NRC Regulatory Guide Research Articles

Age-dependent Radiation Dose Rates from Canine Sn-117m Treatments.

Tin-117 m (Sn-117m) is used to treat dogs with osteoarthritic joints by radiosynoviorthesis. The decay process for Sn-117m is internal conversion wherein IC electrons and auger electrons provide the therapeutic effect. Additionally, the most prominent gamma emission is 158.6 keV. The effective dose rate received by a person interacting at close distances with a treated dog is needed to determine the person's total dose and thus regulatory compliance. Simple measurement of the dose rate at a given distance does not provide an accurate measurement of the effective dose to a person due to the non-uniform nature of the radiation field at close distances. MNCP models of the interactions of five ages of humans at three distances were created to determine the effective dose rates using the methodology from NRC Regulatory Guide 8.40. Ratios of the effective dose rate to the person to the measured dose rate at 1 m from the same source were calculated.

FALCON code-based analysis of PWR fuel rod behaviour during RIA transients versus new U.S.NRC and current Swiss failure limits

Outcomes of the FALCON code analysis-related part of the STARS-ENSI Service Project on Evaluation of the new U.S.NRC RIA Fuel Safety Criteria and Application to the Swiss Reactors are presented. Substantial conservatism of the updated safety limits for high-temperature and PCMI cladding failure, as proposed in the NRC Regulatory Guide RG 1.236, is confirmed. Applicability of the updated failure limits to fuel safety analysis in the Swiss PWRs, as applied to standard fuel designs using UO2 fuel pellets and SRA Zry-4 as cladding materials is discussed. Conducting of new integral RIA tests with irradiated samples using doped- and gadolinia fuel pellets to support appropriate fuel safety criteria for RIA events is recommended.

Seismic Response Amplification Factors of Nuclear Power Plants for Seismic Performance Evaluation of Structures and Equipment due to High-frequency Earthquakes

Analysis of the 2016 Gyeongju earthquake and the 2017 Pohang earthquake showed the characteristics of a typical high-frequency earthquake with many high-frequency components, short time strong motion duration, and large peak ground acceleration relative to the magnitude of the earthquake. Domestic nuclear power plants were designed and evaluated based on NRC's Regulatory Guide 1.60 design response spectrum, which had a great deal of energy in the low-frequency range. Therefore, nuclear power plants should carry out seismic verification and seismic performance evaluation of systems, structures, and components by reflecting the domestic characteristics of earthquakes. In this study, high-frequency amplification factors that can be used for seismic verification and seismic performance evaluation of nuclear power plant systems, structures, and equipment were analyzed. In order to analyze the high-frequency amplification factor, five sets of seismic time history were generated, which were matched with the uniform hazard response spectrum to reflect the characteristics of domestic earthquake motion. The nuclear power plant was subjected to seismic analysis for the construction of the Korean standard nuclear power plant, OPR1000, which is a reactor building, an auxiliary building assembly, a component cooling water heat exchanger building, and an essential service water building. Based on the results of the seismic analysis, a high-frequency amplification factor was derived upon the calculation of the floor response spectrum of the important locations of nuclear power plants. The high-frequency amplification factor can be effectively used for the seismic verification and seismic performance evaluation of electric equipment which are sensitive to high-frequency earthquakes.

A One-Year Review of the Radiation Safety Procedures and Protocols for Peptide Receptor Radionuclide Therapy (PRRT) Using Lutetium Lu 177 Dotatate

