Question: measurement uncertainty analysis in nuclear regulation

With regards to the second part of RAI-10d, it is recalled that measurement uncertainty recommended analysis does not appeal to the classical statistical processing of a finite dataset for regression analysis. In cases when experimental measurement (equipment, input data, etc.) uncertainties are known, an uncertainty propagation analysis is carried out to combine measurement uncertainties with all other known uncertainties.

Another case in point is the presentation of the thermal expansion data. As illustrated in Figure 10-19, also taken from Reference 4, the uncertainty bounds cover all the data considered in the analysis. The thermal expansion dataset is relevant to the current discussion because it contains a sub-set (Christensen data) that shows a systematic larger deviation than the rest of the data.


Question: The licensing of nuclear plants is a complex process involving multiple regulatory assessments.

Nuclear plant licensing was halted for approximately 18 months while the AEC addressed the decision.

A Tough Manager Takes Over at AEC, Business Week, July 24, 1971, 20.


Question: What are the NRC Review Guidance paths for resolving in-vessel downstream effects?

The NRC Review Guidance for the resolution of in-vessel downstream effects provided four different paths (identified as Box 1 through Box 4 paths) that PWR licensees can use to resolve the issue based on the AFP analysis in WCAP-17788-P, Revision 1.

The fiber that passes through the strainer is split based on the ratio in flow rate between the emergency core cooling system (ECCS) pumps and CS pump.


Question: What are Risk Management Assessments (RMAs) in nuclear safety?

Determining RMAs involves the use of both qualitative and quantitative considerations for the specific plant configuration and the practical means available to manage risk.

Use temporary equipment to provide backup power, ventilation, etc.


Question: picocuries per gram units

Results for the requested radionuclides of concern were provided in units of picocuries per gram (pCi/g).

Each container was labeled and security sealed in accordance with ORISE procedures.


Question: What is RAI-34 in the context of nuclear regulatory filings?

RAI-34 will be addressed at a later date.

This report discusses the various safety analyses that must be conducted during the licensing process.


Question: The analysis of critical heat flux is essential for ensuring reactor safety during operation.

COBRA-FLX, a methodology used to analyze the critical heat flux, ensures that the reactors operate safely under various conditions.

The design of the reactor core must account for neutron flux distribution to optimize performance.


Question: What are the environmental considerations for the subsequent license renewal of nuclear plants?

The Applicant’s Environmental Report examines various environmental factors related to the proposed continued operation of the Monticello Nuclear Generating Plant Unit 1, including effects on local wildlife and water resources.

The Monticello Nuclear Generating Plant Unit 1 is located within a 50-mile radius of several urban areas, affecting local infrastructure and emergency response capabilities.


Question: The NRC stipulates specific procedures for analyzing the gamma spectrum of samples.

The efficiency calibration for the gamma spectrum analysis was performed on 7/8/2014.

The sample size recorded for this analysis was 792 grams, which falls within standard parameters.


Question: What is the significance of energy peaks in nuclear submittals?

The safety assessment must include considerations for environmental impacts and adequate financial assurance.


Question: What are the characteristics of M5 cladding in nuclear reactors?

M5 cladding, which is an alloy of zirconium, is designed to provide enhanced corrosion resistance and improved mechanical properties at high temperatures compared to traditional zirconium alloys.

Regular maintenance of nuclear reactors is crucial to ensure operational efficiency and compliance with safety regulations.


Question: What is the significance of the kappa value in seismic analysis?

The empirical kappa value to be employed is considered conservative as a deeper sedimentary column (over 10,000 ft at the site) over hard rock is expected to show larger kappa values (more cumulative damping) compared to shallower sedimentary columns.

The general lateral uniformity of the deeper velocities (deeper than several hundred feet) is consistent with other large basins, e.g. Los Angeles and Imperial Valley, California and provides a basis for employing a generic deep profile beneath the local site profile.


Question: The NRC has assessed the potential increases in disease and predation vulnerability for aquatic life due to exposure to harmful thermal conditions.

Fish and shellfish that are exposed to the thermal effluent of a nuclear power plant may experience stunning, disorientation, or injury. These sublethal effects can subsequently affect an organism’s susceptibility to predation, parasitism, or disease.

The 1996 LR GEIS (NRC 1996-TN288) and the 2013 LR GEIS (NRC 2013-TN2654) concluded that the losses from predation, parasitism, and disease among organisms exposed to sublethal stresses would be SMALL during the initial license renewal term.


Question: The 1-203-3D Relief Valve was not functioning properly during a critical situation when it should have closed.

The 1-203-3D Relief Valve spuriously opens.

Torus temperature is 90°F and RISING at ~5°/minute.


Question: fire protection system

Piping, piping components are part of the Fire Protection System and serve as a pressure boundary made of stainless steel, intended for use in raw water (internal) environments, with aging effects requiring management such as flow blockage and loss of material.

The Fuel Oil Transfer Pump's pump casing is made of carbon steel and is designed to manage fuel oil in internal and external environments, particularly regarding loss of material but is not part of the fire protection system.


Question: The facility must adhere to strict compliance regulations to maintain operational safety standards.

The regulations pertaining to nuclear licensing are crucial for ensuring that safety protocols are correctly implemented.

The external cooling systems are designed to enhance the efficiency of energy generation in various conditions.


Question: Nuclear Energy Institute submissions

The Nuclear Energy Institute (NEI) submitted NEI 20-07, Draft B, which includes guidance on the use of risk-informed decision-making in regulatory processes.

The meeting included various stakeholders from the nuclear industry, such as Duke Energy and Westinghouse, to discuss the implications of the proposed guidelines.


Question: What is the purpose of the Core Tube End Cover in nuclear systems?

The Core Tube End Cover is designed to provide structural support and integrity to the reactor core, ensuring that it can withstand seismic events and maintain safety protocols during operation.

The Substructure Post Jack Bolt and Baseplate play a critical role in securing the reactor's substructure components and ensuring stability under different load conditions.


Question: The reliance on GALILEO derived thermal-mechanical properties is essential for the fuel pin evaluation in the REA analysis.

The AREA™ methodology is limited to only the GALILEO™ derived thermal-mechanical properties of fuel pins.

The limitations conditions for each of the topical reports used in the REA analysis are addressed in this section.


Question: What is the significance of MDA values in nuclear regulatory processes?

* MDA value not calculated  * Half-life too short to be able to perform the decay correction  * CAUTION: MDA value is inconsistent with Currie MDA at 95% confidence level

The NRC requires a detailed analysis of safety systems to ensure compliance with regulatory standards.


Question: What regulations guide the safety analysis reports for nuclear power plants?

Content is driven by: 10 CFR 20, 10 CFR 100, and 10 CFR 50

NURIE-0800 provides guidelines for seismic analysis, which is essential for evaluating plant safety.


Question: isotope activity measurement

Distributor/Distributed By: Thermo Process Instruments, L.P.


Question: The components of the Core Support Shield are evaluated for their significance in inspections and categorization.

CSS top flange was classified as a Primary category component in MRP-227 Rev. 1-A, indicating its importance in safety assessments.

Audit to cover why the CSS cylinders were not included as assessed components in Table 3-1 of MRP-227 Rev. 1-A and leaving the final screening inspection categorization basis up to a license-specific determination (i.e., E or NAM).


Question: reactor vessel integrity

Events which lead to a relatively rapid and severe reactor vessel downcomer cooldown can result in a thermal shock to the vessel wall that may lead to a small flaw, which may already exist in the vessel wall, growing into a larger crack. The growth or extension of such a flaw may lead, in some cases (where propagation is not stopped within the wall), to a loss of vessel integrity.

Isolate steamflow by closing: FCV-37 (AFW S/G 1-2 supply to TD AFW Pp)


Question: nuclear activity of nuclides

K-40 has an activity of 8.55485E+004 pCi/M².

The reactor safety assessment includes multiple scenarios requiring comprehensive evaluation.


Question: The deadline for submitting abstracts for the conference has already occurred, limiting access to relevant systems.

Since the formal deadline for abstract submissions has passed, I am no longer able to access the Indico system through the PLiM website (https://conferences.iaea.org/indico/event/134/).


Question: The determination of radioisotope concentrations involves analyzing specific units such as pCi/g.


Question: The reactor operates under various neutron flux levels including high neutron flux in both the low and high ranges as part of its performance indicators.

HIGH NEUTRON FLUX (HIGH) 2/4 POWER RANGE


Question: The ECCS pumps are activated during a loss of coolant accident to ensure coolant circulation.

During a LOCA the RRHPs are utilized as low pressure ECCS pumps.

An 8” motor operated gate valve is in each CL injection header (1, 2-FCV-63-93, 94) is open for CL injection and recirc and closed by the MCR operator, for HL recirc. Body and bonnet material is SS.


Question: maximum condenser leak

The maximum condenser leak under these conditions is 27 gal/min.

The design maximum influent impurity level is operationally accommodated by the powdered resin units at normal coating coverage.


Question: 10 CFR Part 21 notifications

The inspectors determined that Business Services Company staff did not adequately review the purchase order requirements. As a result, Business Services Company staff failed to identify the part 21 during the Quality Inspection.

The inspectors reviewed the component tracking database and determined that the IR associated with the Part 21 was reasonable to identify by a qualified procurement engineer.


Question: minimum cooling time for PWR fuel

Minimum Cooling Time (years) for the assembly average burnup of 40 < Assembly Average Burnup ≤ 41 GWd/MTU for PWR fuel is detailed in the table.

The table displays the initial assembly average enrichment levels for different reactor designs.


Question: nuclear safety operational review

Operating activities that affect nuclear safety are reviewed. The review program was implemented prior to initial fuel loading and ensures review and evaluation of tests and experiments, unplanned events, and proposed changes. The program complies with the requirements of 10 CFR 50.59 relating to proposed changes, tests, and experiments.

The vice president-Vogle has responsibility for safe operation of the plant. He is kept abreast of plant operating conditions by the managers, superintendents, and supervisors who are knowledgeable of and experienced in their areas of job responsibility.


