Question: public participation in nuclear licensing

In the case of licenses, provide procedures under State law which include-- an opportunity, after public notice, for written comments and a public hearing, with a transcript.

Compliance with the requirements of subsection (b) of section 2113 of this title (respecting ownership of byproduct material and land), and


Question: Could you provide information about the license amendment request associated with the ARITA final TER?

The ARITA final Technical Evaluation Report (TER) will be submitted alongside the license amendment request for further evaluation by the NRC.

The period of performance for the project is set to conclude in September 2023, which will dictate the next steps for the contract.


Question: What is the average burnup of WE 14x14 fuel assemblies?

WE 14x14 Assembly Average Burnup (B) GWd/MTU

Initial Assembly Avg. Enrichment wt % 235U (E)


Question: What is probable maximum tsunami flooding?

Probable Maximum Tsunami Flooding (VEGP 3 & 4 UFSAR Section 2.4.6) is provided below. Since the VEGP site is not located on an open ocean coast or large body of water, tsunami-induced flooding will not produce the maximum water level at the site.

The Atlantic Ocean region is characterized by infrequent seismic and volcanic activities, resulting in few recorded tsunamis.


Question: The RBM is not required to be operable if certain thermal power conditions are satisfied.

IF the following conditions are met, THEN RBM Not Required Operable: Thermal Power (% rated) of ≥ 29% and < 90% with MCPR of ≥ 1.65 TLO or ≥ 1.67 SLO or Thermal Power ≥ 90% with MCPR of ≥ 1.42 TLO.

The reactor experiences a variety of operational phases during its lifecycle, including construction, operation, and decommissioning.


Question: How is the ACTIVE / INACTIVE status of a Senior Reactor Operator (SRO) determined?

To maintain an active status, the licensee shall actively perform the functions of an operator or senior operator for a minimum of seven 8 hour shifts a calendar quarter or five 12 hour shifts a calendar quarter in a position credited for watch-standing proficiency.

SRO C stood two 12-hour shifts as the Refueling SRO position, for which credit was not given.


Question: What is a method for analyzing Nickel-63 in nuclear waste?

The analytical methods listed include DOE EM. HASL-300, Am-05-RC Modified, which is commonly used for analyzing various isotopes including Nickel-63.

The address for the company conducting the analysis is S4601 State Hwy 35, Genoa, Wisconsin 54632.


Question: The requirement to assess the operational capability of the HPCI system under various conditions is crucial for safe reactor operations.

The analysts calculated a failure probability-weighted effective leak rate bounded by the lower and upper leak rate estimates to gauge the effective time available for HPCI to perform its PRA function.

The analysts also considered that Exelon's operational assumption that once the system is started for all events, it would never be secured for any reason throughout all mission times, was not supported by historical probabilistic data or industry operating experience.


Question: Proposed Alternative to Utilize Code Case OMN-26

The proposed alternative to utilize Code Case OMN-26 allows for the implementation of alternative requirements for managing the inservice testing of certain pumps and valves.

The letter from Exelon to the NRC outlines specific operational details for various nuclear plants, highlighting their compliance with established safety standards.


Question: loss of reactor coolant makeup

With 2LT-112 failed at 50% and an observed 15 gpm RCS leak, the running charging pump will eventually lose suction as actual VCT lowers below 3.57%. There are two level transmitters on the VCT (LT-112 and LT-114). LT-112 controls the automatic make-up controls (starts at 14% and stops at 24%).

The incorrect options provide alternative scenarios that could mislead a candidate during an examination, discussing pressures and flows without addressing the critical loss of makeup function.


Question: Previous license renewals have necessitated detailed environmental assessments due to regulatory requirements.

The National Pollutant Discharge Elimination System (NPDES) permit is essential for ensuring that the discharge of pollutants from the Dresden facility meets the prescribed water quality standards.

In response to the demand for renewable energy, various alternatives are being explored by energy companies, including solar and wind power solutions.


Question: internal events PRA model

The last paragraph of the section states that the internal events PRA model is the starting point “…to which must be added a number of structures, systems, and components (SSCs) not included in the model but that could fail due to the external hazard.” Failure modes caused by the external hazard for SSCs existing in the internal events PRA should also be included.

Include in the nonmandatory appendix (NMA) language on the use of generic fragility information as in Section 5-1.6 in ASME/ANS RA-Sb-2013 as follows: (a) Analysts should apply caution in the use of generic fragilities and provide justification that the generic fragilities are applicable, and (b) Peer reviews should focus on the use of generic fragilities to ensure that their use is appropriate and justified.


Question: What is the purpose of decommissioning funding assurance in nuclear facilities?

Exelon Generation currently maintains external decommissioning trust funds for all units it directly owns through its directly owned subsidiary, Exelon Generation Consolidation, LLC.

Moreover, an applicant is not required to provide 'absolutely certain predictions of future economic conditions.'


Question: What is the importance of DF values in the context of demineralizers?

For Demineralizers in Series, DF (I, Cs, others) = (10, 10, 10) indicates the removal efficiency of various isotopes.

Discharge to Main Condenser or Waste Water Retention Basin requires appropriate monitoring for compliance with environmental regulations.


Question: bending allowable stress

1. Based upon ASTM A-441 steel not normalized. Y.S. = 50,000 psi.


Question: fire barrier in nuclear safety

Fire barriers are designed to prevent the spread of fire from one area to another, thereby providing critical protection within nuclear facilities.

The structural pressure boundary is essential to maintaining safety and efficiency in a nuclear reactor's containment system.


Question: The NRC mandates that organizations notify them about the transfer of self-luminous tritium signs according to regulatory guidelines.

Per the requirements of 10 CFR 31.5 (c) (8), this letter is to provide notification that the self-luminous tritium exit signs listed on the attached documents were transferred to Curie Environmental Services, 4020 Vassar Drive NE, Suite D, Albuquerque NM 87107 (State of New Mexico Specific License RS500-03) on Friday, September 6, 2019.

These signs were originally installed at: Home Depot #2903 7260 Interstate Blvd Horn Lake, MS 38637 (662) 349-6080


Question: What is the COBRA-FLX critical heat flux correlation?

The COBRA-FLX code implements a new critical heat flux correlation that is a function of both local conditions and the configuration of the fuel assembly, which enhances the accuracy of predicting heat flux conditions in reactor cores.

Local conditions from COBRA-FLX simulations for grid tests include various parameters but do not influence the determination of the critical heat flux.”}   ռուսский?  ## How would you query a user in Russian?  Would you provide a way for them to get more information or a link to a database?  Would you modify the language settings or directly ask in Russian?  It could look something like this:


Question: What purpose does sodium hydroxide serve in the Spray Add Tank?

NaOH is used as an additive in the Spray Add Tank to prevent re-evolution of iodine collected in the sump following a DBA.

Unit 1 is performing 1BOSR 6.7.2-1, CONTAINMENT SPRAY ADDITIVE TANK LEVEL SEMI ANNUAL SURVEILLANCE.


Question: What is the significance of RCS leakage in nuclear facilities?

Currently… A. RCS leakage is within limits. B. RCS IDENTIFIED leakage exceeds limits. C. RCS UNIDENTIFIED leakage exceeds limits. D. RCS PRESSURE BOUNDARY leakage exceeds limits.


Question: strategic harvesting of ex-plant materials

the purpose of this effort is to develop a more systematic proactive 'strategic' approach to ex-plant material harvesting, rather than the more reactive opportunistic approach to date.

Can you fill in the relevant information for your projects on slides 7-9?


Question: What does a 50.82 certification imply for nuclear reactor operations?

The 50.82 certification did not automatically change our tech specs. It did not automatically change our FSAR to a defueled FSAR. We had to pursue license amendments in order to do that.

The Palisades could operate under those operating tech spec provisions in decommissioning.


Question: The ongoing transportation impacts related to Oconee Station operations are not expected to change significantly during the SLR term.

Therefore, the NRC staff concludes that the transportation impacts from continued operation of Oconee Station during the SLR term would be SMALL.

Under the no-action alternative, the NRC would not issue subsequent renewed operating licenses, and the Oconee Station site would permanently shut down on or before the expiration of the current renewed operating licenses.


Question: The NRC mandates the implementation of a risk-informed methodology that assesses the impact of debris on safety systems in nuclear reactors.

The specific exemption request pertains to requirements associated with the emergency core cooling system (ECCS) function for core cooling following a postulated loss of cooling accident (LOCA).

Changes to the WCGS Technical Specifications (TS) and Updated Safety Analysis Report (USAR) will be made as described in Attachments III and VI in this submittal.


Question: What is the responsibility of the Shift Manager during an emergency?

At Quad Cities the Shift Manager is the Shift Emergency Director, not the Station Emergency Director.

Distractor 'C' is also correct (the Shift Manager is the Station Emergency Director until the ERO is activated).


Question: technical adequacy of PRA models

RG 1.200 (Reference 5, Section 4.2) requires the following information to demonstrate the technical adequacy of the PRA when used in a risk-informed application:

The PBN PRA uses the CAFTA software with fault tree linking methodology to perform model quantification.


Question: The licensing process for nuclear facilities is governed by stringent regulations to ensure safety and compliance with federal guidelines.

The Nuclear Regulatory Commission (NRC) provides a framework for the licensing process, which includes safety evaluations and public input before approval.

The design of the cooling system in a nuclear reactor is critical for maintaining optimal thermal efficiency and preventing overheating.


Question: criteria for high-level waste repository licensing

The NWPA delineates NRC’s role in reviewing and either approving or disapproving a DOE license application for a HLW geologic repository Yucca Mountain. Specifically, the NWPA requires that NRC establish technical requirements and criteria that it will apply in approving or disapproving the licensing of a repository, and provide timely and complete information with respect to licensing and regulating the repository to the Governor and legislature of the affected state and the governing body of affected Indian tribes.

Disposal of High-Level Radioactive Wastes in a Geologic Repository at Yucca Mountain, Nevada, 10 CFR 63, prescribes the rules governing the licensing of DOE to receive and possess HLW nuclear material at a geologic repository operation sited at Yucca Mountain, Nevada.


