Offsite Dose Research Articles

A Consequence-Informed Licensing Path Selection for the Design of Physical Protection Systems at Commercial Nuclear Power Facilities

The survivability of the domestic nuclear power industry depends on the cost-competitiveness of safe and secure nuclear power generation. Advanced reactor design concepts aim to have increased safety margins over traditional large light water reactors (LWRs). With increased safety margins comes the potential for a corresponding decrease in off-site risk to the general public from a hypothetical release of radioactivity due to sabotage or theft. Without sacrificing safety or security, advanced reactor designers may be able to achieve operational cost improvements over current LWRs in part by designing less burdensome physical protection systems (PPSs) and by replacing on-site response forces with off-site response forces. To accommodate these developments, the U.S. Nuclear Regulatory Commission is drafting new rulemakings for physical security when licensing through the current frameworks in 10 CFR 50 or 10 CFR 52 along with drafting an entirely new licensing framework: 10 CFR 53. A novel technology-inclusive consequence-informed methodology for the selection of the optimal licensing path for the design of PPSs at advanced fixed-site commercial nuclear power facilities is presented herein. This methodology proposes integrating security considerations at the beginning of a reactor facility design effort to streamline the licensing process. Off-site total effective dose equivalents at the exclusion area and low population zone boundaries were identified as the key metrics when determining a design’s most appropriate licensing path that in turn affects the design requirements placed upon the PPS. Given these metrics, source-term generation of potential adversary-induced physics-based sabotage actions utilizing severe accident modeling software and off-site plume dispersal modeling were identified as appropriate for determining siting constraints, potential target sets for hypothetical sabotage events, and their subsequent off-site dose consequences. The methodology proposes using the consequence results from the sabotage modeling, in combination with desired cost-saving PPS characteristics, to help inform the licensing path selection. Once a licensing path is chosen, the methodology utilizes the Design and Evaluation Process Outline to evaluate an effective PPS following the licensing requirements placed on the facility. This paper also presents examples of hypothetical commercial nuclear power facilities with varying consequence levels and demonstrations of how to select the optimal licensing pathways for each.

Monte Carlo Simulation of Prompt Dose in the National Ignition Facility

The National Ignition Facility (NIF) at Lawrence Livermore National Laboratory is the world’s largest and most powerful laser system for inertial confinement fusion. The NIF is a 192-laser beam facility that produces more than 2 MJ and 500 TW of ultraviolet light. The facility is expected to generate shots with varying fusion yield (up to 20 MJ or 7.1x1018 neutrons per shot) and a maximum annual yield of 1200 MJ. Monte Carlo simulations have been performed to estimate prompt dose values (neutrons and photons) inside the facility as well as at different locations outside the facility shield walls. The Target Chamber shielding, along with Target Bay and Switchyard walls, roofs, and doors will reduce exposures in occupied areas to acceptable levels. The thicknesses of the different shield doors in the facility were optimized such that prompt dose in occupied areas is limited to < 50 μSv per 20 MJ shot and < 1 mSv for 1200 MJ of annual fusion yield. Most of the normally occupied areas experience limited prompt dose values of < 50 μSv per shot. Low dose values are also expected outside the NIF facility. The maximum annual off-site dose to a maximally exposed individual is small and estimated at 24 μSv for 1200 MJ of fusion yield.

Offsite Dose Assessment to the Public from Residual Radioactivity due to Mining Activities in Adamawa State, Nigeria.

Offsite Dose Assessment to the Public from Residual Radioactivity due to Mining Activities in Adamawa State, Nigeria.