The treatment of gastroenteropancreatic neuroendocrine tumors (GEP-NETs) using drugs known as Peptide Receptor Radionuclide Therapy (PRRT) has been active for 14 months at our institution. The radionuclide used in the PRRT Program is Lutetium Lu 177 dotatate. This research seeks to share the policies and procedures of our radiation safety program, as well as our experiences in the first year of this therapy program. The key guidelines used to establish a radiation safety program include the principle of ALARA (as low as reasonably achievable), the NRC guidelines and any state-specific requirements, the properties of the radionuclide itself, and the management of contamination. Following the calculation formalism outlined in NRC Regulatory Guide 8.39 Release of Patients Administered Radioactive Materials, policy was set to release patients with survey readings at or below 2.1 mR/hr at 1 meter. Patients receive PRRT in a shielded room; the infusion pumps/stands are covered with chuck pads. The primary excretion pathway is through urine, thus absorbent pads are also around the toilet and on the floor of the patient restroom. Vendor estimates of excretion through urine are 58% in the first 24 hours and over 99% in 14 days. The night after the treatment patients either go home or stay in our onsite hotel. Upon checkout from the hotel room, physics staff conducts a room release survey. A Ludlum Model 26 Integrated Frisker calibrated to a collection efficiency of 20%. All linens and surfaces, specifically in the bathroom, are surveyed. If contamination is detected, the following take place: 1. If it is removable (e.g., bed sheet, towel), it is double bagged, labeled, and taken to short term decay. 2. If it is not removable (e.g., toilet, floor in bathroom, chair), it is cleaned and re-measured prior to the room being opened up for a new patient. After noticing the restroom floor was often contaminated, we began lining the floors of all Lu 177 patient restrooms with chuck pads. The rate of restroom floor contamination in the patient rooms has dropped significantly since making this procedural change. To date, 25 patients have received a total of 73 treatments. All were prescribed a 200 mCi dose of Lu 177 dotatate. The actual amount of Lu 177 infused ranged from 191.67-215.106 mCi (mean 202.2 mCi). Time from the start of injection to patient release ranged from 2 hours to 4.68 hours (mean 3.1 hours). Survey readings of patients after injection ranged from 0.54 mR/hr to 4.1 mR/hr (mean 1.5 mR/hr) at 1 meter. From our experience in the first year of this program, contamination risk is best minimized through stringent patient education on care when using the restroom and proper preparation of hotel rooms where patients will stay. Ethical considerations should be made when considering sending patients to local hotels. Also, permissible exposure levels to caregivers should take into account the fact that most patients receive 4 treatments.

Seismic Response Analysis of Nuclear Power Plant Structures and Equipment due to the Pohang Earthquake

The probabilistic seismic safety assessment is one of the methodology to evaluate the seismic safety of the nuclear power plants. The site characteristics of the nuclear power plant should be reflected when evaluating the seismic safety of the nuclear power plant. The Korea seismic characteristics are strong in high frequency region and may be different from NRC Regulatory Guide 1.60, which is the design spectrum of nuclear power plants. In this study, seismic response of a nuclear power plant structure by Pohang earthquake (2017.11.15. (KST)) is investigated. The Pohang earthquake measured at the Cheongsong seismic observation station (CHS) is scaled to the peak ground acceleration (PGA) of 0.2 g and the seismic acceleration time history curve corresponding to the design spectrum is created. A nuclear power plant of the containment building and the auxiliary buildings are modeled using OPENSEES to analyze the seismic response of the Pohang earthquake. The seismic behavior of the nuclear power plant due to the Pohang earthquake is investigated. And the seismic performances of the equipment of a nuclear power plant are evaluated by the HCLPF. As a result, the seismic safety evaluation of nuclear power plants should be evaluated based on site-specific characteristics of nuclear power plants.

Cyber Security Risk Evaluation of a Nuclear I&C Using BN and ET

Abstract Cyber security is an important issue in the field of nuclear engineering because nuclear facilities use digital equipment and digital systems that can lead to serious hazards in the event of an accident. Regulatory agencies worldwide have announced guidelines for cyber security related to nuclear issues, including U.S. NRC Regulatory Guide 5.71. It is important to evaluate cyber security risk in accordance with these regulatory guides. In this study, we propose a cyber security risk evaluation model for nuclear instrumentation and control systems using a Bayesian network and event trees. As it is difficult to perform penetration tests on the systems, the evaluation model can inform research on cyber threats to cyber security systems for nuclear facilities through the use of prior and posterior information and backpropagation calculations. Furthermore, we suggest a methodology for the application of analytical results from the Bayesian network model to an event tree model, which is a probabilistic safety assessment method. The proposed method will provide insight into safety and cyber security risks.

Modeling of the ORNL PCA Benchmark Using SCALE6.0 Hybrid Deterministic-Stochastic Methodology

Revised guidelines with the support of computational benchmarks are needed for the regulation of the allowed neutron irradiation to reactor structures during power plant lifetime. Currently, US NRC Regulatory Guide 1.190 is the effective guideline for reactor dosimetry calculations. A well known international shielding database SINBAD contains large selection of models for benchmarking neutron transport methods. In this paper a PCA benchmark has been chosen from SINBAD for qualification of our methodology for pressure vessel neutron fluence calculations, as required by the Regulatory Guide 1.190. The SCALE6.0 code package, developed at Oak Ridge National Laboratory, was used for modeling of the PCA benchmark. The CSAS6 criticality sequence of the SCALE6.0 code package, which includes KENO-VI Monte Carlo code, as well as MAVRIC/Monaco hybrid shielding sequence, was utilized for calculation of equivalent fission fluxes. The shielding analysis was performed using multigroup shielding library v7_200n47g derived from general purpose ENDF/B-VII.0 library. As a source of response functions for reaction rate calculations with MAVRIC we used international reactor dosimetry libraries (IRDF-2002 and IRDF-90.v2) and appropriate cross-sections from transport library v7_200n47g. The comparison of calculational results and benchmark data showed a good agreement of the calculated and measured equivalent fission fluxes.