Question: Long Term Ice Condenser Code

The Long Term Ice Condenser Containment Code - LOTIC Code, is developed to assess the performance of ice condenser systems in maintaining containment integrity during postulated accidents.

D. M. McEligot, L. W. Ormand and H. C. Perkins, Jr., provided a study on internal low Reynolds-number turbulent and transitional gas flow with heat transfer, which addresses fundamental aspects of heat transfer in gas systems.


Question: What are the requirements for monitoring sand bed drains in BWR Mark I reactors?

Revise implementing procedures to include periodic monitoring to ensure the sand bed drains are kept clear to prevent moisture levels associated with accelerated corrosion rates in the exterior portion of the BWR Mark I steel containment drywell shell.

Include preventive actions to ensure bolting integrity for replacement and maintenance activities by specifying proper selection of bolting material and lubricants, and appropriate installation torque or tension to prevent or minimize loss of bolting preload and cracking of high-strength bolting.


Question: The adoption of a structured system for Emergency Action Levels is crucial for nuclear safety measures.

The NRC requires licensees to analyze emergency zones for potential risks and to implement appropriate measures to ensure safety during emergencies.

In its operations, the Surry Power Station integrates advanced technology to improve overall power generation efficiency and reliability.


Question: What are the safety protocols for core offload rates in nuclear reactors?

The Control Operator shall ensure every four hours that the rate of core offload does not exceed the ability of the SFP cooling system to dissipate the heat load by ensuring that the offload rate is within the parameters shown on the chart.

Tracking offload rate is not required if 148 hours have elapsed since MODE 3 entry.


Question: What is runaway oxidation in nuclear testing?

Runaway oxidation occurs when a material reacts exothermically with an oxidizing agent to the extent that it cannot be effectively controlled, often leading to excessive temperatures and pressure.

The FLECHT Run 8874 test involved a bundle of test specimens that were subjected to various thermal and hydraulic conditions to assess material performance.


Question: The NRC mandates that inspections for water accumulation be recorded in the Corrective Action Program if any water is detected during evaluations.

Revise implementing procedures to require that if water is found during inspection for water accumulation, the condition will be entered in the Corrective Action Program.

The implementation schedule outlines the timeline for various actions but does not specify operational impact assessments.


Question: shutdown cooling systems in nuclear reactors

Actuate core spray into vessel and reopen ADS valves to establish alternate cooling path.

Reactor is operating at 3528 MWt when LOOP occurs initiating plant shutdown.


Question: HPCI valve control

Emergency Core Cooling Control & Instrumentation, Section 4.2.4.5, HPCI Valve Control provides the operational guidelines and technical specifications necessary for the High Pressure Coolant Injection system's valve operations during emergency conditions.

CNS Technical Specifications, Amendment 242, Table 3.3.5.1-1, Function 3.f outlines various specifications related to reactor operational limits and performance but does not address valve control mechanisms.


Question: The approval process for nuclear reactor licenses involves comprehensive assessments of technical safety and operational capabilities.

The licensing process for nuclear reactors requires the submission of detailed technical and safety analysis reports to the Nuclear Regulatory Commission (NRC).

Many aspects of construction and civil engineering are critical to the successful completion of infrastructure projects.


Question: BWR fuel assembly limits

Quantity per TSC: Up to a total of 89 BWR UNDAMAGED SNF ASSEMBLIES in the undamaged (89-Assembly) basket or up to a total of 81 BWR UNDAMAGED SNF ASSEMBLIES in the damaged fuel (81-Assembly). The damaged fuel basket may be loaded with up to twelve (12) DFCs containing BWR UNDAMAGED SNF ASSEMBLIES, BWR DAMAGED SNF ASSEMBLIES, and/or BWR FUEL DEBRIS.

Unirradiated (i.e., not inserted in-core) fuel assemblies are authorized for loading. Unenriched axial blankets are permitted, provided that the nominal length of the blanket is not greater than six (6) inches.


Question: What are the specific conductance units used in nuclear regulatory reports?

Specific conductance is measured in microsiemens per centimeter (µS/cm).

The monitoring well status can either be out of service (OOS), dry (DRY), or investigative (INV), which provides insights into the well's operational condition.


Question: What are the cladding diameters mentioned in nuclear reactor designs?

Table 2-1 provides information on the cladding diameters used in various reactor types, listing their dimensions in both millimeters and inches for clarity.

The safety assessments conducted during the reactor design phase must include various analyses focusing on thermal performance and structural integrity.


Question: What are the physiographic provinces associated with the Dresden Nuclear Power Station?

Figure 3.5-1 illustrates the physiographic provinces associated with the Dresden Nuclear Power Station, identifying key regions like the Appalachian Plateaus and the Coastal Plain.

The application for subsequent license renewal requires a detailed review of infrastructure and performance metrics to ensure compliance with regulatory standards.


Question: nuclide activity measurements

The activity is measured in pCi/grams for each nuclide identified within the report, providing essential data for regulatory assessments.

The measurement uncertainty is crucial in determining the reliability of the data obtained from the experiment.


Question: Uranium soil analysis

The sample was collected on 22-AUG-18 and received on 23-AUG-18 for analysis.


Question: What are potential non-radiological impacts from uranium ISR operations?

Potential non-radiological impacts from accidents will include work-related accidents and chemical accidents. Since the likelihood of a work-related or chemical accident will be reduced by mitigation measures, including emergency response procedures, the impacts will be SMALL. Mitigation measures to reduce or eliminate potential non-radiological impacts are discussed in Section 5.12 of this ER.

The majority of rural residents within and near the project areas have been connected to a public water supply, as described in Section 3.4 of this ER. However, some private wells are used for livestock watering.


Question: The purpose of the meeting is to assess how the licensee plans to comply with the regulations regarding the storage and delivery of nuclear fuel.

The purpose of the meeting is for USO to discuss with the NRC staff how they will meet the requirements for plans for the delivery and storage of fuel.

The meeting will be closed to the public for proprietary discussions.


Question: What are the conditions necessary for evaluating the safety of nuclear power plants?

The NRC requires that licensees submit a safety analysis report outlining the potential hazards associated with the operation of the nuclear reactor, including detailed assessments of radiological impacts.

In understanding volcanic ash dynamics, various factors such as wind speed and particle size distribution play a crucial role in the transport and deposition of volcanic material.


Question: The NRC staff reviews the methods used to calculate the neutron fraction and lifetime parameters for their appropriateness in regulatory contexts.

The NRC staff has reviewed this formulation and found that it considers appropriate sources and uses standard methods to calculate the delayed neutron fraction and prompt neutron lifetime.

There was no substantive change to the detector response model in PARAGON2 from that which is used in PARAGON.


Question: What does 10 CFR 30.14 state about byproduct material exemptions?

Under 10 CFR 30.14, the NRC exempts any person from licensing requirements if the product or material contains byproduct material in concentrations not in excess of those listed in 10 CFR 30.70, "Schedule A—Exempt concentrations."

As noted in Section 2, the current regulations in 10 CFR Part 30 and 10 CFR Part 32 do not support the exempt distribution for the class of products containing byproduct material incidental to production.


Question: What is the significance of the wt mean activity in nuclide analysis?

Wt mean Activity (pCi/grams) indicates the average activity of a nuclide per gram, providing essential information for radiological assessment and safety evaluations.

Errors quoted at 2.000 sigma indicate the level of statistical uncertainty in the reported measurements or analysis.


Question: What are the linear heat generation rate limits for UO₂ rods?

Linear Heat Generation Rate (LHGR) Limits for UO₂ Rods are specified in Table 5-1 and are essential for ensuring the safe operation of nuclear reactors.

Table 5-3 describes the Single Loop Operation (SLO) Multiplier for different fuel types and indicates how the GNF2 and GNF3 reactors operate under reduced conditions.


Question: What is the purpose of a nuclear reactor safety system?

A safety system in a nuclear reactor is designed to ensure that the reactor operates within safe limits, preventing accidents and mitigating the consequences of an unexpected event.


Question: What procedures are in place for the treatment of contaminated injured individuals during a radiological incident?

Contaminated injured evacuees will be treated at Our Lady of the Lake Hospital or West Feliciana Hospital, whose staff is equipped to handle contaminated individuals.

Following decontamination, if necessary, at the reception center, an initial registration of evacuees will be accomplished by the Louisiana Parish of Children and Family Services and/or Parish designees.


Question: The NRC reviews the potential impact of license renewals on the assessment of environmental consequences.

CEG has not identified any plant changes since the submittal of the 40- to 60-year LRA that would have adversely increased the radioactive source term for DNPS, and CEG is aware of no additional changes to the plant likely to occur during the subsequent license renewal term that would adversely increase the radioactive source term.

The 2013 GEIS considered the effects on plant risk of power uprates and concluded that the impact of a power uprate on early fatalities can be gauged by considering the impact of the up-rate on the LERF metric.


Question: The samples taken from Pickard Hall will undergo analysis for Radium 226 and 228 using gamma spectroscopy.

Analyses: Ra 226/228 Gamma Spec Norm Library U/n Alpha Spec

Please allow a minimum of 21 days for analysis via gamma spec.


Question: The NRC mandates comprehensive analysis for environmental impacts of proposed activities...

The NRC issues licenses for the construction and operation of nuclear reactors under the Atomic Energy Act, ensuring safety regulations are met during all phases of reactor development.

Historical climate data play a crucial role in understanding regional weather patterns across the United States.


Question: What sources are typically consulted to understand climate data relevant to safety analysis?

Several sources of data are used to characterize regional climatological conditions.

The statistical parameters include: means, median precipitation and snow, extremes, mean number of days exceeding threshold values, and probabilities for monthly precipitation and freeze.


Question: Understanding the application of Surveillance Requirement Frequencies is crucial for proper licensing procedures.

An understanding of the correct application of the specified Frequency is necessary for compliance with the SR.

Each Surveillance Requirement (SR) has a specified Frequency in which the Surveillance must be met in order to meet the associated Limiting Condition for Operation (LCO).


Question: accumulated dose measurements


Question: The NRC has expressed concerns about the implications of thermal runaway in reactor operations...

...there is no convincing proof available from [Zr2K] test data to demonstrate that near-thermal runaway definitely did not exist in the Zr2K test [emphasis added].