Derivation of a new dose constraint applicable to radioactive discharges from Korean nuclear power plants through retrospective dose assessment

A new methodology to derive a dose constraint for radioactive effluent from a unit of nuclear power plant (NPP) through retrospective assessment was developed to reflect operational flexibility in line with international standards. The new dose constraint can retain the safety margin between the offsite dose and the past dose constraints. As case studies, the new approach was applied to 24 Korean NPPs to address the limitations of the existing seven dose constraints that do not fully comply with current international radiation protection standards. Therefore, an effective dose constraint for Korean NPPs was proposed as no less than 0.15 mSv/y, which is comparable to the international practices and previous studies (0.05–0.3 mSv/y). Although the lower bound of the equivalent dose constraint was calculated as 0.17 mSv/y, it is not proposed in this study since the compliance with the derived effective dose constraint can prevent accompanied equivalent doses to any organs from exceeding equivalent dose limits. The new framework and the case studies are expected to contribute toward and support the revision of existing dose constraints for radioactive effluent from NPPs, ensuring better compliance with the current international safety standards as well as reflect the operational flexibility in practice.

The establishment of analysis methodology of NRCDose3 for Kuosheng nuclear power plant decommissioning

Abstract In this research, the analysis methodology for Kuosheng BWR/6 nuclear power plant (NPP) decommissioning is established by using NRCDose3 code. NRCDose3 code includes XOQDOQ, GASPAR II, and LADTAP II programs that can assess atmospheric dispersion factor, gaseous radioactive effluents and offsite doses, and liquid radioactive releases and offsite doses, respectively. Therefore, this analysis methodology of NRCDose3 assesses the offsite doses of the gaseous radioactive effluents and liquid radioactive releases for Kuosheng NPP decommissioning. According to the analysis results, the results of NRCDose3 are consistent with Kuosheng NPP data.

Monitoring of helium gas leakage from canister storing spent nuclear fuel: Radiological consequences and management

Chloride-induced stress corrosion cracking (CISCC) of canisters has been identified as a potential degradation mechanism requiring ageing management — especially in a marine air environment. Methods for monitoring helium gas leakage from canisters have been developed on the basis of a phenomenon that the bottom temperature of the canister increases as the top temperature of the canister decreases when helium gas leaks during storage. In this paper, we consider the radiological consequences of CISCC based on two scenarios of helium gas leakage and propose effective management that includes a timeline for monitoring and mitigation. Monitoring helium gas leakage by measuring the difference in temperature between canister surfaces in the dry cask storage of spent fuel (SF) can detect leakage before SF undergoes oxidation as a result of air ingress and before country-specific regulatory limits of off-site doses are exceeded, thereby enabling mitigatory action.

POTENTIAL ENVIRONMENTAL HAZARD TO THE PUBLIC FROM THE OPERATION OF URANIUM MINING AND MILLING FACILITY.

Mining and milling processes could cause potential radiological exposures to the public. The objective of this study was to estimate the off-site radiological doses expected to be received by the public as a result of uranium mining and milling activities at Mkuju River Project in the United Republic of Tanzania. MILDOS-AREA was used to estimate off-site doses along with RESRAD-OFFSITE for comparison and verification. Since the mining has not yet started, the conceptual scenario was chosen. Using the concept of the most exposed individual, the location of the nearest residence (receptor) was chosen at 2.5km from the site with other receptors being the fence boundary and grazing area being at 1.0 and 1.8km, respectively. Yellowcake stack (point source), ore pad and two tailing piles where each had an area of 2.5×105m2 were chosen to be the source of radiological contamination. The radiological source term was obtained from the concentration of 226Ra and 232Th in soil obtained from the previous studies of environmental impact assessments. Meteorological and site-specific data were used for this analysis. The estimated total effective dose equivalent (TEDE) for the nearest residence which was calculated by MILDOSE-AREA ranged from 2.5×10-2 to 4.45×10-2mSv/y during the operation of 13y. The result of RESRAD-OFFSITE ranged from 7.19×10-2mSv/y for the first year to 7.43×10-2mSv/y in the final year. This implied all the estimated TEDEs were below the dose limit and dose constraint of 1 and 0.3mSv/y, respectively, as suggested by the International Atomic Energy Agency. Hence, it was found that there was no potential radiological concern of uranium mining at Mkuju River Project. It was found that using MILDOS-AREA it is possible to estimate dose at different distances from the facility. Therefore, this study apart from estimating the off-site doses, it can be used for planning of public and social premises before the commencement of the project. That is the distance from the facility where the public should be located as well as other locations for social activities.