Question: nuclear reactor safety analysis

The safety analysis of the Oconee Nuclear Station, Units 1, 2, and 3, must comply with 10 C.F.R Part 50 and include evaluations of accident scenarios, including loss-of-coolant accidents (LOCA).

The certification process requires that all documents are filed electronically through the NRC’s E-Filing System.


Question: How does climate change influence the transportation of radioactive materials?

Will expected climatological changes affect the impact level assigned to the transportation of radioactive materials in Chapters 3 or 4?

Considering responses to previous questions, justify your answer in 4 sentences or less:


Question: The measurement of radiation levels is crucial for assessing safety in nuclear facilities.

Activity (dpm/100 cm³) represents the concentration of radionuclides in a given volume, important for regulatory compliance and safety assessments.

Gamma results show a wide variation across different measurements, highlighting the need for ongoing monitoring and analysis of radiation levels.


Question: What is the preservation method for total/dissolved mercury samples?

HG-1631 (total/dissolved mercury by method 1631)

The preservation of the following bottle types is not checked at sample receipt: Orange (oil and grease), Purple (total cyanide), Pink (dissolved cyanide), Brown (arsenic speciation), Sterile (fecal coliform), EDTA (sulfate), HCl preserved vial (organics), Na2S2O3 preserved vial (organics).


Question: licensing plans for nuclear facilities

We look forward to further discussion of our licensing plans with the NRC.

Please contact Mr. Don Shaw at 410-910-6878 if you have any questions or require further clarification.


Question: maximum power level in nuclear reactors

Southern Nuclear is authorized to operate the facility at steady-state reactor core power levels not in excess of 2,804 megawatts thermal.

The Technical Specifications (Appendix A) and the Environmental Protection Plan (Appendix B), as revised through Amendment No. 325, are hereby incorporated in the renewed license.


Question: plume exposure pathway emergency planning zone

...means to provide early notification and clear instruction to the populace within the plume exposure pathway Emergency Planning Zone have been established.

The alert and notification systems (ANSs) used to alert and notify the general public within the plume exposure pathway EPZ and methods of activation are described.


Question: Discuss the water balance management at the BFN site including the reliance on surface water resources.

ER Section 3.5.1.3 discusses Clean Water Act Section 303(d) water quality impairments to Wheeler Reservoir.

Please provide a time series figure illustrating the monthly average intake and discharge temperatures for the available period of record.


Question: The space surrounding the reactor vessel where coolant can flow is known as the reactor vessel annulus.

Reactor Vessel Annulus measures 0.721 FT³ for specified configurations.

The Upper Reactor Cavity is reported to hold a volume of 16,270 FT³.


Question: What is SCC in the context of aging management for nuclear systems?

The One-Time Inspection program (B.2.210) will be used to manage cracking due to SCC of the stainless steel ducting, ducting components exposed to air, condensation in the Normal Ventilation, Standby Gas Treatment, Air Conditioning and Containment Systems.

Consistent with NUREG-2191. The Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components program (B.2.124) will be used to manage loss of material of the copper alloy and stainless steel piping, piping components, and tanks exposed to waste water in the Station Drainage and Radwaste Systems.


Question: operator requalification plan

The licensee recently updated their operator requalification plan awaiting review and approval by the operator licensing examiners.

Due to reactor vessel cleaning activities that will be completed in 2025, the inspectors will continue to monitor the licensee’s performance in this area.


Question: What are the drainage zones considered for safety-related facilities at CNS?

Figure 2.1-2 shows the four drainage areas considered in the main plant area, identified as Zones A, B, C, and D. All safety-related facilities at CNS are located within these four zones.

The jersey barriers are 112.5 ft long and 32 inches (h) high and the Kontek blocks are 10 ft long and 3 ft high.


Question: The operator utilizes the phone labeled "NETS" to initiate communication with the relevant authorities.

Operator simulates the following: 1. Selects the phone labeled "NETS" 2. Depresses pushbuttons to dial 5400


Question: risk-informed high energy line break

It’s a sampling population, but it makes sure that it elevates certain pipes or certain welds that have a higher consequence of failure to just make sure that you have a certain percentage of those welds in your inspection program.

So that’s all we’re doing here, entirely analogous with what, you know, consistent programs that we’ve been doing with, you know, with that industry and the NRC have been -- that the industry has proposed and the NRC has approved.


Question: Residual Heat Removal System procedures

Swapping trains of RHR from the Reactor Coolant System can increase radiation levels significantly in areas adjacent to the oncoming RHR train’s system piping.

PRT parameters should be monitored for indications of relief valve lifting, after starting pump.


Question: What are newly identified systems under 10 CFR 54.37(b)?

Any structures and components identified in this SLR-ISG as requiring aging management that were not previously identified in earlier versions of the SRP-SLR or GALL-SLR Report are considered to be newly identified structures and components under 10 CFR 54.37(b).

The NRC staff’s position is based upon the following considerations: - The ISG positions do not constitute backfitting, inasmuch as the ISG is guidance directed to the NRC staff with respect to its regulatory responsibilities.


Question: What are nuclear fuel assembly characteristics?

Assembly characteristics represent cold, unirradiated, nominal configurations.

The NRC ensures that all nuclear facilities maintain a high standard of safety and security in their operations.


Question: concurrent cladding strain and oxidation in BWR experiments

A 1997 ORNL paper states that concurrent cladding strain and oxidation occurred in the BWR CORA experiments as a whole.

When NRC chooses to invalidate experimental data, which is important for simulating accidents, with unsubstantiated postulations, NRC undermines its own philosophy of defense-in-depth, which requires the application of conservative computer safety models.


Question: The aging management program for Class 1 Small-Bore Piping will align with the guidelines set in NUREG-2191.

The ASME Code Class 1 Small-Bore Piping aging management program will be consistent with the 10 elements of aging management program XI.M35, ASME Code Class 1 Small-Bore Piping, specified in NUREG-2191.

This new program will be implemented and one-time inspections will be performed within the six year period prior to the subsequent period of extended operation.


Question: What are the requirements for PWR undamaged SNF assemblies in a TSC?

Quantity per TSC: Up to a total of 37 PWR UNDAMAGED SNF ASSEMBLIES including up to four (4) DFCs containing PWR UNDAMAGED SNF ASSEMBLIES, PWR DAMAGED SNF ASSEMBLIES, and/or PWR FUEL DEBRIS loaded in the DFC locations No. 4, 8, 30 and 34, as shown on Figure B2-3, for the DF Basket Assembly.


Question: What are the cladding temperature values at the PCT and burst elevations?

The cladding temperature at PCT and burst elevations is specified with parameters C₀ = 0.6, HIGH Tₐvg, MIN SI, COSINE POWER SHAPE, NON-IFBA.

The data presented in the table shows the temperature values at various time intervals during the test.


Question: The safety measures in place for the reactors include a comprehensive assessment of potential failure scenarios.

The analysis of the defense in depth and diversity common principles is essential for evaluating the reliability of safety systems in nuclear reactor operations.

The cooling systems of the reactors are designed to maintain optimal temperature ranges during all operational phases.


Question: What is the required RCS temperature during a steam break event?

MONITOR RCS temperature stable at or trending to 557°F: - RCS Loop T-swg with any RCP running, - OR RCS Loop T-cold with RCPs out-of-service.

The event description included a large steam break inside containment that could lead to significant safety challenges and requires immediate operator actions.


Question: plume exposure pathway emergency planning zone sizing methodology

This report presents the methodology for sizing the plume exposure pathway emergency planning zone (EPZ) associated with the Natrium™ power plant technology, focusing on the potential radiological impacts and the required protective actions for the surrounding population.

The Natrium™ system includes advanced safety features designed to minimize the likelihood of accidents and to mitigate the consequences of an unplanned event.


Question: Following a design basis accident, the integrated strength of gamma and beta radiation is measured across different times.

Integrated gamma ray and beta source strengths at various times following a design basis accident show significant variations in energy release levels.

The measurement of temperature changes in the reactor is crucial for ensuring safe operational limits.


Question: license amendment request

Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (NSPM), submitted a license amendment request (LAR) for Renewed Facility Operating License No. DPR-22 for the Monticello Nuclear Generating Plant (Monticello).

The audit plan included audit questions for the discussions between NSPM and the NRC staff during the week of August 17, 2020.


Question: To assess seismic influences, it is essential to develop a detailed hazard curve specific to the site conditions.

Model the site-specific foundation properties for the 10^-6/yr seismic hazard


Question: The safety functions of the main steam supply system must remain intact even during significant seismic events.

All the power supplies and controls necessary for safety-related functions of the main steam supply system are Class 1E, as described in chapters 7 and 8.

The components of the main steam supply system are qualified to function in normal, test, and accident environmental conditions.


Question: Strontium-90 regulatory requirements

**Notes:**  Counting Uncertainty is calculated at the 95% confidence level (1.96-sigma).


Question: What are the electrical systems monitoring the Unit SAT Banks?

Two independent offsite sources to the Unit SAT Banks are available.

Normal power: SAT 142-1 and 142-2 energized or, U-1 SAT crosstie links installed and either SAT 142-1 or 142-2 energized.


Question: What is the significance of Americium-241 in nuclear safety assessments?

Americium-241 is a radioactive isotope that may be present in nuclear waste and needs to be quantified for safety assessments.

The NRC mandates that all regulatory filings must be signed by the authorized individual or representatives of the licensee.


Question: The examination of the condition of fire protection components is crucial for maintaining reactor safety.

The NRC requires a thorough management of aging effects for components like spray nozzles and sprinkler heads within fire protection systems.

Emergency procedures for reactor operators focus primarily on immediate response to potential malfunctions and external threats.


Question: This meeting will provide insights into the methodology that assesses radiological consequences.

The purpose of this preapplication meeting is to discuss with SMR, LLC's (A Holtec International Company) Introduction to Radiological Consequences Methodology Topical Report.

The NRC requires public feedback on proposed nuclear regulations to ensure safety and compliance.