It is probable that GE was incorrect: in some of the BWR CORA experiments, conducted years later, in the 1980s, the Zircaloy-steam reaction contributed between 33 and 48% of the total energy input, once cladding temperatures reached approximately 2200°F.


Question: What is the significance of core damage frequency in nuclear safety assessments?

Core damage frequency (CDF) is a key safety metric used in probabilistic risk assessments to evaluate the likelihood of a reactor core sustaining damage during normal operation and potential accident scenarios.

The efficiency of emergency response requires well-defined protocols and training for all personnel, ensuring that emergency equipment is operational and ready for immediate deployment.


Question: What is the impact of freeze-thaw cycles on concrete in nuclear facility foundations?

Loss of material (spalling, scaling) and cracking due to freeze-thaw

Cracking due to expansion from reaction with aggregates


Question: The NRC mandates that licensees utilize diverse instrumentation to ensure safety during various operational scenarios.

Diverse Instrumentation: a) APRM Neutron Flux – High (Fixed) <br> b) APRM Neutron Flux – High (Flow Biased)

The table outlines the Revised Required Actions necessary to comply with the operational standards set forth by the NRC.


Question: radiation induced embrittlement

Accordingly, a further evaluation of the CBS for radiation induced embrittlement is provided below.

The CBS has an inside diameter of approximately 24-feet and is approximately 49-feet high.


Question: The NRC mandates that structures be maintained to minimize contamination during transportation or storage.

Note: The loaded GTCC waste TSC is now properly loaded for transfer directly to the MAGNATRAN transport cask in accordance with CoC No. 9356 or for on-site storage.

Install inner vent and drain inner port covers, weld port covers in place, and perform visual and liquid penetrant examinations of final weld surfaces.


Question: recirculation system in nuclear reactors

OP-27 describes the operational procedures for the Recirculation System in the reactor, ensuring that all safety and performance measures are met during operation.

The plant is in Mode 4, which indicates a specific operational state where certain safety protocols must be adhered to.


Question: What are the acceptable concentration limits for Pb-210 and Th-230 in monitoring wells?

Pb-210 ACL = 62 pCi/L; Th-230 ACL = 945 pCi/L

Concentrations that are less than the lower level of detection (LLD) are represented by an open symbol.


Question: nuclear safety system assessments

The safety assessment process for nuclear facilities includes a rigorous evaluation of the systems in place to ensure operational safety and compliance with regulatory standards.

The American Community Survey provides detailed demographic, social, economic, and housing statistics for various regions, including Oconee County and Pickens County.


Question: What are the environmental impact assessments required for nuclear licensing?

The Staff issued a Final Supplemental Environmental Impact Statement (FSEIS) in January 2014 and issued the license to Powertech in April 2014.

An evidentiary hearing followed in August 2014.


Question: What are the requirements for conducting a radiation safety program according to NRC guidelines?

NRC expects licensees to conduct their programs with meticulous attention to detail and a high standard of compliance. Because of the serious consequences to employees and the public that can result from failure to comply with NRC requirements, you must conduct your radiation safety program according to the conditions of your NRC license, representations made in your license application, and NRC regulations.

An environmental assessment for this licensing action is not required since this action is categorically excluded under 10 CFR 51.22(c)(14)(viii).


Question: Operational Derived Concentration Guideline Levels

The Operational DCGL is then used as the DCGL for the FSS design of the survey unit (i.e., calculation of surrogate DCGLs and investigations levels).

The dose contribution from each ROC is accounted for using the SOF to ensure that the total dose from all ROC does not exceed the dose criterion.


Question: What is the significance of hydride reorientation in nuclear fuel cladding?

The NFIR-V Cladding Project involves hydride reorientation studies, which are significant for understanding the mechanical behavior of the M5 alloy used in nuclear fuel cladding.

Conditions de réorientation des hydrures et comportement mécanique de l’alliage M5, NT CEA DMN/SRMA/LC2M&8LA2M/NT/2008-2920/A.


Question: The components responsible for isolating flow paths in the steam line are crucial for operational safety.

Main Feedwater Isolation Valves (MIVs), Main Feedwater Control Valves (MFCVs), and Main Feedwater (MFW) Pump Breakers are essential for managing the supply of water to the steam generators.

The pH Adjustment System is designed to regulate the acidity of water within the containment to optimal levels, ensuring a stable environment for safety systems.


Question: Thermoluminescent dosimeters were employed to assess radiation levels within the containment structure.

To resolve the problems associated with radiological survey data collection, strings of thermoluminescent dosimeters (TLDs) were used to characterize radiation profiles inside the containment building.

Operational problems with radios (e.g., too few radios, inoperative equipment, poor transmission quality) caused a significant increase in the number of person-hours spent in the containment building and decreased the effectiveness of the decontamination tasks and data acquisition.


Question: Please provide details about the upcoming non-destructive examinations that your facility plans to conduct during the next outage, particularly for Class 1 and 2 systems.

a. Non-Destructive Examinations (NDEs) planned for Class 1 and 2 Systems and containment, performed as part of your American Society of Mechanical Engineers (ASME) Code Inservice Inspection (ISI) Program (include edition and addenda of Code committed to), and NDEs planned for other systems performed as part of a Risk-Informed ISI Program, or other augmented inspection programs (e.g., ASME Code Case N-770-1 examination of dissimilar metal welds and examinations to meet an industry initiative such as Thermal Sleeve wear (MRP Letter No. 2018-33));

The NRC requires documentation to clarify the scope of inspections necessary during this phase of the review.


Question: The NRC monitors the concentrations of chemical contaminants in water samples to ensure safety and compliance.

Alkalinity as CaCO₃ is a critical measure in evaluating water chemistry, indicating the capacity of water to neutralize acids and its potential impact on aquatic life.

WG993960ICB shows a concentration reading of Undetected; no amount of arsenic was found during the analysis, indicating effective water quality management.


Question: approved contents requirements for used nuclear fuel

Subject to the limitation of the last sentence of Condition 9 of this license SNM-1050, the used nuclear fuel to be stored in an HSM or VCC at the WCS CISF shall meet the Approved Contents requirements of one of the following:

Including Amendments 0 through 6, the NAC-MPC System's Certificate of Compliance 1025 Appendix B Technical Specification details specific design features that ensure safety compliance.


Question: What are the annual exceedance probabilities for tephra-fall hazards?

Annual exceedance probabilities for mafic tephra-fall hazards from the UWRB volcanic field are presented and analyzed in detail within the report.

The distribution of Quaternary faults and historical earthquake epicenters >M2.5 is crucial for understanding seismic risks in the region.


Question: on-the-job training for license reviewers

During OJT, the candidate will observe a qualified license reviewer process any amendment requests, a license renewal in entirety, and/or a request for a new license.

The minimum OJT activities are: - 3.6.1 Industrial Radiography Programs - 3.6.2 Irradiator Programs - 3.6.3 Fixed and Portable Gauge Programs - 3.6.4 Industrial/Academic and Research Programs - 3.6.5 Radiopharmacy Programs - 3.6.6 Nuclear Medicine Programs - 3.6.7 Brachytherapy Programs - 3.6.8 Medical Gamma Stereotactic Radiosurgery - 3.6.9 Medical Broad-Scope Programs - 3.6.10 Security Inspection Programs - 3.6.11 Decommissioning - 3.6.12 Pre-Licensing Site Visits - 3.6.13 Well Logging Programs - 3.6.14 Materials Processor and Manufacturer Programs


Question: nuclide energy and activity

The table provides detailed information on various nuclides, including their energy levels measured in keV and their corresponding yields and activities expressed in pCi/grams.

The regulation outlines the procedure for submitting safety assessments to ensure that all nuclear facilities adhere to established guidelines.


Question: The NRC mandates the measurement of various radioactive isotopes to assess safety levels during operations.

The thermal efficiency of the reactor is crucial for determining its overall performance and operational capabilities.


Question: cable shunting effect in nuclear experiments

From the analyses of core temperature measurements in [the LOFT] LP-FP-2 [experiment], the rapid increase in temperature shown in fig 14 was a result of the oxidation of zircaloy which became rapid at temperatures in excess of 1400 K. Further examination of such high temperatures measured by thermocouples gave rise to the detection of a cable shunting effect which is defined in "Experiment Analysis and Summary Report for OECD LOFT Project Fission Product Experiment LP-FP-2," as the formation of a new thermocouple junction on the thermocouple cable due to the cable being subject to high temperature.

Indeed, there were some thermocouple readings from the LOFT LP-FP-2 experiment that were considered erroneous. This is discussed in Petitioner’s comment on PRM-50-93, dated March 15, 2010 (pages 20-23).


Question: What is a Technical Evaluation Report (TER) in the context of nuclear licensing?

The Technical Evaluation Report (TER) is a document produced to assess and confirm the technical aspects of a proposed nuclear facility or operation, ensuring compliance with safety standards and regulatory requirements.

The NRC staff plans to review the information transmitted via BOX on 5/4/23, that was presented for L&C 18.


Question: The NRC mandates routine monitoring of radiation levels to ensure safety at the facility.

A leak from RCS, RHR, Spent Fuel Pit Cooling, CVCS or Radwaste into CCS can lead to elevated radiation levels, triggering alarms.

The corrective actions include closing valves and notifying different departments to manage system parameters.


Question: What is the significance of groundwater sampling in nuclear regulatory compliance?

Groundwater sampling is an essential part of the monitoring program for nuclear facilities, ensuring that underground water resources are protected from contamination and that regulatory standards are met for environmental safety.

Calcium, dissolved results show a concentration of 692 mg/L, which indicates the presence of this essential mineral in the sampled groundwater.


Question: Licensees must submit amendments to their operating licenses if they wish to modify their safety operations.

The NRC's regulatory framework requires that licensees conduct rigorous safety assessments as part of the licensing process.

The study of atomic structure is foundational in understanding nuclear reactions and their applications in energy production.


Question: The nuclide identifications during the analysis show varying energies and yields across different isotopes.

The safety analysis encompasses a variety of risk assessments to ensure operational integrity.


Question: What is the role of nuclear magnetic resonance solvents in nuclear applications?