An online operator support tool for severe accident management in nuclear power plants using dynamic event trees and deep learning

Operating staffs of a nuclear power plant (NPP) are responsible for returning the NPP to a stable state and alert authorities if there is the potential for offsite radiological consequences following an accident. An operator support tool (OST) using deep learning techniques and trained by data from dynamic probabilistic safety/risk assessment (DPSA/DPRA) is proposed to assist the NPP personnel in decision-making. The DPSA/DPRA methodology employs time-dependent branching conditions based on the evolving state of the NPP and accounts for complex hardware/process/software/human interactions to predict possible outcomes of the initiating event. A large number of scenarios generated from the DPSA/DPRA performed for a pressurized water reactor station blackout as a function of time were used to train the OST to predict possible offsite dose outcomes at 2-mile and 10-mile site boundaries for emergency response planning. The results show that the OST can predict offsite dose levels with greater than 90% accuracy.

Performance Improvement of Liquid Waste Management System for APR1400

The Liquid Waste Management System (LWMS) is a radioactive waste storage and treatment facility equipped in the nuclear power plants to ensure the safe management of radioactive waste generated in the process of operation. New and improved materials and processes are continuously under consideration and development in various countries to reduce waste generation and to minimize costs. To meet the increasing pressure to reduce the release of radioactive into the environment, this study discusses the possible performance improvements of the existing liquid waste management system of APR1400 without additional equipment installation or design changes of the system. This study proposes three possible items for improvement of the Advanced Power Reactor 1400 (APR1400) liquid waste management system which are chemical additions, increasing ion exchange (I.X.) selectivity and optimal operation conditions to reduce the radioactivity concentration in the effluent and solid waste generation. The results of off-site dose calculated by ENDOS-L shows that maximum individual dose is reduced by 12.7% from original values. And the reduction of solid waste generation from the extended replacement period of the membrane is expected to reduce the annual disposal cost by 16.7% from original cost. Improvements do not involve equipment modification so these can be easily applied to currently operating NPPs. It will be helpful to reduce the offsite public dose and operating cost of Nuclear Power Plants(NPPs).

Estimate of radiation release from MIT reactor with un-finned LEU core during Maximum Hypothetical Accident

The Massachusetts Institute of Technology Reactor (MITR-II) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. At 6 MW, it delivers neutron flux and energy spectrum comparable to light water reactor (LWR) power reactors in a compact core using highly enriched uranium (HEU) fuel. In the framework of non-proliferation policy, the international community aims to minimize the use of HEU in civilian facilities. Within this context, research and test reactors have started a program to convert HEU fuel to low enriched uranium (LEU) fuel. A new type of LEU fuel based on a high density alloy of uranium and molybdenum (U-10Mo) is expected to allow the conversion of U.S. domestic high performance reactors like MITR. The current study focuses on the impacts of MITR Maximum Hypothetical Accident (MHA), which is also the Design Basis Accident (DBA), with LEU fuel. The MHA for the MITR is postulated to be a coolant flow blockage in the fuel element that contains the hottest fuel plate. It is assumed that the entire active portion of five fuel plates melts. The analysis shows that, within a 2-h period and by considering all the possible radiation sources and dose pathways, the overall off-site dose is 302.1 mrem (1 rem = 0.01 Sv) Total Effective Dose Equivalent (TEDE) at 8 m exclusion area boundary (EAB) and a higher dose of 392.8 mrem TEDE is found at 21 m EAB. In all cases the dose remains below the 500 mrem total TEDE limit goal based on NUREG-1537 guidelines.