Propped cantilever mesh convergence study using hexahedral elements

AbstractThe Task Group on Computational Modelling for Explicit Analyses in the ASME Boiler and Pressure Vessel Code Committee was set up in August 2008 to develop a quantitative finite element modelling guidance document for the explicit dynamic analysis of energy limited events. This guidance document will be referenced in the ASME Boiler and Pressure Vessel Code Section III Division 3 and NRC Regulatory Guide 7·6 as a means by which the quality of a finite element model may be judged. In energy limited events, which the guidance document will address, ductile metallic materials will suffer significant plastic strains to take full advantage of their energy absorption capacity. Accuracy of the analyses in predicting large strains is therefore essential. One of the issues that this guidance document will address is the issue of the quality of a finite element mesh, and in particular, mesh refinement to obtain a convergent solution. That is, for a given structure under a given loading using a given type of ...

Development of integrated dose assessment program for accident analysis of LILW managed in temporary storage facility

In this study a program, INDAC++, was developed to conduct accident analysis on low and intermediate level waste (LILW) managed in a temporary storage facility (TSF). In order to conduct this work, a variety of parameters were considered such as risk factor, radionuclide inventory, atmospheric dispersion factor ( χ/ Q), dose conversion factor, and so on. The main risk factors leading to arbitrary accidents were developed by the master logic diagram (MLD) method based on a fault tree analysis (FTA). The dose to curie (DTC) method was used to calculate the radionuclide inventory. Then, the release rates with respect to the type of wastes and radionuclides were considered in order to evaluate the amount of radionuclides released by the risk factors. Furthermore, the atmospheric dispersion factor ( χ/ Q) was calculated by the PAVAN code based on various models and assumptions suggested in the U.S. NRC Regulatory Guide 1.145. The dose assessments for arbitrary accidents originating in the TSF for LILW management are conducted by considering the external exposure dose by radioactive plume and the internal exposure dose by breathing. Synthesizing many parameters, the INDAC++, which is an integrated dose assessment code package for normal reaction operation and for accident analysis of LILW managed in the TSF, was developed using Visual C# in Microsoft Visual Studio.NET 2003. Furthermore, the INDAC++ can calculate the radiological risk with regard to the main risk factors for workers and the public.

Establishing a Threshold for 131I Bioassay in Nuclear Medicine Personnel

Since the late 1970's, manufacturers in nuclear medicine have reformulated the I solution to reduce the volatility of the iodine. There has also been an increase in use of the iodide in encapsulated form. Per the requirement of the current U.S. Nuclear Regulatory Commission (U.S. NRC) regulation, with the available results on the volatility of the reformulated radioiodine, we review the I bioassay program for nuclear medicine workers. Our analysis shows the threshold quantity for bioassay monitoring for the routine use of I in nuclear medicine is much higher than the criteria set in U.S. NRC Regulatory Guide 8.20. The latter is a broad bioassay guideline for the general usage of radioactive iodine. For treatment of thyroid carcinoma and hyperthyroidism, a single therapeutic I dose large enough to yield a detectable thyroid burden is very unlikely to occur in a nuclear medicine clinic. Accidental ingestion or inhalation would be an exception to our conclusion. Based on this analysis, we propose a new bioassay policy for the routine use of I in nuclear medicine clinics.

Epistemic Uncertainty in the Ranking and Categorization of Probabilistic Safety Assessment Model Elements: Issues and Findings

In this work, we study the effect of epistemic uncertainty in the ranking and categorization of elements of probabilistic safety assessment (PSA) models. We show that, while in a deterministic setting a PSA element belongs to a given category univocally, in the presence of epistemic uncertainty, a PSA element belongs to a given category only with a certain probability. We propose an approach to estimate these probabilities, showing that their knowledge allows to appreciate "the sensitivity of component categorizations to uncertainties in the parameter values" (U.S. NRC Regulatory Guide 1.174). We investigate the meaning and utilization of an assignment method based on the expected value of importance measures. We discuss the problem of evaluating changes in quality assurance, maintenance activities prioritization, etc. in the presence of epistemic uncertainty. We show that the inclusion of epistemic uncertainly in the evaluation makes it necessary to evaluate changes through their effect on PSA model parameters. We propose a categorization of parameters based on the Fussell-Vesely and differential importance (DIM) measures. In addition, issues in the calculation of the expected value of the joint importance measure are present when evaluating changes affecting groups of components. We illustrate that the problem can be solved using DIM. A numerical application to a case study concludes the work.