Question: reactor vessel support

The reactor vessel is supported by six sliding foot assemblies mounted on the neutron shield tank as shown in Figure 5.5-8. These foot assemblies were fabricated from modified AISI 4330 steel forgings.

Operating-basis earthquake and design-basis earthquake have been concluded to yield generally similar loads that are a small percentage of the total design load.


Question: The NRC mandates that license holders take necessary actions to comply with safety regulations and requirements under the provisions stated in 10 CFR.

Accordingly, the NRC has concluded that these measures are necessary to ensure adequate protection of public health and safety under the provisions of the backfit rule, 10 CFR 50.109(a)(4)(ii), and is requiring Licensee or CP holder action.

These CP holders shall complete full implementation prior to issuance of an operating license.


Question: What is the significance of reflood rates in determining peak cladding temperature during a LOCA?

[...] it does not go far enough. As previously stated in PRM-50-93 (page 13), regarding the significance that coolant flood rates played in the PWR FLECHT test program, the “PWR FLECHT Final Report” states, “[i]n general, the effect on heat transfer coefficient[s] of varying system parameters was clearly discernible, with flooding rate being by far the most influential.”

FLECHT run 9573 was 1.24 kW/ft. The lower fuel rod power of the TH-1 tests would not have affected their overall PCT increases as much as the higher fuel rod power of FLECHT run 9573 affected its overall PCT increase.


Question: The Joint Information Center is activated to provide a centralized source of information during emergencies, especially at alert levels.

The JIC is usually activated at an Alert or higher ECL.

The telephone number is published in the yearly public postcard directing to the Entergy Public Information website and on Entergy press releases during an event.


Question: The NRC Form 313 is submitted as part of the application process for renewing a materials license.

Please find enclosed NRC Form 313 and supplementary materials in support of our application to renew our NRC Materials License No. 52-23059-01, which is scheduled to expire on March 31, 2023.

If you need any further information, please feel free to contact David Rhoe, UCC Radiation Safety Officer, at (787) 245-7248 (crmidmr@aol.com) or Dr. Richard Hann, UCC Radiation Safety Committee Chairman, at (787) 798-3001 x2031 (richard.hann@uccaribe.edu).


Question: What are the results from the Recirculation System Fly-wheels Test?

Results from the Recirculation System Fly-wheels Test – Recirculation pump 1 Speed

Technical drawings will be evaluated during the design review process.


Question: periodic verification of isolation devices

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.

The method of isolation must include the use of at least one isolation barrier that cannot be adversely ...


Question: Risk-Informed Categorization and Treatment of Structures, Systems, and Components

Response to Request for Additional Information Regarding LaSalle License Amendment Request to Renewed Facility Operating Licenses to Adopt 10 CFR 50.69, 'Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors,' dated October 1, 2020 (ADAMS Accession No. ML20275A292).

Staff Review of Spent Fuel Pool Evaluation Associated With Reevaluated Seismic Hazard Implementing Near-Term Task Force Recommendation 2.1 and Staff Closure of Activities Associated with Recommendation 2.1, dated January 19, 2017 (ADAMS Accession No. ML16351A231).


Question: nickel alloy core shroud

Nickel alloy core shroud and core plate access hole cover (welded covers) exposed to reactor coolant | Cracking due to SCC, IGSCC, irradiation-assisted SCC | AMP XI.M9, "BWR Vessel Internals," and AMP XI.M2, "Water Chemistry" | Yes | Consistent with NUREG-2191 with exceptions.

Not Applicable. The BFN BWR design does not include an isolation condenser.


Question: What are the expected characteristics of the discharged fuel assemblies?

The discharged assembly (highest power) is rated at 1.72 with a maximum temperature observed at 3900°F.

The temperature distribution indicates that at fuel volumes below 2500, the local fuel volume reaches 80% with a power generation of 19.18 MWt.


Question: What is a well driller's responsibility in nuclear site evaluations?

The driller is responsible for the proper surveying and documentation of boreholes, ensuring compliance with environmental and safety regulations during the drilling process.

Attempted to contact many of the drillers in this table. Some drillers were not surveyed because phone messages were not returned after more than one attempt.


Question: Public attendees will be able to observe the meeting without participating until the designated public input segment.

Public participants are observers in listen-only mode until the public participation session toward the end of the meeting.

Meeting slides can be found in the NRC’s ADAMS (ML25034A175) and on the public meeting schedule page for this meeting.


Question: The NRC has set forth specific guidelines on the necessary cool time periods for fuel assemblies with high burnup levels.

For fuel assembly average burnup greater than 45 GWd/MTU, cool time tables have been revised to account for a 5% margin in heat load.

Loading tables provide an overview of various reactor types based on their specifications and capabilities.


Question: silicic tephra-fall hazards

Table 50: Summary of conditional tephra-fall hazards at the proposed site, from silicic volcanoes in the site region.

Table 54: Combined mafic (Table 26) and silicic (Table 51) tephra-fall hazards from all potential volcanic sources in the proposed site region.


Question: What is the significance of the Keowee Dam in flood modeling?

The Keowee Dam (upstream) is the primary control on the water level upstream of the canal, which allows flow from the Keowee River Basin into the Little River Basin and eventually into the intake canal.

Model runs 100B and 100F were selected for the more detailed 2-D assessment of the downstream Keowee tailrace area.


Question: What is the impact of oscillations on SG/DHRS heat transfer performance?

Therefore, the oscillations are expected to have minimal impact on the overall SG/DHRS heat transfer performance. This conclusion of minimal impact is consistent with the results of the NIST-2 SG/DHRS testing. The testing results for Run 2 show that measured heat transfer is reasonably predicted by NRELAP5.

The examination of the structural integrity of the containment building is crucial for ensuring operational safety throughout the facility's lifespan.


Question: nuclear regulatory process

The Nuclear Regulatory Commission (NRC) is tasked with establishing regulations to govern the licensing of commercial nuclear reactors and ensuring compliance with safety standards.

The turbines are designed to convert steam energy into mechanical energy, which is then transformed into electrical energy through a generator.


Question: Xe-100 reactor technology

engagements relative to the generic design, analysis, and licensing basis of the Xe-100 reactor technology are contained in X-energy’s REP and are not repeated here.

Beginning in the third quarter of 2024, EN expects to enter into interactive meetings with NRC staff on licensing topics relevant to the EN SMR project.


Question: What is the significance of defense-in-depth in nuclear safety?

Defense-in-depth is a key safety principle applied in the design and operation of nuclear facilities, aimed at preventing accidents and mitigating their consequences through multiple layers of safety measures.

The Turkey Point Unit 3 and 4 facilities utilize advanced cooling systems to enhance efficiency and reduce operational costs.


Question: What is the role of phase A containment isolation in nuclear reactor safety?

Phase A containment isolation, whose function is to prevent fission product release to the site boundary by isolating all nonessential lines.

Containment pressure is sensed by four physically separated differential pressure transmitters mounted by strong supports outside of the Containment, which are connected to the Containment atmosphere by stainless steel impulse tubing.


Question: What activities were associated with the characterization survey methods during decommissioning?

Section 2.1.3, “Chronology of Decommissioning Planning and Characterization,” of the LTP summarizes the history of the licensee’s decommissioning planning activities.

The licensee provides a summary description of the planning activities associated with decommissioning and license terminations.


Question: The NRC mandates the evaluation of thermal hydraulic performance to ensure adequate cooling during a loss of coolant accident.

The COBRA-FLX thermal hydraulic code is used for determining the critical heat flux based on the input parameters and operational conditions.

Table B-1 outlines the various configurations tested during the non-mixing grid experiments, including the geometric and thermal properties of each assembly.


Question: What is the weighted mean activity in pCi/grams for nuclides?


Question: nuclear area measurement

The subtracted area is defined in the context of peak analysis and represents the difference between the original area and the ambient background, calculated for each peak measured in keV.

The ambient background is a crucial aspect in many measurements, but the specific count for each energy peak varies and must be documented accurately.


Question: Open-Cycle Cooling Water System aging management

The DCPP Open-Cycle Cooling Water System AMP provides reasonable assurance that the effects of aging will be managed so that the intended function(s) of components within the scope of the AMP will be maintained consistent with the CLB during the PEO.

In 2020, during coatings maintenance inside the ASW System Pump Vault, minor rust and corrosion on the pipe coating was identified.


Question: What is the concept of plume exposure pathway EPZ in nuclear regulations?

The decisions regarding EP should be based upon projected offsite dose from such accidents and the pre-determined plume exposure pathway EPZ for pre-planned protective measures.

The general considerations from the existing planning basis for EP, established in NUREG-0396/EPA 520/1-78-016, “Planning Basis for the Development of State and Local Government Radiological Emergency Response Plans in Support of Light Water Nuclear Power Plants,” introduced the concept of generic EPZs as the basis for preplanned response actions.


Question: What is selective leaching in nuclear materials?

The DCPP Selective Leaching AMP is a new condition monitoring AMP that manages the loss of material due to selective leaching for gray cast iron, ductile iron, and copper alloys (except for inhibited brass) containing greater than 15 percent zinc or greater than 8 percent aluminum components within the scope of license renewal that are exposed to raw water, treated water, closed cooling water, or ground water.

This new DCPP One-Time Inspection of ASME Code Class 1 Small-Bore AMP will be implemented prior to November 2, 2024 and August 26, 2025 for Units 1 and 2, respectively.


Question: zoning map changes for nuclear facilities

By amending the zoning map which is made a part of the ordinance by changing the zoning district of Roane County tax map 021, parcel 002.00, also known as Horizon Center Development Area/Lot 6A, containing approximately 110 acres, from IND-2, Industrial District, to IND-3, Industrial District.

Effective Date: September 22, 2022. (2 pages)


Question: What is the TRITON11 Dryout Correlation in the context of nuclear licensing?

TRITON11 Dryout Correlation (Submitted in Nov 2023, DRAFT SER in December 2024) is a key element of the licensing process for analyzing fuel behavior under various operating conditions.

We previously communicated plan to submit an amendment to the ADOPT LTR to include BWR applications.