Consequence Assessment of the WIPP Radiological Release from February 2014

On 14 February 2014, a continuous air monitor (CAM) alarm at the exit of panel 7 in the Waste Isolation Pilot Plant (WIPP) underground facility caused the mine ventilation to shift from unfiltered air over to HEPA filtration for its effluent. Subsequent measurements of the effluent at both pre- and post-HEPA filtration using representative sampling demonstrated that a release had occurred. Using modeling based on measured effluent activity, onsite dose estimates were calculated and later measured via bioassay to be less than 0,1 mSv from intakes of radioactivity. The maximum offsite dose potential to nearby dwellings was modeled to be 1 µSv or less, which was consistent with air samples being taken at those dwellings during the release, demonstrating impressive accuracy and precision. No worker or public dose limits have been exceeded, and the release was substantially below the annual release limits for the WIPP site.

Evaluation of the offsite dose contribution to the global risk in a Steam Generator Tube Rupture scenario

The current main figure of merit for risk based decision making process based on Probabilistic Safety Assessment level 1 is usually related with the fuel failure (i.e., Peak Cladding Temperature (PCT)>1477.15K). In this approach, the core damage is the first and necessary step in a potential radiological release, being the containment failure the second one. Nevertheless, SGTR sequences in PWR plants are able to release large quantities of radioactive products without previous core damage or containment failure. For that reason, it seems necessary to analyze which sequences exceed the allowed offsite dose criteria prior to the core damage criterion.The aim of this analysis has been to evaluate the risk contribution due to the offsite dose and the core damage in case of Steam Generator Tube Rupture (SGTR) sequences at full power in a 3-loop Pressurized Water Reactor (PWR) Westinghouse-design. The study has been performed with SCAIS/MAAP and RADTRAD codes. For that purpose, this analysis unfolds the SGTR Dynamic Event Tree for both the core damage and the offsite dose risk metrics. The results indicate that dose criteria complement the PCT criterion and allow quantifying both risk contributions in SGTR sequences.

Environmental impact assessment of a new nuclear power plant (NPP) based on atmospheric dispersion modeling

The US Environmental Protection Agency’s short-range atmospheric dispersion model (AERMOD 12345) is a good candidate for radiological dose calculations to the general public and the environment. It advanced capability should provide better confidence in the accuracy of offsite dose and risk assessment. The code has been used to compute the scaling factors for air concentration and ground level deposition of fission products based on routine and hypothetical accident releases from the NPP site in Geregu, Nigeria (7°33′N, 6°41′E). All computations were within the 16 km emergency planing zone of a generic reactor design considered by the study. The scaling factors have been used to assess the potential environmental risk of the NPP using an integrated approach to the assessment and management of environmental risks from ionizing radiation (D-ERICA). Obtained results should provide baseline information for decision making in terms of operation license provision for the pioneer NPP.

Severe accident source terms for a sodium-cooled fast reactor

In order to support the demonstration of a risk-informed approach to the design optimization of a sodium-cooled fast reactor (SFR), it was necessary to make realistic estimates of the consequences of severe accident scenarios. This paper describes the database, models, and assumptions used to estimate the offsite consequences of characteristic severe accident scenarios. As required for comparison with the NRC’s technology neutral framework limit curve, the offsite dose at one mile from the plant boundary is calculated using conservative meteorology.The reference plant design is a 1000MWt pool-type design with metallic fuel. Because an integrated analysis tool comparable to MELCOR does not exist for SFR accident scenario analysis, it was necessary to write a computer code that would assess release of radionuclides from the fuel and transport within the reactor primary system and to link those analyses with results from existing computer codes that assess the dynamic response of the reactor, containment thermal–hydraulics, and radionuclide transport processes within the containment. The analyses indicate that the offsite source terms for SFR severe accident scenarios tend to be small because of the low melting temperature of the fuel, likelihood of significant retention of fission products within the sodium pool, augmentation of containment deposition processes by interaction with sodium oxide aerosols, and small driving force for release from the containment to the environment. A number of major sources of modeling uncertainty are identified as requiring further development effort. An integrated modeling capability, similar to the MELCOR code, is required to direct further research and ultimately to support the licensing of an SFR.