EVALUATION OF THE POTENTIAL ABSORBED DOSES FROM PATIENTS BASED ON WHOLE-BODY 131I CLEARANCE IN THYROID CANCER THERAPY

The evaluation of the absorbed dose from radioactive patients during the treatment of thyroid disease is an important factor in establishing precautions in these procedures, and the I retention/excretion by patients' bodies provides additional information to medical and radioprotection service. In 94 patients, the measurement of exposure rates was performed over 7 d following NaI administration, and the rates permitted the study of the dynamics of excretion and the potential dose evaluation. The administered activities ranged from 3.7 GBq (100 mCi) to 16.65 GBq (450 mCi), and the results proved that the majority of the activity is excreted by patients in the first 3 d after NaI administration. The average (131)I activity excreted at 24, 48, 72, 96, and 120 h after oral administration was (72 +/- 10), (91 +/- 6), (97 +/- 3), (98.9 +/- 1.5), and (99.6 +/- 0.7)%, respectively. According to the administered activity, the evaluation of the accumulated absorbed dose from patients ranged from 3.0 +/- 0.7 to 8.4 +/- 1.1 mSv at 1 m and 1.2 +/- 0.4 to 3.2 +/- 0.4 mSv at 2 m. The data reported here are important to radioprotection policy and to add to and improve on the guidelines reported in U.S. NRC Regulatory Guide 8.39.

ICONE11-36186 INTEGRATED LEAK RATE TESTING OF CONTAINMENTS-A PERSPECTIVE

Appendix J of 10 CFR 50,A Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors, @ sets the testing requirements for preoperational and periodic verification of the leak-tight integrity of the primary reactor containment, including systems and components which penetrate containment of light water-cooled power reactors, and establishes the acceptance criteria for such tests. The purposes of the tests are to assure that leakage through the primary reactor containment and systems and components penetrating primary containment would not exceed allowable leakage rate values as specified in the plants=technical specifications. The current Appendix J requirements provide two options for performing the tests. Option A (fully deterministic) requires that after the preoperational leakage rate tests, a set of three integrated leak rate tests (ILRT-termed as Type A tests) shall be performed, at approximately equal intervals during each 10-year service period. Option B does not provide a quantitative requirement for scheduling the periodic Type A tests. However, NEI 94-01, A Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50,Appendix J, @ and NRC Regulatory Guide (RG) 1.163, A Performance-Based Containment Leak Test Program, @ provide guidelines for determining the frequencies of preoperational and periodic leak rate tests using a performance based approach. The Option B requirements and NEI 94-01 guidance are based on NUREG-1493, A Performance Based Containment Leak-Test Program. @ Based on the information provided by the operating data and their risk significance, the study calculated the change in risk (in person-rem) to public for the 15 alternatives considered in the study. However, recognizing the non-sensitivity of risk to change in Type A leak rate testing frequencies, as depicted in NUREG-1493,a number of licensees are proposing changes to their Type A test frequencies using NRC risk-informed guidance in RG 1.174, A An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis. The paper discusses certain key deterministic aspects of the risk-informed decision for the plant-specific changes in the ILRT frequencies.

Some experiences with treating thyroid cancer patients.

U.S. NRC Regulatory Guide 8.39 provides for the release of patients treated with 131I provided that predetermined calculations indicate that no member of the public will receive a total dose equivalent in excess of 5 mSv (500 mrem). When this condition cannot be met or there are other reasons for keeping the patient hospitalized after treatment, control of contamination and exposure from the patient must be taken into consideration. If the patients are hospitalized following treatment, decontaminating the patient's room after discharge and controlling the exposure potential from the patient are considerations for the hospital radiation safety staff. This paper reviews the experiences from fifty patients treated as inpatients over the past two years.

Crack-opening-area analyses for circumferential through-wall cracks in pipes—Part I: analytical models