Question: impact of nuclear plant construction on local economy

The major indirect tax impacts which will be experienced by the governmental units of Jackson County, Scottsboro, Hollywood, and their school districts are the additional property and sales taxes paid by new resident temporary construction workers and permanent plant employees.

We have concluded that the only direct tax impact which the governmental units of Jackson County, Scottsboro, Hollywood, and their school districts will experience as a result of the location of the plant will be caused by the removal of 1500 acres of land from the tax rolls and the attendant loss of property taxes paid by the previous private owners.


Question: high pressure fire protection systems

The High Pressure Fire Protection (Diesel Driven Pump) System is designed to function as a pressure boundary for fire suppression, utilizing metallic materials that must be managed for aging effects like general corrosion and pitting.

The environment of the facility includes various systems that operate under different temperature and pressure conditions, requiring tailored approaches for each type.


Question: What is the purpose of using a 500 ml Marinelli sample in gamma spectrum analysis?


Question: nuclear ITAAC notification process

The ITAAC notification process involves the submission of ITAAC notifications by the licensee to inform the NRC of the status of ITAAC completion and any corresponding reviews, ensuring regulatory oversight and compliance.

The NRC conducts various inspections to verify that construction activities comply with the approved design and regulatory requirements, focusing on safety and performance evaluations.


Question: What is Loss of Power (LOP) instrumentation?

**LCO 3.3.8.1:** The LOP instrumentation for each Function in Table 3.3.8.1-1 shall be OPERABLE.

The required maintenance of the plant's emergency coolant system is critical for safe operation during routine and non-routine conditions.


Question: MNGP reactor building construction phases

MNGP Reactor Building Facing West-Northwest near Completion of Construction, 1969

Potential Areas on the MNGP Site for Alternatives Solar Installation


Question: nuclear reactor operator licensing requirements

The requirements for licensing nuclear reactor operators are set forth in 10 CFR Part 55, which includes provisions for qualifications, examinations, and the necessary skills to ensure the safe operation of the facility.

The emergency cooling system was designed to function without the need for external power supply, relying entirely on gravity-fed water flow to maintain reactor core temperatures during a loss of coolant accident.


Question: The critical boron concentration is influenced by various factors including the temperature of the moderator and isotopes' behavior after shutdown.

The critical boron concentration in the analysis is primarily dependent on the moderator temperature coefficient, along with the temperature change of the moderator (i.e. cooling), and the buildup and decay of isotopes in the core after shutdown.

Table 1 shows biases and conservatisms applied to the boron dilution analysis with a summary of the impact on the analysis. The biases and conservatisms include prescriptive deterministic assumptions required for regulatory approval, such as the assumption that the highest worth control rod remains withdrawn from the core for at least 72 hours, as well as analysis-specific conservatisms such as assuming.


Question: radioactive materials license

These distributions were performed under Massachusetts Radioactive Materials License # 20-6752.

These distributions were performed under appropriate guidelines to ensure safety and compliance during operations.


Question: non-cited violation

One finding of very low safety significance (Green) is documented in this report. This finding involved a violation of NRC requirements. We are treating this violation as a non-cited violation (NCV) consistent with Section 2.3.2 of the Enforcement Policy.

On December 31, 2024, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Saint Lucie Plant, Units 1 and 2.


Question: What is the significance of near field settlement in nuclear head loss testing?

Wolf Creek is not crediting near field settlement of chemical precipitate in chemical head loss testing. See Figure 3.0.2-1.

Measures taken during the test, as described in the Response to 3.f.12, to keep debris suspended and transportable to the test strainer, prevented notable...


Question: minimum cooling time for PWR fuel assemblies

The data also includes assembly types from manufacturers such as CE and B&W.


Question: What is the purpose of an ITAAC in nuclear licensing?

An ITAAC, or Inspections, Tests, Analyses, and Acceptance Criteria, is used to ensure that safety-related systems and components have been constructed and will perform as intended before a nuclear reactor can begin operations.

The Federal Register Notice published for the ITAAC indicates the status of the regulatory process and the necessary documentation for compliance and oversight.


Question: The crew must decide whether to stay in 2-EOP-LOCA-1 or transition to a new procedure.

In accordance with 2-EOP-LOCA-1, which ONE of the following completes the statements below? The crew will ______.


Question: What is 10 CFR 50.59 in nuclear regulations?

Changes, Tests, & Exp. are governed by 10 CFR 50.59, which outlines the regulatory requirements for modifications to nuclear facilities.

Consider other connected regs. & guidance, as needed -- e.g., 50.55a, 21.3, 50.69, NSR guidance, etc.


Question: What are NELAP and ISO 17025 standards in nuclear testing?

Test results for NELAP or ISO 17025 accredited tests are verified to meet the requirements of those standards, with any exceptions noted.

Our policy is to provide high quality, personalized analytical services to enable you to meet your analytical needs on time every time.


Question: NAVD88 elevation measurement

Figure 2.3-15: Building Roof Elevations Included in Model (ft NAVD88)

Dam breaches can lead to significant environmental impacts and require thorough analysis to understand the extent of damage.


Question: What are CRGT guide cards and their inspection criteria?

The NRC staff finds that the methodology served as the technical basis for implementing a conservative, component-specific minimum coverage criterion change for the specified population of CRGT guide cards requiring inspection in WCAP-17451-P, Revision 1, from the prior 20 percent minimum population of CRGT guide cards requiring inspection in RMR-227-A.

The NRC staff confirmed that the changes made in WCAP-17451-P, Revision 2, were supported by the PWROG’s updated position in its letter No. OG-18-76 (PWROG 2018a) and provided a more conservative basis for selecting the minimum population of CRGT guide cards needed for inspection.


Question: What is the maximum amount of thorium that can be possessed under nuclear licensing?

Product stored in Kings Mt, NC on shelving, no other equipment necessary.


Question: What is the significance of managing surface water runoff in nuclear facility operations?

Surface water runoff at BFN is derived from precipitation remaining after losses due to infiltration and evapotranspiration. It can generally be classified as local surface runoff or streamflow. Surface water runoff from the plant site is to Wheeler Reservoir.

Field assessments of onsite surface water features were conducted in September 2021 to determine stream, drainage, pond, and wetland presence, extent, and condition within the BFN site (TVA 2021a). Wetland features are discussed in Section 3.6.1.


Question: containment sump system inspections

new SR will also require visual inspection of the containment sump system (including the containment drainage flow paths, any design features upstream of the containment sump that are credited in the containment debris analysis, the containment sump strainers, the pump suction trash racks, and the inlet to the ECCS and CSS piping for evidence of structural degradation, potential for debris bypass, and presence of corrosion or debris blockage) to ensure no loose debris is present and there is no evidence of structural distress or abnormal corrosion.

The NRC staff finds the proposed change is acceptable since it continues to provide remedial actions for when the containment sump is inoperable for reasons other than Condition A and ensures safe operation of the plant.


Question: What is the purpose of the Emergency Core Cooling Systems (ECCS)?

The Emergency Core Cooling Systems (ECCS) are designed to provide cooling water to the reactor core in the event of a loss of coolant accident (LOCA) to maintain the core temperature within safe limits and prevent core damage.

The containment structure is designed to prevent the release of radioactive materials into the environment under accident conditions.


Question: What is a general license registration in the context of nuclear regulation?

General License Registration Number: GL-700091-26

The address listed for the company is 177 Angell Street, Battle Creek, MI, 490378274.


Question: The management of solid waste generated at the facility is done according to applicable regulations.

Non-AEA-regulated solid wastes generated at the CUP will include construction debris, decontaminated materials and equipment, general office trash, boxes, miscellaneous wood packaging and products, steel, pipes, and solid hazardous waste. These materials will be stored in commercial-sized dumpsters and periodically disposed of by a commercial waste disposal operation, except hazardous waste, which will be stored in secure containers in designated hazardous waste storage areas and transported by an appropriately licensed transporter. All non-AEA-regulated solid waste will be managed in accordance with existing regulations and disposed of in a landfill that has been permitted under subtitle D of RCRA or a hazardous waste recycling or disposal facility permitted through NMED or a nearby site.

Pursuant to 10 CFR Part 40, Appendix A, Criterion 9, NuFuels has provided a plan for adequate financial assurance for the project NRC approved revised RAPs for each of the project areas in 2011. NuFuels will submit annual financial assurance updates to NRC for review and approval in accordance with LC 9.5 of SUA-1580.


Question: licensed operator requalification program

The inspectors reviewed and evaluated the licensed operator examination failure rates for the requalification annual operating exam completed on November 13, 2024.

The inspectors evaluated internal flooding mitigation protections in the Units 1 and 2 ECCS pump rooms on November 21, 2024.


Question: What is the purpose of a Quality Assurance Program in nuclear licensing?

A Quality Assurance Program is essential for ensuring that all activities relating to the design, construction, and operation of nuclear facilities meet safety standards and regulatory requirements.

The fuel handling system is critical for maintaining the reactor's operational efficiency and involves the management of fuel assemblies throughout their lifecycle.


Question: holding time for samples

Holding Time Samples (See Below) were received and analyzed outside the hold time.

Samples 493624002 (L3-012-101-RAAGS-004-SS) and 493624012 (L2-011-104-FSGS-010-SS) were recounted due to results more negative than the three sigma TPU.


Question: What are the requirements for the containment sump in nuclear safety regulations?

The new specification retains the existing requirements on the containment sump and the actions to be taken when the containment sump is inoperable with the exception of adding new actions to be taken when the containment sump is inoperable due to containment accident generated and transported debris exceeding the analyzed limits.

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or SR.


Question: What is the residual heat removal system (RHRS)?

The Residual Heat Removal System (RHRS) is a safety feature designed to remove residual heat from the reactor core under low-power or shutdown conditions to prevent overheating.

The NRC places significant emphasis on the importance of operator training and the assessment of knowledge and abilities (K/A) that are critical for ensuring safe reactor operations.


Question: The regulatory framework mandates thorough assessments of potential environmental impacts during nuclear licensing procedures.