Evaluation and Measurements of Radioactive Air Emission and Off-Site Doses at SLAC

SLAC, a high-energy (GeV) electron accelerator facility, performs experimental and theoretical research using high-energy electron and/or positron beams that can produce secondary neutron and gamma radiation when beam losses occur. Radioactive gas production (mainly C, N, O, Ar) and release is one of the environmental protection program issues. U.S. DOE Order 458.1 requires that 40 CFR 61 Subpart H's NESHAP requirements be followed. These regulations prescribe a total dose limit of 0.1 mSv y to the Maximally Exposed Individual (MEI) of the general public, a requirement for a continuous air monitoring system if a release point within a facility can cause > 1 × 10 mSv y to the MEI, and a requirement for periodic confirmatory measurements for minor sources which give releases that contribute ≤ 1 × 10 mSv y to the MEI. At SLAC, all air release points for current operations are evaluated to be minor sources. This paper describes SLAC's evaluation following NESHAP requirements; measurements using the Air Monitoring Station (AMS) as periodic confirmatory measurements are also discussed.

Analysis of the operator action and the single failure criteria in a SGTR sequence using best estimate assumptions with TRACE 5.0

Steam Generator Tube Rupture (SGTR) sequences in Pressurized Water Reactors are known to be one of the most demanding transients for the operating crew. SGTR are a special kind of transient as they could lead to radiological releases without core damage or containment failure, as they can constitute a direct path from the reactor coolant system to the environment.The first methodology used to perform the Deterministic Safety Analysis (DSA) of a SGTR did not credit the operator action for the first 30min of the transient, assuming that the operating crew was able to stop the primary to secondary leakage within that period of time. However, the different real SGTR accident cases happened in the USA and over the world demonstrated that the operators usually take more than 30min to stop the leakage in actual sequences. Some methodologies were raised to overcome that fact, considering operator actions from the beginning of the transient, as it is done in Probabilistic Safety Analysis.This paper presents the results of comparing different assumptions regarding the single failure criteria and the operator action taken from the most common methodologies included in the different Deterministic Safety Analysis. One single failure criteria that has not been analysed previously in the literature is proposed and analysed in this paper too. The comparison is done with a PWR Westinghouse three loop model in TRACE code (Almaraz NPP) with best estimate assumptions but including deterministic hypothesis such as single failure criteria or loss of offsite power. The behaviour of the reactor is quite diverse depending on the different assumptions made regarding the operator actions. On the other hand, although there are high conservatisms included in the hypothesis, as the single failure criteria, all the results are quite far from the regulatory limits. In addition, some improvements to the Emergency Operating Procedures to minimize the offsite release from the damaged SG in case of a SGTR are outlined taking into account the offsite dose sensitivity results.

Assessment of radiation dose due to the accidental release of radionuclides from a DCLL reactor

A conceptual design for a DEMO fusion reactor based on a dual coolant He/Pb15.7Li breeding blanket (DCLL) is being developed within the Spanish Breeding Blanket Technology Programme: TECNO_FUS. The production of tritium and activation products of LiPb might be a concern from the radiological safety point of view. Thus, in this contribution, an accidental release in atmosphere of radionuclides from LiPb breeder has been studied.Activation calculations have been performed with ACAB code assuming an irradiation scenario of 5 FPY for the maximum neutron fluence rate in the equatorial breeding zone. The results in terms of specific activity, surface gamma dose rate and committed effective dose (CED) due to inhalation at different times have been used to chose the potentially more hazardous radionuclides.Dispersion of the selected radionuclides has been modeled with HOTSPOT code using the Gaussian plume model and two different atmospheric conditions. Offsite dose (for external irradiation and inhalation) due to an accidental release of 1kg of activated LiPb has been calculated after 5 FPY of irradiation (shutdown) using HOTSPOT atmospheric dispersion in class D weather conditions. According to the results, fulfilling the dose requirement for no evacuation would permit to release up to 40kg of activated LiPb, without taking into account the possible isotopic purification and detritiation systems. This value can be compared with the actually released amount in a given accidental scenario for designing the reactor confinement barriers and safety systems.