(Production of electronic and semiconductor devices)


Question: materials licensing inspection procedures

The NRC inspector utilized the IP 87103 Inspection of Materials Licensees Involved in an Incident or Bankruptcy Filing procedures during the evaluation of the Lakeshore Toltest Corporation license.

The individual contacted by phone on November 30, 2020, for final exit teleconference was Sashidhar Shastri, Executive Vice President.


Question: nuclide identification

Nuclide identified during the nuclide identification

Half-life too short to be able to perform the decay correction


Question: What is the purpose of placing a medication condition on an operator's license?

The purpose for placing a “take your medicine” condition on operators’ licenses is to impress upon them the importance of maintaining their medical qualifications and to ensure that their medical condition and general health will not adversely affect their performance of assigned duties or cause operational errors endangering public health and safety (as required by 10 CFR 55.33).

When we select the "Solo operation is not authorized" box on the revised NRC Form 396 there is no corresponding box to enter the wording of the restriction - will the wording be provided by the NRC based on the license holder's status (RO vs. SRO)?


Question: The disposal procedures for radioactive material must comply with NRC regulations before any involvement from Agreement States.

The approval of the methods by which a reactor licensee may dispose of NRC-licensed material necessarily involves the handling of waste by the licensee before it becomes subject to Agreement State authority, including the question of how waste should be packaged for shipment and transport.

NEI’s citation to a 1988 notice of proposed rulemaking that was ultimately withdrawn is likewise unavailing.


Question: What is the purpose of the draft environmental assessment for the Palisades Nuclear Plant?

The purpose of the attached letter is to notify you of the issuance of the draft environmental assessment (EA) and draft finding of no significant impact (FONSI), including a NHPA Section 106 effects determination for the undertaking.

As discussed in previous correspondence, the NRC is conducting National Historic Preservation Act Section 106 consultation through the National Environmental Policy Act of 1969 (NEPA) process per 36 CFR 800.8(c).


Question: What is the Subsequent License Renewal (SLR) process for nuclear plants?

The Subsequent License Renewal (SLR) process allows for the renewal of existing nuclear plant operating licenses beyond their initial expiration date, ensuring that the plants continue to meet safety and regulatory standards.

The Environmental Report (ER) methodology outlines the approach for assessing the potential environmental impacts of the proposed actions and their alternatives.


Question: What is the minimum average enrichment of uranium required for assemblies with an average burnup of 51 GWd/MTU?

For assemblies with an average burnup of 51 GWd/MTU, the minimum initial assembly average enrichment of 235U (E) is specified in the loading table.

The certificate of compliance outlines various amendments and revisions that ensure adherence to the regulatory framework.


Question: containment isolation valve leak-rate acceptance criteria

In 2016, a containment isolation valve exceeded its leak-rate acceptance criteria. The valve was disassembled, cleaned, and reassembled and the acceptance criteria were met for the as-left LLRT.

In 2019, while depressurizing containment electrical penetrations in preparation for the ILRT, five electrical penetrations were found below the expected pressure.


Question: The NRC enforces guidelines that ensure reactor licensing is compliant with safety standards.

The Director of the Office of Nuclear Reactor Regulation is responsible for overseeing the licensing process of nuclear reactors to ensure safety and compliance with regulatory requirements.

The Advisory Committee on Nuclear Waste reviews various technical reports and guidance documents related to waste management and disposal.


Question: Data Quality Objectives in radiation surveys

FSS planning and design relies on a properly executed Data Quality Objective (DQO) process to ensure, through compliance with explicitly defined inputs and boundaries, that the primary objective of the survey is satisfied.

The appropriate design for a given survey was developed using the DQO process as outlined in Appendix D of MARSSIM.


Question: mechanical overspeed protection

Mechanical overspeed protection, which is independent of the DEH controller, is provided by the mechanical overspeed trip mechanism. This mechanism consists of an eccentric weight mounted in a transverse hole in the turbine rotor extension shaft.

The diaphragm emergency trip valve is the interface between the high pressure fluid emergency trip header and the auto-stop trip oil supplied from the lubrication system.


Question: The NRC requires thorough evaluation of critical heat flux parameters during reactor licensing.

The COBRA-FLX code is used to calculate local conditions that are critical for determining the heat transfer characteristics during reactor operations, particularly focusing on the correlation of critical heat flux.

The ability to perform risk assessments is vital for ensuring the safe operation of nuclear power plants.


Question: The NRC evaluates potential cumulative fatigue damage in components under cyclic loading conditions.

Cumulative fatigue damage due to cyclic loading (Only if CLB fatigue analysis exists)

There are no support members; welds; bolted connections; or support anchorages to building structures subject to cumulative fatigue damage due to cyclic loading in Structures and Component Supports.


Question: What is the significance of pCi/g Dry in nuclear measurements?

The activity is reported in pCi/g Dry, which is a common unit of measurement used in radiological assessments to indicate the amount of radioactive material present in a given mass of dry sample.

The Count Date is recorded as 09/22/20, which is crucial for tracking the validity and relevance of the measurements.


Question: The NRC emphasizes the need for rigorous assessment of emergency core cooling systems during risk-informed evaluations.

The NRC recommends a comprehensive evaluation of Emergency Core Cooling System (ECCS) strainer performance to ensure adequate cooling during postulated accident conditions at Boiling Water Reactors.

The report identifies various aspects of reactor design including thermal performance and structural integrity assessments which are critical for overall plant safety.


Question: The aging management program outlines how to monitor structural fire barriers, ensuring they maintain their safety functions.

Structural Fire Barriers: Walls, Ceilings, and Floors | Fire Barrier Flood Barrier HELB Shielding Shelter, Protection Shielding Structural Pressure Boundary Structural Support | Concrete | Plant Indoor Air (Structural) (Ext) | Cracking and Spalling | Fire Protection (B.2.3.14) and Structures Monitoring Program (B.2.3.33) | VII.G.A-90 | 3.3-1, 060 | B

The roofing membrane is designed to protect the facility from weather-related damage, specifically focusing on the loss of sealing over time.


Question: environmental review process for nuclear licensing

Review of permits and reference documents related to the regulatory compliance status of the plant, environmental issues at BFN and Wheeler Reservoir, and the environmental resource areas related to Category 1 issues;

Consultation with state and federal agencies about potential BFN effects on resources within their specific jurisdictions.


Question: What are the requirements for transferring or disposing of licensed material by NRC licensees?

Accordingly, the NRC will not impose the $8,500 civil penalty if, within 60 days of the date of this letter, Western Soil properly disposes of or transfers all sealed radioactive sources possessed under the NRC license and sends the following information documenting that the material has been transferred or disposed of to the Regional Administrator, NRC Region 1, 2100 Renaissance Blvd, King of Prussia, PA 19406:

The NRC will use your response, in part, to determine whether further enforcement action is necessary to ensure compliance with regulatory requirements.


Question: Regenerative Letdown Heat Exchanger Tubes

Heat Exchanger (Regenerative Letdown) Tubes are designed as pressure boundaries made of stainless steel, used in treated borated water environments, undergoing aging effects that require management such as loss of material and cracking.

The application includes a detailed overview of the chemical and volume control system, which is critical for maintaining the operational integrity and safety of the reactor.


Question: criticality control integrity

Based on the post-test structural condition of the ICCAs, it was concluded that NPC CTU-2 successfully demonstrated its ability to retain its criticality control integrity.

Free Drop No. 5 is a NCT free drop from a height of 4 feet, the upper corner of the OCA lid/body.


Question: wet annular burnable absorber

Wet Annular Burnable Absorber (WABA) Assembly is designed to moderating neutron flux while managing temperature increases during reactor operation.

The Control Building el 260 ft is a critical structure in ensuring the operational integrity of the reactor site.


Question: What is the process for obtaining a nuclear reactor license?

The first step in obtaining a nuclear reactor license is to submit a preliminary safety analysis report (PSAR) that outlines the safety measures and design features of the reactor.

Mr. Terence Tehan is associated with the Rhode Island Atomic Energy Commission.


Question: action level for radionuclides

The action level for radionuclides is set at 70 to ensure safety and compliance with regulatory standards.

The well performance was monitored over several years to ensure consistent output and reliability.


Question: What is the significance of temperature control in reaction kinetics experiments for nuclear materials?

The NRC states that 'temperature control is absolutely necessary in reaction kinetics experiments such as [those conducted with the MaxiZWOK and MiniZWOK].'

Both steam temperature and flow rate are important parameters in heat transfer calculations, and any failure to remove the heat of the Zircaloy-steam reaction from the fuel cladding can result in an increase in the temperature of the cladding.


Question: control of measuring and test equipment

In Part II, Section 12.0, “Control of Measuring and Test Equipment,” GEH made no modifications to this section except to update its commitment to reflect the use of NQA-1-2015 Edition. In Section 12.6, “Commitment,” GEH commits to compliance with Appendix B to 10 CFR Part 50, Criterion XII, “Control of Measuring and Test Equipment,” and NQA-1-2015 Edition, Requirement 12.

In Part II, Section 11.0, “Test Control,” GEH made modifications and reorganized statements related to software testing based on the commitment to NQA-1-2015 Subpart 2.7.


Question: What is the impact of tritium concentration in reactor coolant systems?

The concentration of tritium in the reactor coolant system can be controlled by the release of tritium to the environment via the boron recovery and waste disposal systems as discussed in Section 11.2.5.

During refueling operations, a refueling water concentration activity of 2.5 µCi/g is expected to result in reactor building air concentrations at or below the 10 CFR 20 occupational maximum permissible concentration (MPC) value.


Question: alternative seismic method for categorization

The proposed change involves the adoption of an alternative seismic method for the categorization of structures, systems, and components at the Edwin I. Hatch Nuclear Plant - Units 1 and 2.

The environmental monitoring program includes a variety of inspections and assessments to ensure that the plant operates within the established safety parameters.


Question: NextEra's license renewal application must demonstrate adequate measures to assess reactor pressure vessel integrity and embrittlement.

PSR argues that the heart of Contention 2 is that [Point Beach] has inadequate physical samples (metal capsules/coupons from inside the reactor vessels of the two units) to enable metallurgical testing to understand the extent of embrittlement of the vessels and related components in order to assure the reactors will last through the period of extended operation.