Nuclear licensing involves the detailed evaluation of an applicant's ability to operate a facility safely and in compliance with regulatory requirements.

The geology of the Columbia North Quadrangle was extensively studied in 1949, focusing on the local stratigraphic relationships and formations.


Question: What is Excess Letdown in the context of reactor coolant systems?

Excess Letdown is a configuration of the reactor coolant system that allows for controlled removal of coolant to accommodate operational needs or issues such as maintenance and leakage.

Unit 1 is at 100% Power and the RP Shift Supervisor has been informed of the intent to establish Excess Letdown.


Question: What regulations must be followed for nuclear licensing submissions?

Nuclear licensing involves a comprehensive review process that encompasses safety assessments, environmental evaluations, and the examination of the proposed facility's compliance with regulatory requirements.

KH Imports, LLC is a vendor that provides various items for nuclear facilities; the catalog includes multiple equipment codes and descriptions for specific parts.


Question: The NRC dictates that safety analysis must accompany all license applications under Title 10 of the CFR.

The KP-FHR relies on a functional containment approach similar to the Modular High Temperature Gas-Cooled Reactor (MHTGR) instead of the typical light water reactor (LWR) low-leakage, pressure retaining containment structure.

The reactor coolant is a chemically stable molten fluoride salt mixture, 2LiF·BeF2 (Flibe with [[ ]]), which also provides retention of fission products that escape from fuel defects.


Question: What are Chi/Q calculations in nuclear safety assessments?

Short-term Chi/Q analyses were performed using the PAVAN model. The results of the PAVAN modeling analysis are summarized in Table 2.7-51 and Table 2.7-52.

In addition, Case 2 was evaluated with and without building wake effects.


Question: The structural support components of the heat exchanger are analyzed for material loss in various environments.

Heat Exchanger (Aux Steam Drain Rec Vent Cond) Tube Side Components | Structural Support                           | Carbon Steel      | Closed-Cycle Cooling Water (Int) | Loss of Material                  | Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components (B.2.3.24)

Heat Exchanger (Auxiliary Steam Drain Receiver Pump Discharge Sample Cooler) Shell Side Components | Leakage Boundary (spatial)                     | Copper Alloy (≤15 inc and ≤8 Aluminum) | Plant Indoor Air (Ext)         | None                              | None


Question: What is the significance of SCC in aging management for metallic storage tanks?

Stainless steel tanks (within the scope of AMP XI.M29, "Outdoor and Large Atmospheric Metallic Storage Tanks") exposed to air, condensation are subject to cracking due to SCC.

Not Applicable. There are no aluminum tanks (within the scope of AMP XI.M29, "Outdoor and Large Atmospheric Metallic Storage Tanks") exposed to air, condensation, soil, concrete, raw water, waste water within the screened-in portions of the Steam and Power Conversion Systems. See Subsection 3.4.2.2.7.


Question: NRC exemptions process

Exemptions approved by the NRC are detailed in historical information and these references: 4.1.3.2, 4.1.3.3, 4.1.3.4.

Historical data plays a crucial role in the development of nuclear safety protocols and regulatory measures that ensure the protection of public health.


Question: The 14-day Completion Time is designed for planned maintenance or inspection activities, allowing operations to continue without interruption.

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.

In the event the inoperable DG is restored to OPERABLE status prior to completing either B.3.1 or B.3.2, the plant corrective action program will continue to evaluate the common cause possibility.


Question: 10 CFR 50.69 categorization process

A recent Brunswick LAR dated August 17, 2023 (ML23229A456) also asserts how the applicant will follow the same alternative seismic approach in the 10 CFR 50.69 categorization process as that approved by the NRC staff for LaSalle with the following two exceptions.

Please describe if the proposed amendment incorporates site-specific LaSalle information in the approved LaSalle 10 CFR 50.69 LAR and RAI responses.


Question: What is the activity of EU-154 in pCi/grams?

EU-154 has an activity of 6.90E-03 pCi/grams.

The nuclide MDA represents the minimum detectable activity, which is the lowest activity level that can be reliably detected in measurements.


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

The activity (pCi/grams) of a nuclide indicates the rate at which it undergoes radioactive decay, which is crucial for assessing potential radiation exposure and safety protocols in nuclear facilities.

Emergency procedures must include steps for evacuation and communication in the event of a radiation incident, ensuring that all personnel are accounted for and that the public is informed.


Question: loss of offsite power

During the loss of offsite power, the DG powers the motor driven auxiliary feedwater pumps which are automatically started after expiration of the appropriate time delay by the LSELS.

The closure of tie breaker 52NG0116 between 480 Vac buses NG01 and NG03 or tie breaker 52NG0216 between 480 Vac buses NG02 and NG04 to repair failed or degraded equipment.


Question: What is the reason for blocking the Steamline Isolation during a reactor cooldown?

Per step 8 of 1BEP ES-1.2, the crew will block Steamline Isolation SI once PZR pressure is less than 1930 PSIG, and this will allow the continuation of dumping steam from intact S/Gs to the condenser.

The written examination for an operator for a facility will include a representative sample from among the following 14 items, to the extent applicable to the facility.


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

The identified nuclides report lists various isotopes including their energy levels, yields, and activities, which are crucial for assessing the radiological impact and ensuring compliance with regulatory standards.

Energy tolerance is defined as 1.000 keV, ensuring statistical validity in measuring spectra during analysis.


Question: The reactor protection system's trip commands are relayed to activate safety mechanisms like the scram solenoid valves.

As shown in Figure 3.5.6-1, the HARP portion of the RPS TU in each division includes the MCR manual scram pushbutton input signal that will actuate the solid-state relay output relays.

The solid-state relays in the Division 2 and 4 RPS TUs are hardwired together to create a functional OR, so that either pair of relays de-energizing will generate a scram output from the PPS to the scram pilot solenoid valves (along with the associated aforementioned values).


Question: The NRC conducts evaluations of the impacts on diverse species that may inhabit or move through the action area, especially during the license renewal process.

This section evaluates nine species, eight of which are listed and one of which is proposed for listing under the ESA, that may be present in the action area.

In 1999, the NRC staff performed an environmental review for the Oconee Station's licensing, which involved numerous administrative processes and technical assessments regarding its operation.


Question: sensitized stainless steel in nuclear components

Sensitized stainless steel is defined as those types of unstabilized austenitic stainless steels (typically types 304 and 316) which have been exposed to elevated temperatures (800°F to 1500°F) for extended periods of time.

Control of welding processes and procedures to avoid HAZ sensitization.


Question: normal range for instrument air pressure

The normal range for instrument air is 110 - 115 PSIG.

See above. Plausible misconception that the normal range of IA is from 110 - 130 PSIG. The second part is correct.


Question: What is the significance of Radium 226 in groundwater sampling?


Question: What are the actions required when thermal power changes exceed 15%?

WHEN rated thermal power change exceeds 15% in one hour, THEN NOTIFY Chemistry to initiate 1-SI-68-28.

As Chemistry, acknowledge notification of power change.


Question: What is the project number for the Duke Energy Crystal River plant?


Question: What is the significance of temperature measurements in nuclear reactor operations?

Temperature measurements are critical in nuclear reactor operations as they ensure that the reactor operates within specified thermal limits to prevent overheating and maintain safety protocols.

In contrast, the financial data presented in the report indicates an increase in expenditures related to reactor maintenance and operational procedures.


Question: What are QPTR values in nuclear regulatory processes?

Applicant calculates QPTR values by dividing the respective Relative Powers [Step 5] by the Average Relative Powers for each NIS channel detector [Step 6]. The results recorded in appropriate blocks on Attachment 4.

The examiner notes that the acceptable range is ± 0.0002 to account for rounding errors.


Question: The functions of the Station Emergency Organization are critical in managing emergency situations effectively.

The Station Emergency Organization is responsible for coordinating the onsite emergency response efforts during an emergency event.

Effective communication is also crucial in managing relationships with media representatives during an emergency.


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


Question: What factors must be evaluated when assessing plant changes in relation to LOCA frequencies?

Option one is to directly evaluate the effects of the change on those variables that are most important that affect both the frequencies that are articulated in 1829. And these are things like, you know, did the change affect the plant materials, the environment, the loading, degradation rates, the geometry, and then either the maintenance or the mitigation practices associated with that particular system?

So, in terms of plant changes and LOCA frequencies, again they need to demonstrate that they don’t -- There’s, sorry, a couple different options here.


Question: radiological protection staffing

The proposed change combines the Protective Action and Radiological Assessment functions and utilizes the two (2) on-shift RP Technicians for performance of these tasks as noted in NUREG-0654, Revision 2.

Per NUREG-0654, Revision 1, the Firefighting functional area is addressed by use of a Fire Brigade and managed in accordance with site T.S.


Question: What is the purpose of a probabilistic risk assessment in nuclear safety?

Probabilistic risk assessment (PRA) is a comprehensive and systematic evaluation of the risks associated with a nuclear facility, incorporating the likelihood of different types of events and the potential consequences of those events.

The coolant system in a reactor is designed to remove heat from the core and transfer it to the steam generator, ensuring that the reactor operates within safe temperature limits.


Question: TVA has announced plans to pursue new advancements in nuclear technology and enhance safety measures related to licensing processes.

TVA Board Authorizes New Nuclear Program to Explore Innovative Technology. February 10, 2022. Accessed October 5, 2022 at https://www.tva.com/newsroom/press-releases/tva-board-authorizes-new-nuclear-program-to-explore-innovative-technology.

The TVA's strategy includes a wide range of renewable energy projects, such as wind and solar initiatives, which are separate from nuclear energy advancements.


Question: The relationship between the feedwater NPSH and the system's capability to respond effectively to low NPSH conditions is crucial for operation.

The examinee must understand the relationship between FW NPSH and Condensate system response to a FW NPSH signal.

The Gland Condenser Bypass Valves will get a close signal on the Main Feed pump LOW NPSH alarm in order to bypass the Gland Steam condenser and ensure that maximum flow is getting to the condensate and feed system.


Question: What is the aging management process for concrete structures in nuclear power plants?