Leak-before-break (LBB) analyses for circumferentially cracked pipes are currently being conducted in the nuclear industry to justify elimination of pipe whip restraints and jet impingement shields which are present because of the expected dynamic effects from pipe rupture. The application of the LBB methodology requires calculation of leak rates. The leak rates depend on the crack-opening area of the through-wall crack in the pipe. In addition to LBB analyses which assume a hypothetical flaw size, there is also interest in the integrity of actual leaking cracks corresponding to current leakage detection requirements in NRC Regulatory Guide 1.45, or for assessing temporary repair of Class 2 and 3 pipes that have leaks, as are being evaluated in ASME Section XI. The objectives of this study were to review, evaluate, and refine current predictive models for performing crack-opening-area analyses of circumferentially cracked pipes. A three-phase effort was undertaken to accomplish this goal. It is described here in a series of three papers generated from this study. In this first paper (Part I — Analytical models), a comprehensive review is performed to determine the current state-of-the-art in predicting crack-opening displacements for circumferentially cracked pipes under pure bending, pure tension, and combined bending and tension loads. Henceforth, new and improved analytical models and some preliminary results are presented for cases where current methods are inadequate or there are no available methods. Also, based on this review, a number of appropriate predictive models are identified for a systematic evaluation of their accuracy. The results of their evaluations will be presented and examined in the forthcoming companion papers (Part II — Model validations [1] and Part III — Off-center cracks, restraint of bending, thickness transition, and weld residual stresses) [2].

Study on the behavior of129I in the terrestrial environment

Transfer parameters for dose assessment such as deposition velocity (Vg), fraction of directly deposited on plants to the edible parts (r) and transfer factors (Biv) have been evaluated. Deposition velocity under drying and wet conditions was evaluated about 0.8 cm/s. Fraction of directly deposited on plants and transfer factors in plants showed a lognormal distribution. Geometric means of fraction of deposited in rice, leafy vegetable and grass were <9·10−3, 0.2 and about 9·10−2, respectively. Geometric mean of the transfer factor in leafy vegetable was about 3·10−2. These results were compared with the parameters given in the US NRC Regulatory Guide and previous literature values. The results showed that the fraction of directly deposited on plants were smaller than the recommended value in the NRC Regulatory Guide. They also showed that the deposition velocity and transfer factors were almost similar to the results reported in the literature.

Prenatal Radiation Exposure Policy

A policy on prenatal radiation exposure at two nuclear power plants was revised to give better assurance of compliance with NCRP recommendations on fetal radiation exposure. This action was taken after publication of NCRP 91 in June 1987 to provide better assurance that a total dose equivalent limit to an embryo-fetus be no greater than 0.5 mSv (0.05 rem) in any month and no more than 5 mSv (500 mrem) for a gestation period. For any female worker to receive radiation exposure greater than 1.5 mSv (0.15 rem) in a month at these nuclear power plants, she was asked to initiate an administrative request for radiation exposure in excess of this limit. In this request, she was asked to acknowledge that she was aware of the guidance in U.S. NRC Regulatory Guide 8.13. A worker who had the potential for radiation exposure in excess of 1.5 mSv (0.15 rem) refused to process this request and was consequently denied overtime work. She filed a grievance for denial of overtime, and this grievance was submitted for labor arbitration in June 1988. The arbitration decision and its basis and related NRC actions are discussed.

Computer Code TERFOC-N to Calculate Doses to Public Using Terrestrial Foodchain Models Improved and Extended for Long-Lived Nuclides

A computer code TERFOC-N has been developed to calculate doses to the public due to atmospheric releases of radionuclides in normal operations of nuclear facilities. The code calculates the highest individual dose and the collective dose from four exposure pathways; internal doses due to ingestion and inhalation, external doses due to cloudshine and groundshine. A foodchain model, which is originally referred to the U. S. NRC Regulatory Guide 1.109, has been improved to apply to not only LWRs but also other nuclear facilities. This report describes the models employed and gives a sample run performed by the code. The parameters which were sensitive to ingestion dose were identified from the results of sensitivity analysis. The models which significantly contributed to the dose were identified among the models improved and extended here.

A simplified approach of a systematic functional task analysis (SFTA) for a nuclear power plant

A systematic functional task analysis (SFTA) for control room instrumentation of a nuclear power plant has been developed. The methodology involves a systematic review of the plant worst-accident conditions, NRC Regulatory Guide 1.97 postaccident monitoring requirements, and multiple failure conditions. Computerized database management of the analysis is expected to provide consolidated instrumentation (display/control) characteristics required by operators to mitigate plant transient/accident conditions. >

Computer code TERFOC-N to calculate doses to public using terrestrial foodchain models improved and extended for long-lived nuclides.

A computer code TERFOC-N has been developed to calculate doses to the public due to atmospheric releases of radionuclides in normal operations of nuclear facilities. The code calculates the highest individual dose and the collective dose from four exposure pathways; internal doses due to ingestion and inhalation, external doses due to cloudshine and groundshine. A foodchain model, which is originally referred to the U. S. NRC Regulatory Guide 1.109, has been improved to apply to not only LWRs but also other nuclear facilities. This report describes the models employed and gives a sample run performed by the code. The parameters which were sensitive to ingestion dose were identified from the results of sensitivity analysis. The models which significantly contributed to the dose were identified among the models improved and extended here.