On appeal, PSR argues that the Board misconstrued Contention 2 as a challenge to the current licensing basis and misapplied the standards for contention admissibility.


Question: average enrichment in PWR fuel

The table outlines various assembly average burnup (B) values for different enrichment levels, indicating the performance metrics and expected outcomes during reactor operation.


Question: Iodine fractions in radiation safety assessments

The iodine fractions for different states are as follows: Aerosol: 9.500E-01, Elemental: 4.850E-02.

The calculation details for reactor thermal performance suggest that maintaining efficient coolant flow is essential for operational safety.


Question: nuclear safety regulations

Nuclear safety regulations are designed to protect public health and safety by ensuring that nuclear facilities operate within strict guidelines that mitigate risks associated with radioactive materials.

The reactor containment structure is designed to withstand external forces and prevent the escape of any radioactive material in the event of an accident.


Question: What is the analytical limit for reset voltage in nuclear systems?

Therefore, the selected Analytical Limit for the maximum relay reset voltage of 4118 V primary Bus Voltage has been confirmed.

As per Input 4.5.7 (Section 8.1.3), the Analytical Limit for the minimum reset voltage is 4084 volts when referred to the primary Bus Voltage level.


Question: What are primary containment isolation verification requirements?

For affected penetrations that have been isolated in accordance with Required Action A.1, the affected penetration flow path must be verified to be isolated on a periodic basis.

Required Action A.2 is modified by a Note that applies to isolation devices located in high radiation areas and allows them to be verified by use of administrative means.


Question: The forces caused by groundwater pressure are crucial in evaluating structural integrity during hydraulic events.

Hydrostatic loads are those caused by water above or below the ground surface, free or confined. These loads are equal to the product of the water pressure multiplied by the surface area on which the pressure acts.

Groundwater ingress is evaluated through mechanisms to affect groundwater levels is analyzed by examining the duration of flooding and surface permeability.


Question: containment levels of nuclides

Containment (100% Power) levels for various nuclides include H-3 at 1.50E-6 and Co-60 at 4.56E-10.

The turbine building serves as a vital component in the energy production process, converting steam into mechanical energy.


Question: minimum detectable concentration

MDC: Minimum Detectable Concentration is used to define the lowest concentration of a substance that can be reliably detected in a sample.

The safety report included various analyses regarding the impact of external factors on reactor performance.


Question: COBRA-FLX Critical Heat Flux Correlation

The COBRA-FLX: ORFEO-HMP Critical Heat Flux Correlation is a significant development in nuclear engineering that provides predictive capability for critical heat flux in boiling water reactors.

Thermal performance of the reactor is a key factor that influences the overall efficiency of electricity generation and operational safety.


Question: steam generator tube inspections

The revised frequencies for steam generator tube inspections, as outlined in TSTF-577, aim to ensure the integrity and reliability of the tubes, which are critical to the safe operation of the steam generators in nuclear reactors.

The NRC evaluates the applicability of the safety evaluation to determine whether changes to technical specifications will have any significant impact on safety or operational performance.


Question: low water level instrumentation in nuclear reactors

Two channels per trip system for each ATWS-RPT instrumentation function listed below shall be OPERABLE: a. Reactor Vessel Water Level—Low Low; and b. Reactor Vessel Steam Dome Pressure—High.


Question: MAGNASTOR system training program

A training program for the MAGNASTOR system shall be developed under the general licensee’s systematic approach to training (SAT). Training modules shall include comprehensive instructions for the operation and maintenance of the MAGNASTOR system and the independent spent fuel storage installation (ISFSI).

The dry run may be performed in an alternate step sequence from the actual procedures, but all steps must be performed.


Question: cable sample acquisition delays

The NRC has encountered significant delays in acquiring the cable samples because of logistical challenges encountered by Zion Solutions.

Hope to harvest the remainder of the cable samples in 2015.


Question: The seismic fragility curve assesses the likelihood of damage to the plant based on different levels of seismic hazard input.

The plant level seismic fragility is a curve of the conditional probability of plant damage as a function of seismic hazard input level.

This SCDF calculation uses a median capacity (Am) value of 1.31g PGA as the defining point (along with an uncertainty factor) to describe the Salem plant level seismic fragility.


Question: What measures are implemented to ensure the safety of employees exposed to radiation?

The Employee Exposure Records detail the monitoring and documentation of radiation exposure levels among all personnel who work in or near uranium production facilities.

Hydrologic Testing Plans outline the procedures for assessing groundwater quality and flow in the well fields.


Question: loss of coolant accident (LOCA)

The HPCS System is the only high pressure system assumed to function during a loss of coolant accident (LOCA).

The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.


Question: Monitoring well W-39, which is designed to assess groundwater levels, is located within two distinct aquifer zones.

Monitoring well W-39 is screened within the upper and lower surficial aquifer.

The floodplain aquifer monitoring well location is critical for evaluating surface water interactions.


Question: NIS calibration procedures


Question: What is the significance of the nuclide MDA in regulatory filings?

Nuclide MDA (Minimum Detectable Activity) is a critical parameter that indicates the lowest level of activity that can be reliably detected for a particular nuclide in a sample, ensuring compliance with regulatory standards.

The Peak Locate From Channel is an important procedure that helps in identifying specific energy ranges of interest in gamma spectroscopy analyses.


Question: separability of heat transfer mechanisms

The second reason for using rod[s] from [stainless steel] than [zircaloy] rods involves the problems of simplifying heat transfer analyses by separating the [metal-water] reaction from the physical processes of cooling rods which were not undergoing [a metal-water] reaction. It was assumed that the [metal-water] reaction was an independent heat input mechanism to the fuel rods, separable from the basic heat transfer process of cooling.

The NRC also stated, regarding the results of FLECHT run 9573, that the AEC Commissioners were not “concern[ed] about the zirconium-water reaction models.”


Question: What is the relevance of energy calibration in gamma spectrum analysis?

Energy Calibration Used Done On: 7/8/2014


Question: What is the purpose of a survey unit in the context of radiation safety?

A survey unit is defined as a designated area where radiological surveys are conducted to assess the levels of radiation and ensure compliance with safety regulations.

The approval process for submitting regulatory documents often involves multiple layers of review, requiring signatures from various departments and specialists.


Question: chain of custody in sample handling

Sample containers received in good condition.


Question: What are the Surveillance requirements for nuclear activities?

Verify flow is within limits Once within 12 hours prior to starting activity AND 24 hours thereafter.

Each time the example activity is to be performed, the Surveillance must be performed within 12 hours prior to starting the activity.


Question: The NRC mandates that each reactor must possess a valid operating license to ensure that all safety procedures are meticulously adhered to.

License No. NPF-47 is the operating license associated with River Bend Station, Unit 1, which is regulated by the NRC.

The project included the installation of new cooling systems to improve the plant’s efficiency during peak demand periods.


Question: sensitive information licensing procedures

All licensing applications, except for portions containing sensitive information, will be made available for review in the U.S. Nuclear Regulatory Commission (NRC) Public Document Room and electronically at the NRC Library.

The applicant or licensee should identify, mark, and protect sensitive information against unauthorized disclosure to the public.


Question: The NRC mandates that operational safety involves rigorous scrutiny of the systems that monitor conditions within the plant.

PLANT SYSTEMS BASES SOLID RADWASTE SYSTEM involves systems that process and manage solid radioactive waste to ensure safe containment and disposal.

PLANT SYSTEMS BASES FIRE RATED ASSEMBLIES are structures designed to withstand fire exposure while maintaining their integrity during an incident.


Question: diesel generator testing procedures

Tests have been conducted by the manufacturer to document starting and load acceptance reliability on a similar 12-cylinder 38TD8 1/8 engine generator set. The diesel generator was connected directly to a 3000 kW resistive load and automatically set to sequence as follows:

Frequency during the initial loading does decrease below 95% of nominal and is not restored to within 2% of nominal in less than 40% of the initial load block time interval.


Question: What roles are included in the Emergency Response Organization for nuclear facilities?

- **Station Emergency Manager**  - **Chief Technical Spokesperson**    - Dominion Energy Liaison    - Dominion Energy Technical Advisor - **Technical Support Manager**  - **Radiological Assessment Coordinator**    - Accident Assessment Team      - Accident Assessment Team Leader      - Dose Assessor      - Dose Assessor      - Field Team Radio Operator      - Reactor Core Analysis      - HPN Communicator  - **Operational Support Coordinator**  - **Resource Support Manager**  - **Nuclear News Manager**  - **News Team**

The NRC requires a thorough review of safety protocols before allowing any new experimental reactors to operate.


Question: natural uranium radioactive decay

Uranium is widely distributed around the world with an average concentration in the earth’s crust of 4 PPM. Uranium is a heavy metal and is naturally radioactive. Natural uranium contains isotopes: 238-U (99.3%), 235-U (0.7%), and 234-U (0.006%). 238-U constitutes one of the main primordial radioactive decay series and has a long radioactive decay half-life of 4.5 billion years.

The eluant is then acidified with HCl, breaking the bicarbonate complex and forming UO₂Cl₂. This is precipitated with hydrogen peroxide (H₂O₂), forming hydrated UO₄ as described in section 3.7.


Question: What does RIPB mean in nuclear regulation?

The NRC’s efforts to establish a modern regulatory framework, including the proposed Part 53 and Subpart H, do not fully reflect risk-informed and performance-based (RIPB) concepts as they are defined in long-standing Commission policy.

To enable the safe and secure civilian use of nuclear energy and technologies, the NRC must break from conventional thinking and practices and deliver an efficient and effective regulatory framework.


Question: What is the impact of DC power availability on restoration actions in nuclear plants?

Long term DC power to perform restoration actions is provided from the batteries.

Common cause failures and the potential for recovery are treated explicitly in the modeling of support system initiating events.


Question: What are the licensing requirements for commercial nuclear plants?

The licensing process for commercial nuclear plants involves a thorough evaluation of safety, environmental impact, and compliance with regulatory standards set forth by the NRC.

The handout package includes various performance measure tools applied during the NRC exam process to evaluate candidates.