Loss of material (spalling, scaling) and cracking due to freeze-thaw is a significant consideration in the aging management of concrete used for structural support and water retaining boundaries in nuclear power plants.

The NRC requires thorough inspections of various reactor components to ensure compliance with regulatory safety standards.


Question: What is the primary containment isolation instrumentation in nuclear reactors?

The Primary Containment Isolation Instrumentation is outlined in section 3.3.6.1 of the Fermi 2 Technical Specifications, specifying the requirements for operable channels and their function within nuclear safety protocols.

Fermi 2 TS numbering for the RCIC Function 3.3.5.2 is different than NUREG-1433 numbering for the RCIC Function, which is 3.3.5.3.


Question: What is the projected population growth for counties near DNPS?

By 2051, the end of the proposed DNPS operating term, the permanent population (not including transient populations) of the 21 counties is projected to be approximately 9,887,287. Based on 2020–2051 population projections, an annual growth rate of approximately 0.17 percent is anticipated for the permanent population in these 21 counties.

The area within a 50-mile radius of the DNPS site totally or partially includes 19 counties in Illinois and two counties in Indiana and are detailed in Table 3.11-2.


Question: What is the seismic velocity for hard rock?

The seismic velocity for hard, relatively unweathered rock is approximately 13,000 to 16,000 ft./sec.

Depths to rock are shown on the profile sections which accompany this report.


Question: piezometric map

(Piezometric Map by Davis & Floyd, 1981, Fig. 1)


Question: What is the significance of digital modernization in nuclear reactor safety?

Digital modernization enhances the safety and efficiency of nuclear reactors by integrating advanced technology to monitor and control systems, providing improved data analytics and response capabilities during operations.

Florida Power & Light Co. has implemented several enhancements at Turkey Point Unit 3&4 to improve operational performance and reliability.


Question: The assessment of changes in core damage frequency (CDF) is crucial for evaluating safety improvements in nuclear power facilities.

this will result in identifying and assessing the range of reduction in CDF, as well as estimating the representative change for the class.

Analysts should use Table 2-2 or more recent data, as appropriate, to perform a preliminary screening of the merit of the proposed nuclear power requirements for the appropriate class of nuclear power plants.


Question: nuclear material quantity limits

Greater than 10^4 but less than or equal to 10^5 times the applicable quantities of Appendix B to Part 30 in unsealed form.

The NRC provides comprehensive guidelines for the licensing of nuclear reactors, ensuring that safety and operational standards are met.


Question: loss of material due to corrosion in nuclear reactors

Loss of material due to general, pitting, and crevice corrosion could occur in steel elements of inaccessible areas for all types of PWR and BWR containments. The existing program relies on ASME Code Section XI, Subsection IWE, and 10 CFR Part 50, Appendix J and JAMPs, to manage this aging effect. Further evaluation is recommended of plant-specific programs to manage this aging effect if corrosion is indicated from the IWE examinations.

The DNPS Primary Containment design includes four moisture barriers. Three of these act to stop moisture and other foreign material from entering the 2-inch expansion gap between the drywell shell and the containment shell wall in the Reactor Building.


Question: materials for Sampling and Water Quality System

The materials of construction for the Sampling and Water Quality System components are: - Cast Austenitic Stainless Steel - Copper Alloy - Copper Alloy (>15% Zn or >8% Al) - Glass - Gray Cast Iron, Ductile Iron - High-Strength Steel - Nickel Alloy - Polymeric - Stainless Steel - Steel

The following aging management programs manage the aging effects for the Station Drainage System components: - Bolting Integrity (B.2.1.10) - One-Time Inspection (B.2.1.20) - Selective Leaching (B.2.1.21) - External Surfaces Monitoring of Mechanical Components (B.2.1.23) - Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components (B.2.1.24)


Question: What are the design basis requirements for the RHRSW system?

Perform 24.205.06 Division 2 RHRSW Pump and Valve.


Question: core barrel assembly components

VV original locking device type – spring retainer


Question: The NRC standards specify limits on the concentrations of radionuclides in drinking water.

Sample ID: DUP-03-10032024 | Date Sampled: 10/03/24 0:00 | Date Received: 10/05/24 | Sample Matrix: Groundwater


Question: What is the process for performing a crosstie operation between chilled water systems?

At 1355, a briefing was held in the MCR to perform the crosstie of the VS chilled water system to the 0A VC chilled water system per procedure BOP VC-20, VC Chilled Water System to VS Chilled Water System Crosstie Operation with One Train Inoperable for Maintenance, Revision 4.

The station determined the root cause to be inadequate briefing with contributing causes including lacking system knowledge and experience and less than adequate procedural guidance.


Question: What is the purpose of a safety analysis report in nuclear licensing?

A Safety Analysis Report (SAR) is a document that provides a comprehensive analysis of the safety implications of a proposed nuclear facility, addressing the structures, systems, and components important to safety.

The fuel cycle involves the processes of mining, milling, enrichment, and fabrication of uranium into fuel for reactors.


Question: B values for low-level waste (LLW) disposal

The values shown in this table for the years 2024 and 2022 are developed in Appendix B, with all values normalized to the burial costs calculated for the year 1986.

Effective with NUREG-1307, Revision 16, the Compact Waste Facility (CWF) in Andrews County, Texas, is available as a full-service facility for waste generators located in States affiliated with the Texas Compact.


Question: What are the inspection procedures for nuclear facility operations?

The NRC has established Inspection Procedures to ensure that nuclear facilities operate safely and comply with regulations.

The document also includes information about recent staff changes and training schedules.


Question: The NRC mandates that acceptance tests are applied for the nonsafety-related portions of the project.

This test abstract will be performed on Unit 2 via preoperational test(s) on the safety-related portions and acceptance tests(s) on the nonsafety related portions.

The main turbine system will undergo a comprehensive preoperational test to ensure its operational readiness before entering full service.


Question: radioactivity measurement and reporting

The radionuclide K-40 has an activity of 1.318+/-0.01 with an uncertainty of 1.348+/-0.01, measured in g dry on 08/07/20 at 12:41, and reported on 09/22/20 with a count date of 4867.

The uncertainty values indicate the precision of the measurement and are crucial for ensuring regulatory compliance.


Question: The reactor operator accidentally adjusted the control rod to an incorrect position during a routine operation.

The error is immediately reported before selecting the next control rod.

All adjacent control rods are at position 48.


Question: The NRC stipulates that license holders must adhere to the technical specifications outlined in their operating licenses.

The amendment provides for changes to the technical specifications of the facility in accordance with regulatory requirements to ensure compliance with safety standards.

The regular maintenance of the reactor system is essential to prolong its operational life and ensure efficient performance.


Question: MCNP model infinite array

Figure 15 Radial Cross Section View of MCNP Model of an Infinite 8x8 Array Eccentric Positioning, Fuel Assemblies Toward the Center, Reflective Boundary Conditions

The regulatory compliance assessment was conducted to ensure that all safety measures were effectively documented and communicated to the stakeholders.


Question: What is COBRA-FLX and how does it relate to nuclear safety?

COBRA-FLX is a thermal-hydraulic code used to predict critical heat flux in nuclear reactor systems and is essential for ensuring the safety and efficiency of reactor operation.

The report details various test assembly conditions which include factors such as water temperature and pressure but do not directly relate to the COBRA-FLX functionality.


Question: nuclear fuel assembly burnup limits

The enrichment levels determine the amount of fissile material in the fuel.


Question: radiopharmaceutical extravasation management

In summary, there is no clinical data that supports Lucerno Dynamics' claim that extravasation of diagnostic radiopharmaceuticals is a patient safety issue.

In the absence of such data, I believe that extravasations are best managed on an institutional level at the discretion of the authorized user and do not require additional NRC regulation.


Question: content spray system

ENSURE Cntnt spray actuated.
 3. ENSURE Cntnt spray pumps running.
 4. ENSURE Cntnt spray flow.

WHEN 10 minutes has elapsed since Phase B actuated, THEN <br> 10. DISPATCH AUO to perform Attachment 1 (1-E-0), Ice Condenser AHU Breaker Operation.


Question: The document contains replies to inquiries made by the NRC regarding safety analysis and licensing requirements.

Response to NRC Request for Additional Information

The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Mr. Mark D. Sartain, who is Vice President – Nuclear Engineering and Fleet Support, of Dominion Energy South Carolina, Inc.


Question: Philadelphia terrane characteristics

The Philadelphia terrane (Figure 2.5.1-8a and 2.5.1-8b) consists mostly of the Wissahickon Formation along with granodiorite and granitic intrusions of the Wilmington Complex. The characteristics of the Wissahickon Formation are described in Plank et al. (Reference 2.5.1-168). The primary lithologies comprising the Wissahickon Formation are aluminous to quartz-rich gneisses interlayered with amphibolites.

One of the component units consists of low-grade meta-mafic to metalumafric rocks with protoliths that range in composition from pelitic to tuffaceous basalt; these rocks have been interpreted to be amphibolic components and to represent oceanic crust that was substrate to a volcanic arc (Reference 2.5.1-86).


Question: The NRC mandates that any analytical results must meet specified reporting limits to ensure compliance with safety standards.

Counting Uncertainty is calculated at the 95% confidence level (1.96-sigma).


Question: Chemical and Volume Control System

The Chemical and Volume Control System is capable of reducing the concentration of ionic isotopes in the purification stream as required in the design basis.

If the normal charging line is not available, charging to the Reactor Coolant System can be continued via reactor coolant pump seal injection at the rate of approximately 5 gm per pump.


Question: The NRC requires a licensee to submit sufficient information related to ITAAC notifications.

ITACCs are verification items identified in the licensing process that are necessary to ensure compliance with the regulatory framework for plant construction and operation.

In 2020, the NRC emphasized the importance of community engagement in nuclear power plant assessments.


Question: What is the significance of radioisotopes in radiation safety surveys?

Radioisotopes approved for use must be reported for safety evaluations in labs handling such materials.

The RSO quarterly inspection includes checks for compliance with a number of safety measures and inventory records.