Cardiac Safety of Fingolimod 0.5 mg during the First Dose Observation in 4-Month, Open-Label, Multi-Center FIRST Study in Patients with Relapsing MS (S41.003)

Objective: To report the safety results of fingolimod 0.5 mg focusing on cardiac events during the first dose observation in relapsing MS patients both without cardiac risk factors and with cardiac conditions not included in fingolimod clinical trials. Background Fingolimod (FTY720) is an S1PR modulator approved for the treatment of relapsing MS. The initiation of fingolimod treatment results in generally asymptomatic and transient bradycardia and may be associated with atrioventricular blocks (AVB). Prescribing information recommends 6 hours on-site observation for signs and symptoms of bradycardia. Design/Methods: 2417 patients were enrolled. Preliminary results are reported for 2289 patients. Unless patients met pre-defined cardiac risk criteria the first dose administration was permitted off-site. Heart rate and ECG abnormalities were assessed by ambulatory Holter-ECG for 24 hours at screening and for 6 hours after administration of the first dose. Results: In the post-dose Holter ECGs the incidence of Mobitz I second degree AVBs was 1.4% and the incidence of 2:1 AVBs was 0.5%. One patient had a finding of >3 sec pause in both screening and post dose Holter-ECG. Incidence of conduction abnormalities was similar in patients monitored on-site (n=1162) and off-site (n=1127) and in patients with or without cardiac risks. One patient discontinued study drug due to the second degree AVB. No patient had a HR Conclusions: Overall incidence of AVBs on Holter-ECGs was low, irrespective of on- or off-site first dose administration or the presence of potential cardiac risk factors. These results are similar to those observed in fingolimod trials and confirm the benign nature of the cardiac effects of treatment initiation with fingolimod. Supported by: Novartis Pharma AG, Basel, Switzerland. Disclosure: Dr. Comi has received personal compensation for activities with Novartis, Teva Neuroscience, Sanofi-Aventis Pharmaceuticals, Inc., Merck Serono, Bayer Schering, and Biogen Dompe. Dr. Kappos has received research support from Acorda Therapeutics, Actelion, Allozyne, BaroFold, Inc., Bayer Pharmaceuticals Corporation, Bayhill Therapeutics, Biogen Idec, Boehringer Ingelheim Pharmaceuticals, Inc, Elan Corporation, Genmab, GlaxoSmithKline, Inc., Glenmark Pharma, Merck Serono, MediciNova, Novartis, Sanofi-Aventis Pharmaceuticals, Santhera Pharmaceuticals, Shire, Roche Diagnostics, Teva Neuroscience, UCB Pharma, Pfizer Inc, Swiss MS Society, Swiss National Research Foundation, European Union, Gianni Rubatto Foundation, Novartis and Roche Research Foundations. Dr. Palace has received personal compensation for activities with Teva Neuroscience, Biogen Idec, Merck Serono, and Bayer Schering Pharma. Dr. Palace has received research support from GlaxoSmithKline, Inc. Dr. Gottschalk has received personal compensation for activities with Novartis as an employee. Dr. Bijarnia has received personal compensation for activities with Novartis Healthcare Pvt Ltd as an employee. Dr. Keil has received personal compensation for activities with Novartis as an employee. Dr. Tomic has received personal compensation for activities with Novartis as an employee. Dr. von Rosenstiel has received personal compensation for activities with Novartis Pharma AG. Dr. von Rosenstiel holds stock and/or stock options in Novartis Pharma AG, which sponsored research in which Dr. von Rosenstiel was involved as an investigator. Dr. Gold has received personal compensation for activities with Bayer Pharmaceuticals Corporation, Biogen Idec, Merck Serono, Teva Neuroscience. Dr. Gold has received personal compensation in an editorial capacity for Therapeutic Advances in Neurological Disorders. Dr. Gold has received (royalty or license fee or contractual rights) payments from Biogen Idec. Dr. Gold has received research support from Bayer Pharmaceuticals Corporation, Biogen Idec, Merck Serono, Novartis and Teva Neuroscience.

