 and 15.

Gaza, R.↗

Comparisons of Integrated Radiation Transport Models with Microdosimetry Data in Spaceflight

Astronauts are exposed to galactic cosmic rays (GCR), trapped protons, and possible solar particle events (SPE) during spaceflight. For such complicated mixtures of radiation types and kinetic energies, tissue equivalent proportional counters (TEPC's) represent a simple time-dependent approach for radiation monitoring. Of interest in radiation protection is the average quality factor of a radiation field defined as a function of linear energy transfer, LET, Q(sub ave)(LET). However TEPC's measure the average quality factors as a function of lineal energy (y), Q(sub ave)(y) defined as the average energy deposition in a volume divided by the average chord length of the volume. Lineal energy, y deviates from LET due to energy straggling, delta-ray escape or entry, and nuclear fragments produced in the detector. Using integrated space radiation models that includes the transport code HZETRN/BRYNTRN, the quantum nuclear interaction model, QMSFRG, and results from Monte-Carlo track simulations of TEPC's response to ions, we consider comparisons of model calculations to TEPC results from NASA missions in low Earth orbit and make predictions for lunar and Mars missions. Good agreement between the model and measured spectra from past NASA missions is found. A finding of this work is that TEPC's values for trapped or solar protons of Q(sub ave)(y) range from 1.9-2.5, overestimating Q(sub ave)(LET), which ranges from 1.4-1.6 with both quantities increasing with shielding depth due to nuclear secondaries Comparisons for the complete GCR spectra show that Q(sub ave)(LET) for GCR is approximately 3.5-4.5, while TEPC's measure 2.9-3.4 for Q(sub ave)(y) with the GCR values decreasing with depth as heavy ions are absorbed in shielding material. Our results support the use of TEPC's for space radiation environmental monitoring when computational analysis is used for proper data interpretation.

Cucinotta, Francis A.↗

Analysis of Chromosomal Aberrations after Low and High Dose Rate Gamma Irradiation in ATM or NBS Suppressed Human Fibroblast Cells

A detailed understanding of the biological effects of heavy nuclei is needed for space radiation protection and for cancer therapy. High-LET radiation produces more complex DNA lesions that may be non-repairable or that may require additional processing steps compared to endogenous DSBs, increasing the possibility of misrepair. Interplay between radiation sensitivity, dose, and radiation quality has not been studied extensively. Previously we studied chromosome aberrations induced by low- and high- LET radiation in several cell lines deficient in ATM (ataxia telangactasia mutated; product of the gene that is mutated in ataxia telangiectasia patients) or NBS (nibrin; product of the gene mutated in the Nijmegen breakage syndrome), and gliomablastoma cells that are proficient or lacking in DNA-dependent protein kinase (DNA-PK) activity. We found that the yields of both simple and complex chromosomal aberrations were significantly increased in the DSB repair defective cells compared to normal cells. The increased aberrations observed for the ATM and NBS defective lines was due to a significantly larger quadratic dose-response term compared to normal fibroblasts for both simple and complex aberrations, while the linear dose-response term was significantly higher in NBS cells only for simple exchanges. These results point to the importance of the functions of ATM and NBS in chromatin modifications that function to facilitate correct DSB repair and minimize aberration formation. To further understand the sensitivity differences that were observed in ATM and NBS deficient cells, in this study, chromosomal aberration analysis was performed in normal lung fibroblast cells treated with KU-55933, a specific ATM kinase inhibitor, or Mirin, an MRN complex inhibitor involved in activation of ATM. We are also testing siRNA knockdown of these proteins. Normal and ATM or NBS suppressed cells were irradiated with gamma-rays and chromosomes were collected with a premature chromosome condensation (PCC) technique at the first mitosis post-irradiation. Chromosomes were analyzed using a multicolor fluorescence in-situ hybridization (mFISH) chromosome painting method. Preliminary analysis showed that chromosomal exchanges were increased in the cells treated with the specific ATM inhibitor. Possible cytogenetic signatures of acute and low dose-rate gamma irradiation in ATM or Nibrin deficient and suppressed cells will be discussed.

Hada, M.↗