Question: What is a safety analysis report in nuclear regulation?

A Safety Analysis Report (SAR) is a comprehensive document that describes the safety of a nuclear facility and presents the findings of safety analyses and evaluations as required by the NRC.

The construction of the reactor must comply with all local zoning laws and requirements, ensuring that the infrastructure supports the operational mandates established by the governing bodies.


Question: What are the acceptable levels of radiation exposure for nuclear workers?

The NRC establishes limits on occupational radiation exposure to protect workers from the harmful effects of radiation, setting a maximum permissible dose at 50 millisieverts (mSv) per year.

The reactor coolant system must be monitored regularly to ensure proper functioning and prevent potential issues that could affect plant safety.


Question: What is the significance of rod worth minimize (RWM) during reactor startup?

C. Rod worth minimize (RWM) inoperable during reactor startup. | C.1 Suspend control rod movement except by scram.
 OR
 C.2.1.1 Verify ≥ 12 rods withdrawn.

D. RWM inoperable during reactor shutdown. | D.1 Verify movement of control rods is in accordance with BPWS by a second licensed operator or other qualified member of the technical staff.


Question: What components are reviewed in the Off Gas System for aging management?

The aging management review results for these components are provided in: Table 3.3.2-11 Off Gas System Summary of Aging Management Evaluation.

The table lists various component types and their intended functions essential for the operational integrity of the system.


Question: resident population projections near nuclear plants

The 2010 population for the South South East-Residential sector within 10 miles of the reactor was projected to be 108, decreasing slightly to 107 by 2020 and expected to be 105 in 2030.

The nuclear facility's cooling system is designed to handle a variety of thermal loads during operation.


Question: melt pool temperature control

Melt Pool Temperature Control for Laser Metal Deposition Processes—Part I: Online Temperature Control discusses methods for real-time management of the thermal state during laser deposition processes to ensure high-quality outcomes.

A review on measurement science needs for real-time control of additive manufacturing metal powder bed fusion processes highlights various measurement techniques that support the overall production quality but do not specifically address melt pool temperature management.


Question: Geophysical surveys have been carried out to understand the crustal dynamics of the Reelfoot Rift and its implications for safety assessments.

Extensive geophysical investigations of the Reelfoot Rift have been completed for a variety of purposes including deep crustal dynamics, oil exploration, active tectonics, and geotechnical projects (Reference 92).

During Precambrian to Cambrian time, the Reelfoot Rift formed as a result of continental rifting and crustal extension of the North American continent (Reference 27).


Question: seiche flood hazard

A seiche is an oscillation of the water surface in an enclosed or semi-enclosed body of water initiated by an external cause. Once started, the oscillation may continue for several cycles; however, over time it gradually decays because of friction (Reference 2.5-1, Section 3.6).

The potential flooding hazard from a seiche at the Cooper Nuclear Station (CNS) is judged to be negligible because of the site's riverine setting.


Question: What does the term 'Turbine Building' refer to in nuclear reactor safety?

The Turbine Building houses the steam turbines and main generators, as well as other auxiliary equipment and piping systems for SPS Units 1 and 2.

The primary reinforced concrete slab is the operating deck at elevation 58'-6". Below the operating deck is the mezzanine floor and basement.


Question: The NRC oversees various regulatory submissions concerning nuclear materials and their safety protocols.

NOTE: If the Rec% column is null, the high/low limits are in the same units as the result. If the Rec% column is not null, then the high/low limits are in % Rec.

| ACZ ID      | Type  | Analyzed  | PCN/SCN  | QC  | Sample | Error | LLD  | Found | Error | LLD  | Rec% | Lower | Upper | RPD/RER | Limit | Qual |


Question: cracking due to expansion from reaction with aggregates

Cracking due to expansion from reaction with aggregates could occur in inaccessible areas of concrete elements of PWR and BWR concrete and steel containments.

The exterior concrete of the containment subject to the outdoor environment is not exposed to temperatures of 32°F or less for sufficient durations that would cause freeze-thaw aging effects to occur.


Question: What is a TER in the context of nuclear regulatory documentation?

A Technical Evaluation Report (TER) is a document that provides an evaluation of a specific subject related to a proposed regulatory action and serves as a reference in the NRC's Safety Evaluation report (SE).

Ngola, I would typically list it as a reference in the SE if it is of sufficient quality. If not, I would not reference it.


Question: What is the COBRA-FLX model used for in nuclear safety analysis?

The COBRA-FLX code is a thermal-hydraulic analysis tool designed for the evaluation of critical heat flux conditions in nuclear reactor systems.

Table 7-18 presents the overall statistics related to the combined data set, detailing various parameters and their correlation coefficients from the study.


Question: The licensing process for nuclear facilities involves a comprehensive safety review by the NRC that evaluates nuclear reactor operations.

Amendment No. 23 to Facility Operating License No. R-76 for the Washington State University Modified TRIGA Nuclear Reactor has been issued to align with regulatory requirements as stipulated by the NRC.

The facility has undergone various upgrades and changes over the years, ensuring compliance with modern safety standards.


Question: The regulation emphasizes that the safe shutdown process in BWRs must not imply only two paths with identical functions.

This final paragraph of this section includes new language suggesting that there will be two redundant trains, that they will perform “the same” safe shutdown functions, and that they are only differentiated by electrical power division. That is not the case in BWRs, where ECCS system functions and capacities are very diverse, there are many possible safe shutdown success paths, and safe shutdown paths can be made from high-capacity high pressure systems, low-capacity high pressure systems, low pressure systems, etc.

The emergency lighting and portable lighting sections discuss “sealed beam” lamps. The term “sealed beam” specifically describes a specific type of incandescent lamp manufacturing technology that was predominant prior to the 1990’s for automobile headlights. Retaining the term in the RG effectively prohibits modern lighting technology from being adopted. For example, “sealed beam” would not permit a replaceable halogen bulb in a reflector.


Question: heat transfer of multi-rod bundles in LOCA scenarios

Clearly, the Cathcart-Pawel equation is an equation for predicting the oxidation kinetics of Zircaloy specimens in furnaces; it is not adequate for predicting the oxidation kinetics of Zircaloy bundles in a nuclear power plant core in the event of a LOCA.

I therefore believe that serious consideration should be given first to cross-checking different codes and then to verifying ECCS computations by experiments on large scale and, if necessary, on full scale.


Question: The effects of traffic and workforce size on transportation near nuclear facilities remain minimal according to the SLR assessment.

Because no changes to employment are expected from continued operations of DNPS, and no potential refurbishment activities have been proposed that would require a larger workforce, the people living in the DNPS region are not likely to experience any changes in socioeconomic conditions due to the plant during the SLR term beyond the current conditions.

Transportation impacts depend on the size of the workforce, the capacity of the local road network, traffic patterns, and the availability of alternate commuting routes to and from the plant.


Question: NIST SP 800-53 security controls

The security controls in NIST SP 800-53 are particularly suitable for tailoring for ZTA at nuclear facilities because this NIST guidance is updated every 3-4 years, which is more frequently than other security standards and guidance documents, and thus the controls in NIST SP 800-53 are timely and can be used to implement the most current, best security practices.

These tailored selected sets of cybersecurity controls may not include controls listed in general cybersecurity guidance, such as NIST SP 800-53 Security and Privacy Controls for Information Systems and Organizations, and consequently are not implemented in the licensee’s CSP.


Question: What are the design basis considerations for independence and single failure in nuclear safety?

The concepts of independence and single failure, these are design basis considerations covered under separate design basis requirements.

Multidivisional control and display stations should be qualified to withstand the effects of adverse environments, seismic conditions, EMI/RFI, power surges, and all other design basis conditions applicable to safety-related equipment at the same plant location.


Question: What does LCH 480V A represent in nuclear systems?

LCH 480V A refers to a specific low voltage AC output power supply system that is crucial for the operational integrity of safety systems in a nuclear plant.

The 4KW FAST ELEXER BLOCKED indicates a malfunction in a different electrical component which does not relate to the safety systems powered by LCH 480V A.


Question: What are the qualifications for a Radiation Protection Manager in nuclear facilities?

The Radiation Protection Manager shall be a supervisor with line responsibility for operational health physics who meets or exceeds the qualifications of USNRC Regulatory Guide 1.8, September 1975, for a Radiation Protection Manager.

Each member of the unit staff shall meet or exceed the minimum qualifications of ANSI/ANS 3.1-1978, with the following exceptions:


Question: FLEX feed pumps failure rates

The industry average baseline values from NUREG-6928 were used for FLEX feed pumps failure rates.

Supporting requirements for HLR-DA-D (ASME Standard, RA-Sa-2009) are addressed in Table E9-6 for the modeling of the Flex Portable Equipment.


Question: The NRC examiners assessed the qualifications of applicants for licensing as reactor operators at the Comanche Peak facility.

The NRC examiners determined that eleven of the fourteen applicants satisfied the requirements of 10 CFR Part 55, and the appropriate licenses have been issued.

The written examination was administered by the licensee on October 31, 2024.


Question: The NRC conducts research to enhance the understanding of high burnup spent nuclear fuel and its implications for storage.

Update on spent nuclear fuel research activities, including extended storage, high burnup spent nuclear fuel, and other future-focused activities


Question: Nuclear facility operators must apply for licenses that involve thorough safety evaluations and compliance with regulations.

The licensing process for a nuclear facility includes comprehensive assessment of its safety protocols and environmental impact.

The reactor type generally utilized for commercial power generation in the United States is the Pressurized Water Reactor (PWR), which uses water under high pressure as a cooling medium.


Question: The meteorological data collected during November shows average wind speeds recorded at the GGNS site, with varying frequencies based on wind direction.

Percentage frequency of wind direction and speed (MPH) during November, GGNS site is detailed in Table 2.3-42, indicating the distribution of wind speeds across various directional categories.

The NRC continuously reviews applications for new reactor licenses to ensure compliance with existing regulations.


Question: The NRC requires licensees to perform volumetric examinations of reactor vessel pressure-retaining welds during each Inservice Inspection interval...

Category B-A welds are defined as "Pressure Retaining Welds in Reactor Vessel."