Introducing PCTRAN as an evaluation tool for nuclear power plant emergency responses

Protecting the public from radiation exposure is important if a nuclear power plant (NPP) accident occurs. Deciding appropriate protective actions in a timely and effective manner can be fulfilled by using an effective accident evaluation tool. In our earlier work, we have integrated PCTRAN (Personal Computer Transient Analyzer) with the off-site dose calculation model. In this study, we introduce PCTRAN as an evaluation tool for nuclear power plant emergency responses. If abnormal conditions in the plant are monitored or observed, the plant staffs can distinguish accident/incident initiation events. Thus, the responsible personnel can immediately operate PCTRAN and set up those accident/incident initiation events to simulate the nuclear power plant transient or accident in conjunction with off-site dose distributions. The evaluation results consequently help the responsible organizations decide the rescue and protective actions. In this study, we explain and demonstrate the capabilities of PCTRAN for nuclear emergency responses, through applying it to simulate the postulated nuclear power plant accident scenarios.

Radiological risk from consuming fish and wildlife to Native Americans on the Hanford Site (USA)

Historical operations at the Hanford Site (Washington State, USA) have released a wide array of non-radionuclide and radionuclide contaminants into the environment. As a result of stakeholder concerns, Native American exposure scenarios have been integrated into Hanford risk assessments. Because its contribution to radiological risk to Native Americans is culturally and geographically specific but quantitatively uncertain, a fish and wildlife ingestion pathway was examined in this study. Adult consumption rates were derived from 20 Native American scenarios (based on 12 studies) at Hanford, and tissue concentrations of key radionuclides in fish, game birds, and game mammals were compiled from the Hanford Environmental Information System (HEIS) database for a recent time interval (1995-2007) during the post-operational period. It was assumed that skeletal muscle comprised 90% of intake, while other tissues accounted for the remainder. Acknowledging data gaps, median concentrations of eight radionuclides (i.e., Co-60, Cs-137, Sr-90, Tc-99, U-234, U-238, Pu-238, and Pu-239/240) in skeletal muscle and other tissues were below 0.01 and 1 pCi/g wet wt, respectively. These radionuclide concentrations were not significantly different (Bonferroni P>0.05) on and off the Hanford Site. Despite no observed difference between onsite and offsite tissue concentrations, radiation dose and risk were calculated for the fish and wildlife ingestion pathway using onsite data. With median consumption rates and radionuclide tissue concentrations, skeletal muscle provided 42% of the dose, while other tissues (primarily bone and carcass) accounted for 58%. In terms of biota, fish ingestion was the largest contributor to dose (64%). Among radionuclides, Sr-90 was dominant, accounting for 47% of the dose. At median intake and radionuclide levels, estimated annual dose (0.36 mrem/yr) was below a dose limit of 15 mrem/yr recommended by the United States Environmental Protection Agency (USEPA), as well as below a dose limit of 100 mrem/yr proposed by the International Commission on Radiation Protection (ICRP). Similarly, lifetime cancer risk (1.7E-5), calculated with median inputs, was below risk levels corresponding to these dose limits. However, our dose and risk estimates apply to only one pathway within a multidimensional exposure scenario for Native Americans. On the other hand, radiation dose and risk corresponding to onsite tissue concentrations were not significantly different from those corresponding to offsite (background) concentrations. Recognizing uncertainties in exposure and toxicity assessments, our results may facilitate informed decision making and optimize resource allocation within a risk assessment framework at the Hanford Site.