Question: Licensees must obtain permission for specific activities related to the nuclear facility to ensure compliance with regulations.

The U.S. Nuclear Regulatory Commission (NRC) is responsible for licensing and regulating commercial nuclear power plants to ensure they operate safely within the safety standards established by federal regulations.

The maintenance of safety equipment is crucial to the overall operations of a power generation facility, ensuring that all systems are functioning correctly and efficiently.


Question: What is the purpose of a Safety Analysis Report in nuclear licensing?

A Safety Analysis Report (SAR) is a comprehensive document detailing the safety assessment of a nuclear facility, including potential hazards and the measures taken to mitigate risks.

The cooling system in a nuclear reactor is designed to remove excess heat from the reactor core and prevent overheating during normal operation.


Question: What are the failure modes of fission chambers?

Failure modes of fission chambers, ion chambers, and proportional counters    | 3.1        |

Theory and operation of voltmeters, ammeters, frequency detectors, and ground detectors | 2.4 |


Question: What are the implications of interpretive rule changes in nuclear regulation?

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.

Scrap this 'interpretive' rule change. Keep nuclear waste under nuclear regulatory control and make the polluters pay to try to isolate it--not release it!


Question: What is ex-plant materials harvesting in nuclear regulatory research?

The NRC’s Office of Nuclear Regulatory Research (RES) is sponsoring a March 7-8, 2017, workshop on Ex-Plant Materials Harvesting, to be held at NRC headquarters.

Data from ex-plant materials could help address technical gaps identified for extended operation of nuclear power plants due to highly relevant aging conditions.


Question: reasonable accommodation for participation in NRC meetings

The NRC provides reasonable accommodation to individuals with disabilities where appropriate.

Determinations on requests for reasonable accommodation will be made on a case-by-case basis.


Question: What are the entry conditions for Loss of Containment Integrity?

Which event is an entry condition into 1BOA PRI-11, Loss of Containment Integrity?


Question: design basis site-specific parameters

The design basis site-specific parameters and analyses that require verification by the MAGNASTOR® SYSTEM user are: a temperature of 76°F is the maximum average yearly temperature.

The potential for fire and explosion shall be addressed, based on site-specific considerations.


Question: plant licensing basis

The objective of the licensing basis section is to provide information regarding any changes to the plant licensing basis due to the sump evaluation or plant modifications.

The screen for the system is designed to capture large debris while allowing smaller particles to pass into the reactor core.


Question: What is the significance of Plutonium-241 in nuclear safety assessments?

Samples (See Below) were received and analyzed outside the hold time.


Question: Licensees must submit detailed safety analyses and reports to demonstrate compliance with the regulatory requirements.

The NRC requires licensees to submit applications for the issuance of operating licenses, including safety analysis reports.

Many components of the reactor are designed to enhance the overall thermal efficiency and operational performance of the nuclear facility.


Question: radiation protection requirements in nuclear licensing

Radiation protection measures are essential to ensure the safety of workers and the public from radioactive exposures during routine operations and postulated accidents.

Aquatic ecology assessments evaluate the impact of a proposed project on local water bodies and their ecosystems, focusing on fish populations and plant life.


Question: What aging effects are associated with Yard Structures in nuclear facilities?

The following aging effects associated with the Yard Structures components require management: Cracking, Distortion, Increase in Porosity and Permeability, Loss of Bond, Loss of Material, Loss of Material (Spalling, Scaling), Loss of Preload, Loss of Strength.

The materials of construction for the Yard Structures components are: Carbon and Low Alloy Steel Bolting, Carbon Steel, Galvanized Bolting, Galvanized Steel, Reinforced concrete.


Question: What is the significance of performing a circulating water temperature check in nuclear operations?

The operator must perform a circulating water temperature check to ensure that the systems are functioning within the required temperature ranges, which is critical for safe reactor operations.

Appropriate training and evaluation are crucial for ensuring that staff can effectively manage emergency scenarios and maintain compliance with regulations.


Question: fluid quality in nuclear reactors


Question: proposed license amendment request to revise emergency plan

Virginia Electric and Power Company (Dominion Energy Virginia) is requesting Nuclear Regulatory Commission (NRC) approval of proposed changes to the Renewed Facility Operating License Numbers NPF-4 and NPF-7 for North Anna Power Station (NAPS), Units 1 and 2, respectively, and Renewed Facility Operating License Numbers DPR-32 and DPR-37 for Surry Power Station (SPS), Units 1 and 2, respectively.

The extension of facility activation time requirements for the Technical Support Center (TSC), Operational Support Center (OSC), and Corporate Emergency Response Center (CERC) to 90 minutes is vital for ensuring effective emergency response.


Question: average enrichment level of uranium

The table provides data on the Initial Assembly Average Enrichment wt % 235U (E) which is crucial in determining fuel performance and safety margins in BWR reactors.

The document also describes various operational protocols that must be adhered to during safety inspections, emphasizing the need for thorough documentation and adherence to guidelines.


Question: NuScale Standard Design Approval

NuScale Power, LLC is seeking approval for its Standard Design Approval Application, which outlines the technical specifications and safety analysis necessary for the regulatory review process.

The agenda included a discussion on public comments and how they will be considered as part of the decision-making process.


Question: manual safety assessment

Perform a Manual Safety Assessment (Shutdown)

Determine Post-Maintenance Testing Requirements


Question: The NRC has outlined specific timelines for responding to supplementary information requests related to licensing applications.

As indicated in Reference 2, the schedule for responding to the requests for additional information (RAI) and request for confirmation of information (RCI) is within 30 days from the date of the email with the exception of RAI 4.3.2-3 which is due within 45 days from the date of the email.

In Reference 1, Constellation Energy Generation, LLC (CEG) submitted a License Renewal Application (LRA) for Clinton Power Station, Unit 1 (CPS).


Question: The NRC requires specific assessments for determining the impacts on water bodies for new project developments.

The delineation included herein has been conducted to identify the location and extent of the aquatic resource boundaries and/or the jurisdictional status of aquatic resources for purposes of the Clean Water Act for the particular site identified in this request.

Enclosed with this letter is a copy of the PJD. If you agree with the findings of this PJD and understand your options regarding the same, please sign and date the form and return it to this office within 30 days of receipt of this letter.


Question: What is the purpose of the Containment Pressure Control System in nuclear reactors?

These pressure sensors are part of the Containment Pressure Control System which is designed to maintain the containment pressure within design limits.

The spray nozzles can also be periodically tested using a vacuum blower to induce air flow through each nozzle to verify unobstructed flow.


Question: The NRC establishes certification requirements for the contents and configuration of spent fuel storage systems.

The CY-MPC 24-Assembly Basket Fuel Loading Positions are crucial for determining the proper arrangement of spent fuel assemblies during storage.

The NRC's guidelines emphasize the importance of clear documentation and consistent communication among regulatory bodies.


Question: The NRC mandates that licensees submit reports detailing significant events promptly to ensure proper oversight and response actions.

Paragraph 70.74(a) requires each licensee subject to Subpart H of Part 70 to report all events described in Part 70, Appendix A, to the NRC Operations Center, including supplemental information as it becomes available.

Section 70.73 lists the regulations that apply to license renewal applications.


Question: background radiation regulations

Background radiation is defined in 10 CFR § 20.1003 as, “Radiation from cosmic sources; naturally occurring radioactive material, including radon (except as a decay product of source or special nuclear material); and global fallout as it exists in the environment from the testing of nuclear explosive devices or from past nuclear accidents, such as Chernobyl, that contribute to background radiation and are not under the control of the license.”

For the average U.S. resident, the total effective dose from background radiation sources is approximately 3.1 mSv/yr [310 mrem/yr] but varies by location and elevation.


Question: DNBR limit in nuclear safety analysis

For deterministic DNB analyses, the design DNBR limit is set equal to the applicable code/correlation limit and it is termed the DDL.

These design limits are two of the Design Basis Limits for Fission Product Barriers (DBLFPB) described in Reference 13.


Question: The frequency of Plant Review Board (PRB) meetings will be adjusted to not occur on a set monthly schedule.

The quorum requirements of the Plant Review Board (PRB) are removed from the QATR in the proposed change.

Example changes that represent a RIC at Farley, Hatch, and/or Vogtle 1&2:


Question: k-eff calculation

The calculated values are used to determine the adequacy of the safety analysis supporting the licensing process.


Question: The assessment indicates that licensees have the flexibility to maintain their existing treatment of LOCAs due to the voluntary nature of the new regulations.

The assessment states that licensees would not be required to comply with the proposed amendments and would have the option to continue their current treatment of LOCAs.

They need to understand what the expectations are when they submit.


Question: minimum initial assembly average enrichment of U-235

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

The table presents various assembly average burnup values categorized by specific ranges of enrichment.


Question: wind speed data for nuclear sites

Table 2.3-36 shows the percentage frequency of wind direction and speed (MPH) during May at the GGNS site, providing important meteorological data for safety analysis.

The quality assurance program is essential for ensuring compliance with regulatory requirements and enhancing operational safety.


Question: What is the reporting limit in nuclear analytical chemistry?

The reporting limit (RL) is the lowest concentration of a substance that can be reliably detected and reported in analytical testing.

Client Sample ID: L1-SUB-TDS-FSGS-B06-SB


Question: The request involves updating technical specifications to include risk-informed completion times for reactor safety measures.

Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, 'Provide Risk-Informed Extended Completion Times – RITSTF Initiative 4b'

The Grand Gulf Nuclear Station is known for its advanced boiling water reactor technology, which plays a crucial role in achieving safe and efficient energy production.


Question: heat transfer in containment structure

For a conductor representing the containment floor or sump walls that will eventually be covered with water from the break and condensate, the Split heat transfer option is used to switch the heat transfer from the vapor phase to the liquid phase as the liquid level in the containment builds. A quicker transition to liquid heat transfer is more conservative for containment analysis.

Containment walls above grade and the containment dome have a specified external temperature boundary condition with a heat transfer coefficient of 2.0 Btu/hr·ft²·°F to model convective heat transfer to the outside atmosphere.