Volumetric examinations shall be performed for the Sixth ISI interval in accordance with requirements of 10 CFR 50.55a and the Edition and Addenda of the ASME Code, Section XI, applicable at the time of the examination.


Question: The NRC specifies that aging management programs must address the potential degradation of components such as piping and structural materials.

AMP XI.M38, "Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components"

Not Applicable. There are no stainless steel piping or piping components exposed to concrete in the Steam and Power Conversion System.


Question: Isolation devices in high radiation zones should have their functionality confirmed through proper administrative procedures.

Isolation devices in high radiation areas may be verified by use of administrative means.

Verify the affected penetration flow path is isolated.


Question: The analysis of atmospheric stability is crucial for understanding environmental impacts related to nuclear facilities.

Based on Pasquill-Turner calculation of E, F or G from hourly surface observations.

The table lists the mean number of inversion occurrences recorded at various times throughout the year.


Question: What is the Charpy V-notch impact test and why is it important?

The Charpy tests are performed to determine the transition temperature increases at 30 ft-lb and 50 ft-lb points, and the decrease in upper shelf energy.

The calculated fast neutron fluence based on actual cycle power distributions at the capsule location was 1.06 x 10^19 n/cm², which compares favorably (within 11%) with the dosimeter fluence.


Question: regulatory filings nuclear power

The table summarizes the compliance and operational data for nuclear reactors at various levels of power output, measuring both the frequency and magnitude of incidents across different reactor types and geographical locations.

The structure of the nuclear reactor plays a crucial role in determining its overall efficiency and safety characteristics.


Question: What is the mitigative option in nuclear safety regulations?

The mitigative option involves crediting plant mitigative features to protect safety-related SSCs from failures of nonsafety-related SSCs.

Failure in the nonsafety-related piping beyond the above boundary locations would not impact structural support for the safety-related piping.


Question: What operational pressures are necessary for the SDC to function effectively?

NRC: <br> - SDC requires <100 psig RPV pressure. Mechanical vacuum pump can't be started until after >100 psig.

NRC:
 - The reference (T.S. 3.7.b) should not be redacted ... T.S. information 'below the line' is not required to be memorized (references must provide 'an effective and objective way for the applicant to demonstrate knowledge of the topic or concept'). (ES-4.2, B.6.d)
 - Consider adding a failure of an 'important' yet non-T.S. D.C. bus to the stem ... this may make Distractor 'C' more plausible (without giving anything away once the reference is given without redaction).
 Response: 1-hour or less action statements are required to be memorized for ROs and SROs.


Question: low population zone in nuclear regulatory context

The low population zone (LPZ) for VEGP Units 3 and 4 is the same as the LPZ for the existing VEGP units and consists of the area falling within a 2-mi radius of the midpoint between the VEGP Unit 1 and Unit 2 containment buildings.

The nearest population center to the VEGP site with more than 25,000 residents is the City of Augusta, Georgia, with a 2000 population of 195,182 (USCB 2000b). Augusta is approximately 26 miles north-northwest of the VEGP site.


Question: What is the purpose of the Certificate of Compliance in nuclear shipping?

Industrial Nuclear Company (INC) hereby submits this application for a Certificate of Compliance (C of C) for the INC Outer Pack-Raw Material Shipping Container (OP-RMSC) package.

The enclosed SAR includes all test data and analyses for the certification of this shipping container package as required by Title 10, Energy, Code of Federal Regulations, Part 71 (10 CFR 71), Packaging and Transportation of Radioactive Material, 1-1-2018 Edition.


Question: aging management program for reactor vessel internals

Aging Management Program: BWR Vessel Internals (B.2.1.7) indicates the specific items and evaluations related to managing aging effects in reactor vessel internals.

Jet Pump Assemblies: Castings (Inlet elbow, Mixing assembly, Diffuser casting, Restraint Bracket) serve a Direct Flow function and are made from Cast Austenitic Stainless Steel (CASS) in a Reactor Coolant and Neutron Flux environment.


Question: Uranium isotopes analysis


Question: What is the significance of nuclide activity in nuclear safety assessments?

Activity (pCi/M²) refers to the measure of the number of decays of a radioactive nuclide per unit area, which is crucial for assessing radiological hazards in nuclear safety evaluations.

Co-60 has a yield of 100.00% and an energy level of 1173.22 keV, making it relevant for various industrial applications.


Question: What are the components maintained in the pump section of a nuclear facility?

The TRAIN & PUMP SECTION IN MAINTENANCE includes various components such as GATE VALVE 16, GATE VALVE 19, and CHECK VALVE among others that are subjected to regular maintenance checks to ensure operational safety.

The turbine driven pump is crucial for maintaining pressure and flow in nuclear reactors and is regularly inspected to prevent failures.


Question: The analysis must consider the potential risks associated with materials that may impact the control room environment.

The potential effect of an anhydrous ammonia release on control room habitability was evaluated, and it was determined that the probability of an event at this location resulting in uninhabitability of the control room was sufficiently low so as to allow this source to be eliminated from concern.

The 19 materials and their respective quantities (for example, maximum quantity per shipment) were evaluated in the CPS USAR for impact on control room habitability.


Question: how to determine the ACTIVE/INACTIVE status of a reactor operator's license

The applicant determines RO A’s license is INACTIVE

The shift period is defined by the schedule worked by the rotating shift crews.


Question: chemical waste management process

Laboratory samples which contain reagent chemicals (and possibly tritiated liquid) are discarded through a sample room sink which drains to the chemical drain tank. Chemical drains requiring radwaste processing are sent to the waste management system or may be processed through the radwaste processing facility demineralizers and/or the radwaste processing facility filtration system.

Another source of water to the floor drain tank is the chemical laboratory drains. Excess nonreactor grade samples that are not chemically contaminated and laboratory equipment rinse water are drained to the floor drain tank.


Question: What are the minimum and maximum limiting ages for liquefaction features?

Min Age (cal. yr B.P.)¹ and Max Age (cal. yr B.P.)¹ are presented for various sites such as Ashley County, Portland Site, and Morgan Site, indicating the chronological range of geological features identified in these areas.

Data from Cox et al. (2004), Cox and Larson (2004), and Cox (2004, personnel communication) provides insights into geological studies conducted in the region.


Question: What is meant by molten material in nuclear reactor cores?

Video data from the interior of the core bore holes revealed a region of previously molten material directly below the hard crust and a region of intact standing fuel rods extending from the bottom of the previously molten region to the bottom of the core.

The drill cuttings or sandy debris collected at various depths provide insight into the physical conditions of the reactor core but do not necessarily indicate the state of the fuel.


Question: The analysis of licensing requirements for the emergency operating procedures ensures safety in reactor operation.

Licensing requirements for emergency operating procedures (EOPs) are critical to ensuring the safety of nuclear reactor operation, especially during abnormal conditions.

The operator must ensure proper communication with the control room during normal operation to avoid miscommunication during critical events.


Question: The regulatory guidelines specify minimum enrichment levels for reactor fuel assemblies to ensure safety and efficiency.

Minimum Initial Assembly Avg. Enrichment values are critical for determining the fuel design and operational safety of various PWR assembly types.

In addition to fuel assembly designs, the reactor's cooling system capacity must also be robust to maintain safe operating temperatures.


Question: What is a cooling water reservoir in nuclear power plants?

The NRC staff concluded that construction of the cooling water reservoir required to supply water to the cooling system would result in MODERATE impacts on the aquatic environment.

Water quality permits required through Federal and State regulations would control, reduce, or mitigate potential effects on the aquatic environment.


Question: What are the regulatory requirements for emergency response data systems in nuclear power plants?

Outlines a set of system, testing, and implementation requirements for the emergency response data system (ERDS) for operating nuclear power reactor licensees, and § 50.72, “Immediate notification requirements for operating nuclear power reactors,” includes requirements for activation of ERDS.

For example, SMRs or ONTs may need to be prepared for events associated with contiguous or nearby facilities’ hazards.


Question: The actuation logic for the Engineered Safety Feature Actuation System (ESFAS) must meet specific criteria outlined in section 3.3.15.

The ESFAS Actuation Logic is described in detail in section 3.3.15 of the regulatory guidelines, ensuring the system can respond effectively during critical situations.

The frequency of operations for various surveillance activities must be strictly followed, even when no specific requirements are indicated.


Question: pressure boundary evaluations

Reactor coolant pressure boundary components: piping, piping components; other pressure retaining components with fatigue analyses | Pressure Boundary | Stainless steel | Reactor coolant | Cumulative fatigue damage: cracking due to fatigue, cyclic loading | Section 4.3 "Metal Fatigue"

Class 1 valve bodies and bonnets | Pressure Boundary | Cast austenitic stainless steel | Reactor coolant >250°C (~482°F) | Loss of fracture toughness due to thermal aging embrittlement | ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD (B.2.1.1)


Question: regulatory history of the Oconee plant

Petitioners’ desire to present even further detail on the regulatory history of the Oconee plant or Petitioners’ generalized grievances therewith far exceeds the purpose of an appeal.

As the Commission has explained, “[o]ur page limits on appeal briefs are intended to encourage parties to make their strongest arguments as concisely as possible.”


Question: pertinent information program

Technical References: OP-AA-102-104 (PERTINENT INFORMATION PROGRAM)

**Cognitive Level:** Higher  ☐  Lower   ☑


Question: What are the characteristics of the Tuscaloosa Aquifer System?

Some wells in the Hopkins area (Grid 29Q, Fig. 2-3) and many small-diameter domestic wells northeast of the WEC site (Grid 29P) tap artesian sand aquifers within the Tuscaloosa Aquifer System.

To the southeast, where Tuscaloosa aquifers thicken and are more permeable, several large-capacity irrigation wells (26R-1, 27Q-1) tap this aquifer system.


Question: instruments operable with only dc battery power

Instruments and controls operable with only dc battery power available are crucial during a Station Blackout to ensure continued monitoring and safety of the reactor system.

The ability of operators to evaluate feedback from safety systems during normal operations is essential for maintaining plant integrity and preventing incidents.


Question: The flow rate measurements for riser holes is critical for assessing reactor operations.

Total riser hole flow is an important metric that integrates the flow contributions from all holes combined, providing insights into reactor performance.

The cooling system operates by circulating water through the reactor core to manage temperatures during operations.


Question: What is the significance of extravasation in nuclear medicine?

Extravasation of diagnostic radiopharmaceuticals is common.

The Society of Abdominal Radiology and American College of Radiology co-developed a “practice quality improvement project” to study extravasations that occur during CT exams.


Question: LOCA and LOOP probabilities

It is suggested that the "low probability" prediction should be strengthened to also include CCF of DG fail-to-start. Also, LOCA with consequential LOOP may be more probable than independent LOCA and LOOP.

Based on this F&O being resolved except for documentation updates which have no quantitative impact on the PRA model, this open F&O has no impact on the GSI-191 evaluation.


Question: What is a reactor coolant system?

The Reactor Coolant System includes the Pressurizer Safety Valves and Pressurizer Power Operated Relief Valves, which are essential components in managing the pressure and cooling of the reactor core.

The Containment Spray System serves to reduce the temperature and pressure within the containment building during a loss of coolant accident by spraying water inside the containment.


Question: What is the purpose of the Resources Databook in nuclear emergency planning?

The Resources Databook (RDB) is an extremely useful tool in an emergency at River Bend Station.

Volume 1 also contains data by Protective Action Section which includes general population data, listing of individuals requiring transportation and special needs, transportation needs of facilities requiring additional vehicles to accomplish evacuations, listing of transient facilities and evacuation time estimates.


Question: What are the specifications for unirradiated fuel assemblies used in Yankee Class?

Length: Maximum = 111.8 inches, Minimum = 109.0 inches.

The original fuel assemblies for the INTACT FUEL RODS, DAMAGED FUEL RODS, FUEL DEBRIS and damaged fuel cans shall meet the criteria specified in Table B2-2 for the fuel assembly class, and meet the following additional specifications.


Question: What are engineered safety features in nuclear reactors?

Engineered safety features are systems designed to prevent or mitigate the consequences of accidents by ensuring that the reactor remains in a safe condition under a variety of conditions.

A comprehensive overview of the licensing process includes submission deadlines and the requirements for documentation, aiming to streamline operations successfully.


Question: Nuclear regulatory control is essential for ensuring the proper management of nuclear waste.

Keep nuclear waste under nuclear regulatory control and make the polluters pay to try to isolate it--not release it!

This proposal is even more bold and expansive than the many past efforts by the nuclear industry and its 'regulators' which Congress revoked in the 1992 and the public has repeatedly rejected over the decades of the nuclear age.


Question: What are the material properties of ASTM A-572 Gr. 60?

Material properties for ASTM A-572 Gr. 60 include an elasticity modulus (E²) of 29.4 (10³ ksi) at 70°F, yield strength (Sₜ) of 60.0 ksi, and tensile strength (Sₑ) of 75.0 ksi.

The reactor core is designed to achieve optimal neutron flux for efficient nuclear fission.


Question: The NRC mandates that license holders must monitor radioactive emissions in various environments.

Gamma radiation measurements are crucial for determining the levels of radioactivity in a given environment, as reflected by consistent monitoring and compliance with safety standards.

Technical specifications require regular calibration of measurement instruments to ensure accurate performance and adherence to regulatory requirements.


Question: What is the significance of K-40 in radioactive measurements?

The NRC requires detailed analysis of the environmental impact assessments conducted during the licensing process.


Question: What is the significance of energy measurements in nuclear safety assessments?

The Peak Centroid is a key parameter in nuclear spectroscopy, providing the energy at which the maximum counts occur for a specific peak, and is typically measured in kiloelectronvolts (keV).

Errors quoted at 2.00 sigma are used to ensure the statistical validity and reliability of the measurements obtained during experiments.


Question: The NRC mandates that licensees evaluate the safety of operating in extended burnup conditions in their regulatory filings.

In this report, we apply the SSP CMS5 methodology to Pressurized Water Reactors (PWRs) for extended enrichment and burnup analyses relevant to Small Modular Reactors (SMRs).

The validation of numerical codes ensures they accurately predict reactor behavior under various operational scenarios.


Question: operable diesel generators

Required Action B.3.1 provides an allowance to avoid unnecessary testing of OPERABLE DGs. If it can be determined that the cause of the inoperable DG does not exist on the OPERABLE DGs, SR 3.8.1.2 does not have to be performed.

Operation may continue in Condition B for a period that should not exceed 14 days. Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time Program.


Question: What are the components of the steam generator boundary in a nuclear reactor?

Table 5.2-9 lists the Steam Generator Primary/Secondary Boundary Components, detailing descriptions and maximum fatigue usage for conditions such as Normal, Upset, and Test.

The Tubesheet Center component has a maximum fatigue usage of 0.16 under normal conditions, indicating its structural integrity during regular operation.


Question: What is online monitoring in the context of nuclear reactor regulation?

The licensee, SNC, proposes to use online monitoring (OLM) methodology as the technical basis to switch from time-based surveillance frequency for channel calibrations to a condition-based calibration frequency based on OLM results.

The NRC staff from the Office of Nuclear Reactor Regulation (NRR) has initiated its review of the LAR in accordance with NRR Office Instruction LIC-101, "License Amendment Review Process" (ML19248C539).


Question: nuclide activity measurement

The NRC ensures that all safety systems are tested and verified to operate correctly under all conditions.


Question: What actions are necessary to establish feedwater flow in a nuclear plant?

12. ESTABLISH feedwater flow: | BOP | | a. START secondary plant pumps as necessary: | | 1. Hotwell pumps. | | 2. Condensate booster pumps. | | 3. Cond DI booster pumps.


Question: What is the purpose of the Initial Prehearing Conference at the NRC?

The Board issued an Order scheduling the Initial Prehearing Conference for February 12, 2025, at NRC Headquarters.

Joint Petitioners believe that personal interactions with these participants are essential to ensure that the Board fully appreciates the concerns and evidence that Joint Petitioners intend to present.


Question: The NRC mandates that all licensees must conduct thorough assessments of their operational safety protocols.

Nuclear reactor licensees are required to submit a safety analysis report that demonstrates compliance with all applicable regulatory requirements.

The author of this section has provided insights on document approval timelines and the validation process of records.


Question: The NRC mandates that companies must provide thorough assessments of radioactive materials present in soil samples.

The moisture content of the soil sample was determined to be 9.77%.


Question: The NRC requires a comprehensive aging management program for components exposed to reactor coolant.

The BWR Vessel Internals (B.2.1.7) and Water Chemistry (B.2.1.2) program will be used to manage cracking of the nickel alloy core shroud and core plate access hole covers and steam dryers exposed to reactor coolant and neutron flux in the Reactor Vessel Internals.

An exception applies to the NUREG-2191 recommendations for Water Chemistry (B.2.1.2) program implementation.


Question: The NRC mandates that licensees submit detailed analyses concerning safety measures for nuclear operations.

The safety analysis report must demonstrate that the facility can operate within regulatory limits and that measures are in place to mitigate any potential hazards.

In the results section, various statistical methods are employed to interpret the data collected during the experimentation phase.


Question: Residual Heat Removal System


Question: The TSC must be equipped with dependable communications systems to interact with the MCR, OSC, EOF, and various local and state agencies during emergencies.

The habitability capabilities of the proposed TSC as described above provide for the radiological protection of TSC personnel consistent with the guidance provided in NUREG-0696 [Reference 1] and NUREG-0737, Supplement 1 [Reference 2].

Revision 15 of the Emergency Plan [Reference 7] identified several communications systems available to TSC personnel for communications with the MCR, OSC, EOF and NRC. These included commercial.


Question: The regulatory framework mandates that all nuclear facility operators must secure appropriate licenses and permits before initiating operations.

Before operating a nuclear facility, licensees must obtain a license that is granted by the NRC, which includes demonstrating compliance with safety regulations and technical requirements.

The use of advanced cooling systems in nuclear reactors significantly improves efficiency and reduces operational costs across various power plants.


Question: What is the purpose of an interim notification under 10 CFR Part 21?

Our present goal is to complete our evaluation by February 5, 2021 at which time a determination will have been made as to whether a substantial safety hazard exists.

If there are any questions or if additional information is required, please contact the writer using the contact information provided above or by email at dbrown@lakeengineeringri.com.


Question: ASME Code for Operation and Maintenance of Nuclear Power Plants

Specification 4.0.5 establishes the requirement that inservice testing of ASME Code Class 1, 2, and 3 pumps and valves shall be performed in accordance with a periodically updated component of the ASME Code for Operation and Maintenance of Nuclear Power Plants (ASME OM Code) and applicable Addenda as required by 10 CFR 50.55a.

The purpose of this specification is to ensure that system and component OPERATIONAL requirements or parameter limits are met before entry into a MODE or condition for which these systems and components ensure safe operation of the facility.


Question: nuclear reactor regulation

The Office of Nuclear Reactor Regulation is responsible for ensuring the safety and compliance of commercial nuclear power plants according to federal regulations.

The Chief Information Officer oversees the administration of information technology services and systems within the agency.


Question: Turbine Impulse Transmitter failure

Checks rod position above low-low rod insertion limit.


Question: transuranic waste disposal

In early 1976, ERDA officials decided to build the Waste Isolation Pilot Plant (WIPP) in a bedded salt deposit near Carlsbad, New Mexico, to be used for disposal of transuranic wastes from the defense program and for the performance of R&D with other waste materials in salt.

While ERDA was occupied with nuclear energy challenges inherited from the AEC, other energy legislation passed in 1974 compelled it to diversify its energy R&D.


Question: device not subject to registration

Devices that are not subject to registration include those that are not typically categorized under the regulatory guidelines set forth by the NRC, which allows for operational flexibility in certain scenarios.

The Manufacturer License No is essential for tracking and managing registered devices within the regulatory framework.


Question: What is the purpose of a security baseline inspection in nuclear power plants?

On December 4, 2020, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Quad Cities Nuclear Power Station and discussed the results of this inspection with Mr. Ken Ohr, Site Vice President and other members of your staff.

This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with Title 10 of the Code of Federal Regulations 2.390, “Public Inspections, Exemptions, Requests for Withholding.”