Question: What are the actions required in response to a Steam Generator Tube Leak?

Selecting rate of load reduction is a response relevant to a Steam Generator Tube Leak.

The proposed references to be provided include documentation on malfunctions of the reactor control systems.


Question: radioactive shipment survey

If UZ29015, is the Radioactive Shipment Survey Complaint? Yes/No.

The client designated samples as hazardous, including PCBs, Flammable materials, Foreign Soil, RCRA, Asbestos, and Beryllium.


Question: class 1 piping aging management

Cracking due to SCC (for stainless steel or nickel alloy surfaces exposed to reactor coolant only), IGSSC (for stainless steel or nickel alloy surfaces exposed to reactor coolant only), thermal, mechanical, vibratory loading.


Question: What are Completion Times in nuclear regulatory context?

Fermi 2 Specification 3.8.7, Distribution Systems - Operating, has an 8 hour Completion Time for one or more AC electrical power distribution subsystems inoperable (Condition A). One or more required DC electrical power distribution subsystems (Condition B) inoperable.

In addition to these regulatory programs, Section 1.3 of the Technical Specifications is revised to require administrative controls to limit the maximum time allowed for any combination of...


Question: nuclear licensing process

The nuclear licensing process involves a series of detailed regulatory reviews conducted by the NRC to ensure that the proposed facility meets safety, security, and environmental standards.

The safety analysis report must be updated regularly to reflect operational changes and ongoing assessment of the plant's performance.


Question: What safety equipment is required for personnel involved in mitigation activities?

Personnel involved in mitigation activities will need to be equipped with fitted respirators (i.e., N95 or greater) and closed-sided goggles, in addition to other routine personal protective equipment.

The hazards assessments in this report were conducted within a probabilistic framework that evaluates a broad range of potential conditions or scenarios.


Question: The Surveillance Requirements for Control Rod Block Functions detail the actions to be taken when certain conditions are met.

When an RBM channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains control rod block capability.

During the reactor's startup phase, various checks are conducted to ensure all systems are operational and safe for power generation.


Question: The evaluation of digital systems in nuclear facilities must adhere to rigorous safety standards.

The Defense-in-Depth strategy is vital for ensuring that safety systems are robust and capable of addressing potential failures in a nuclear reactor.

Regular maintenance and enhancements of digital systems are essential for operational efficiency at the Turkey Point facility.


Question: IEEE Standard 279-1971

Cables and conduits routed in non-Category I structures associated with safety related functions or anticipatory trips (i.e., turbine trip on reactor trip, reactor trip on turbine trip, loss of feedwater) are designed to meet IEEE Standard 279-1971 including redundancy, separation, and single failure criteria (see detailed description in Section 7.2.1.1.2).

The initial turbine controls trip system was a series of contacts that directly fed the Auto Stop Trip (AST) solenoids that would depressurize the Digital Electro-Hydraulic (DEH) header and trip the turbine.


Question: probabilistic seismic hazard analysis

A probabilistic seismic hazard analysis (PSHA) was performed for the GGNS following the procedure provided in RG 1.165 (Reference 4). The results were deaggregated in terms of earthquake magnitude (Mw) and distance to determine the controlling earthquakes for the site.

The EPRI 2003 ground motion model provides median ground motion models for the Mid-continent and the Gulf region of the central and eastern U.S. (Reference 13). In addition, the model is defined for different seismic source types, including General Area sources, and Fault.


Question: The characteristics of nuclear fuel assemblies typically include parameters such as the number of fuel rods and the dimensions of fuel pellets.

The assembly characteristics represent cold, unirradiated, nominal configurations.


Question: categorical exclusion criteria

Accordingly, the proposed amendment meets the eligibility criterion for a categorical exclusion set forth in 10 CFR 51.22(c)(9).

The proposed amendment does not involve: 1. a significant hazards consideration, or 2. authorize a significant change in the types or a significant increase in the amounts of any effluent that may be released Off-site, or 3. result in a significant increase in individual or cumulative occupational radiation exposure.


Question: Management of structural integrity is crucial for supports like cable trays and HVAC ducts due to aging effects.

Increase in Porosity and Permeability, Cracking, Loss of Material (Spalling, Scaling)


Question: Rad Liquid Scintillation Analysis

Rad Liquid Scintillation Analysis is a method used to detect and measure specific radioactive isotopes in a sample, whereby certain parameters such as the concentration of Nickel-63 are quantified along with their associated uncertainties.

Similarly, various analytical methods like ASTM C 1475-00 Modified and EPA 905.0 Modified are employed for different testing procedures to ensure accurate results in various scientific analyses.


Question: beach heather habitat requirements

In MN, beach heather is sometimes found on high and sandy beaches of large lakes, well beyond the reach of normal wave action.

All the populations of beach heather in Minnesota are small and restricted to active sand dunes.


Question: What is stress corrosion cracking in stainless steel alloys?

Cracking due to stress corrosion cracking (SCC) could occur in indoor or outdoor SS piping, piping components, and tanks exposed to any air, condensation, or underground environment when the component is: (a) uns insulated; (b) insulated; (c) in the vicinity of insulated components, or (d) in the vicinity of potentially transportable halogens. Cracking can occur in environments containing sufficient halides (e.g., chlorides) in the presence of moisture.

Loss of material is not an aging effect for copper alloys in a condensation environment.


Question: dropped rod accident

Figure 15-182 illustrates the changes in the bank control rod position over time during a dropped rod accident, detailing specific measurements at intervals of 0, 30, 60, 90, 120, and 150 seconds.

The reactor core design incorporates advanced safety features to ensure operational stability and efficiency throughout its lifecycle.


Question: What is the importance of diverse instrumentation in reactor safety?

Diverse instrumentation is critical in ensuring that the reactor protection system can respond reliably during various events, such as a decrease in feedwater temperature or a failure of the steam pressure regulator.

Loss of stator cooling can lead to significant operational challenges in electric generators, but it does not directly impact the instrumentation used for reactor protection.


Question: What is meant by 'ordinary high-water level' in the context of dredged material management?

"Ordinary High-Water Level (OHWL)" means the boundary of waterbasins, watercourses, public waters, and public waters wetlands, and shall be an elevation delineating the highest water level which has been maintained for a sufficient period of time to leave evidence upon the landscape, commonly that point where the natural vegetation shifts from predominantly aquatic to predominantly terrestrial.

A significant storm event is characterized by rainfall greater than 1.0 inches and must occur at least 72 hours after the last measurable storm event.


Question: What does the No Action Alternative mean for nuclear licensing?

The best option for NRC, VC Summer, the communities within a 50-mile radius of VC Summer, and the energy consumers and ratepayers is the No Action Alternative.

Dominion Energy’s home state of Virginia rates 20th out of 51 in the US, while South Carolina ranks next to last at 49th.


Question: TMI-2 accident cleanup

EPRI staffed its TMI-2 site office from 1980 to 1992 with the purpose of transferring TMI-2 accident cleanup data and experience to member utilities and to provide technical support to the cleanup program.

The papers in the proceedings may be purchased from the ANS Web site (http://www.ans.org/pubs/journals/nt/v_87).


Question: license renewal process for nuclear power plants

As part of the license renewal process, the U.S. Nuclear Regulatory Commission (NRC) requires that the license renewal application include an environmental report (ER) that assesses the impacts from continued operation and any refurbishment to be undertaken to enable the continued operation of the units.

Figures depicting the station site and the vicinity within a 6-mile radius of the reactor units (Figures 1 and 2) are attached.


Question: Radioactive isotopes of noble gases, such as krypton and xenon, primarily lead to whole-body doses through external exposure after being released into the atmosphere.

Exposure of a man to an atmosphere contaminated with radioactive isotopes of these elements results only in an external, whole-body dose from submersion in the radioactive cloud.

The total whole-body population exposure occurring as a result of increased cosmic radiation during commercial air travel is estimated at 90,000 man-rem in the United States during 1969 (Reference 22).


Question: The analysis of Radium concentrations is vital due to its regulatory implications in monitoring wells.

Concentrations that are greater than the LLD are represented by a closed symbol.

Insufficient Water for Sampling after May 2012


Question: nuclear licensing process

The nuclear licensing process involves a series of regulatory steps to ensure safety and compliance with established standards, including the submission of detailed technical safety evaluations and environmental assessments by the applicant.

The reactor's coolant system must maintain a constant temperature to ensure efficient energy production throughout the operational cycle.


Question: The assessment of the risk metrics showed an increase in Core Damage Frequency attributable to LOCA debris.

The calculated risk increase due to failures caused by LOCA-generated debris (ΔCDF and ΔLERF) shown in Table 5-4 and Table 5-5, along with the baseline CDF and LERF from Table 5-3, is compared with the guidelines in RG 1.174, as shown in the figures below.

A bounding evaluation was performed for the risk contribution of the SSBI using the PBN IE PRA model.


Question: The NRC stipulates that simulation analyses must incorporate seismic dam failure assessments to ensure compliance with safety regulations...

Fukushima NTTF Recommendation 2.1: Seismic Dam Failure Simulations provide critical guidance for evaluating potential failure modes related to seismic events.

Tennessee Valley Authority SQN Drawing 46W455-30 details the specifications for architectural door frames, including material quality and dimensions.


Question: subsequent license renewal application

The application for a subsequent license renewal (SLR) is intended for facilities that have previously received a license renewal, allowing them to operate beyond the originally licensed period, contingent upon meeting regulatory requirements for safety and environmental protection.

The North Anna Power Station operates on nuclear energy, which is generated through a process of nuclear fission, where atomic nuclei are split to release energy.


Question: What is the condition described by the Lawson Criterion regarding plasma performance?

The Lawson Criterion - nTτE is a measure in magnetic fusion energy indicating the product of plasma density (n), temperature (T), and confinement time (τE) that must be achieved for sustained nuclear fusion reactions.

The operational methods listed include various public projects and private companies involved in fusion energy research, but do not pertain to the Lawson Criterion specifically.