Question: The NRC evaluates the potential cracking mechanisms affecting reactor components exposed to coolant.

Cracking due to stress corrosion cracking, intergranular stress corrosion cracking, cyclic loading

Not applicable. This item is only applicable to BWRs.


Question: What is the EPA Method for analyzing dissolved calcium?

- **Company:** Rio Algom Mining Company
- **Project ID:** DUP-06-07132024
- **Sample ID:** L89001-04
- **Date Sampled:** 07/13/24 00:00
- **Date Received:** 07/16/24
- **Sample Matrix:** Groundwater


Question: The regulations include updating terms from 'reactor' to 'production or utilization facility' to clarify definitions.

In § 20.1905(g), remove the word "reactors" and add in its place the phrase "production or utilization facilities".


Question: What are the responsibilities of a Radiation Safety Officer according to NRC regulations?

The scope of my authority and responsibilities are codified in the Nuclear Regulatory Commission license and regulations.

The letter requesting this transfer of control by the current Radiation Safety Officer (and the Authorized User) is shown below:


Question: What are the steps to operate the MOV 29MOV-74 and 29MOV-77 valves?

Open the following valves: At panel 09-3 and 4 candidate opens 29MOV-74 and 77 by rotating control switches clockwise.

At panel 09-4 candidate jogs opens 29MOV-79 until full open by intermittently rotating control switch clockwise.


Question: What is the methodology used for non-LOCA safety analyses at Byron and Braidwood?

The ARITA topical report (Reference 1) establishes a methodology to address the non-LOCA safety analyses per Chapter 15 of the Standard Review Plan (SRP), with the exception of the control rod ejection event which is covered by a separate topical report.

The ORFEO-GAIA and ORFEO-NMGRID topical (Reference 9) provides the Critical Heat Flux (CHF) correlations and the respective design limits to be used for departure for nucleate boiling (DNB) analyses.


Question: What does SRRS stand for in the context of nuclear training?

SRRS: 3D.105 (when utilized for operator initial or continuing training) provides guidelines for the training and assessment of nuclear plant operators.

The external factors influencing reactor design must be carefully considered during the licensing process.


Question: What are the monitoring requirements for fuel storage in nuclear contexts?

The monitoring of radioactivity releases is essential for ensuring environmental safety.


Question: What is a design-basis earthquake?

A design-basis earthquake (DBE) is defined as a set of ground motions that a nuclear facility must be designed to withstand, ensuring that the structure, systems, and components remain functional and protect the health and safety of the public.

Seismic history provides an understanding of past earthquakes and their effects on the region, but it does not define the parameters for designing nuclear facilities.


Question: The process of sealing a well involves using a slurry mixture to ensure that the annulus remains properly secured.

The cement will consist of a slurry of Class A cement, approximately 2% bentonite gel, and water with a weight of approximately 13 ppg.

Historically, wells completed by under-reaming have demonstrated higher volumetric flow rates over those observed in perforated wells.


Question: How will the reporting of extravasation handle different types of radiopharmaceutical injections?

the reporting should be inclusive of diagnostic and therapeutic if there was a certain dose threshold exceeded or should the extravasation reporting be limited to therapeutic radiopharmaceutical injections if they were extravasated?

IV administration is a common technique and part of the scope of practice for nuclear medicine techs.


Question: location of equipment in nuclear facilities

The equipment is strategically located near Curing Tanks to facilitate efficient operations during the curing process.

The notable area serves as an entry point for personnel and visitors, ensuring smooth access to the main facility.


Question: minimum cooling time for nuclear fuel assemblies

Minimum Cooling Time (years) specified for various Initial Assembly Average Enrichment wt % 235U (E) ranges within the Loading Table.

The reactor core is designed to accommodate a range of thermal power outputs depending on the fuel used and the operational procedures implemented.


Question: Loss of reactor coolant

CHECK RCS subcooling greater than 65°F [85°F ADV].

ENSURE no higher priority exists, THEN **GO TO** 1-FR-H.1, Loss of Secondary Heat Sink.


Question: corrective action program inspections

If any projected inspection results do not meet the acceptance criteria prior to the next scheduled inspection, inspection frequencies are adjusted as determined by the site's Corrective Action Program.

The timing of the additional inspections is based on the severity of the degradation identified and is commensurate with the potential for loss of intended function.


Question: The NRC also analyzes the frequency and direction of wind patterns.

Monthly and Annual Joint Frequency Distributions of Wind Speed and Direction by Atmospheric Stability Class

The various stability classes are determined based on the elevation and atmospheric conditions.


Question: The estimated recurrence intervals for seismic events in the New Madrid Seismic Zone suggest variability over centuries.

Kelson et al. (Reference 109) estimated a preferred recurrence interval of about 400 to 500 years for the Reelfoot fault.

Baldwin et al. (Reference 99) conducted paleoseismic investigations to identify possible evidence of surface deformation related to the January 1812 earthquake.


Question: The NRC stipulates that SG tube inspections shall be conducted periodically to identify any flaws present in the tubes.

Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube plugging criteria.

The operational leakage performance criterion is specified in LCO 3.4.13, “RCS Operational LEAKAGE.”


Question: What is the function of ice baskets in nuclear containment structures?

The function of the ice baskets is to contain brotated in 12 inch diameter columns 48 feet high. The ice absorbs the thermal energy resulting from LOCA or steam line break in the Containment structure. The ice baskets are arranged to promote heat transfer from the steam to ice during and following these accidents.

The vertical support columns and brackets which support the lattice frames are structurally analyzed to determine structural integrity.


Question: A Logic System Functional Test verifies the operability of a logic circuit by checking all components from the sensor to the actuated device.

A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all logic components required for OPERABILITY of a logic circuit, from as close to the sensor as practical up to, but not including the actuated device, to verify OPERABILITY.

The analysis in WCAP-18461-P-A (Reference 54) examines the simulated test path for both these Westinghouse tests that are equivalent to the LGS Logic System Functional and depicts the PPS modules involved (see Figures 7.1-2 and 7.1-3 in WCAP-18461-P-A).


Question: What is environmental qualification in nuclear safety?

EQ is a time-limited aging analysis (TLAA) to be evaluated for the subsequent period of extended operation. See the Standard Review Plan, Section 4.4, "Environmental Qualification (EQ) of Electric Equipment," for acceptable methods for meeting the requirements of 10 CFR 54.21(c)(1)(i) and (ii).

The NRC evaluates licensees' compliance with regulations using various methods, including inspections and assessments of safety protocols.


Question: The NRC's oversight of licensee performance regarding reactor safety is critical for identifying operational trends that may indicate safety concerns.

The inspectors reviewed the licensee’s corrective action program to identify potential trends in degrading maintenance fundamentals that might be indicative of a more significant safety issue (CR1962008).

The inspectors did not identify a performance deficiency or violation of NRC requirements.


Question: The disposal of tritium signs must comply with regulatory standards set by the NRC.

This letter is to provide notification that the signs listed on the attached documents have been shipped to SRB Technologies, Inc. (NC Radioactive Materials License # 034-0534-2) for proper disposal.

These signs were originally installed at: Home Depot #6838, 12727 Fm 1960 West, Houston, TX 77065.


Question: The NRC regulates the qualifications and responsibilities of Radiation Safety Officers in medical and research facilities.

Please add Michelle L. Kritzman, M.S. as the Radiation Safety Officer. Ms. Kritzman was listed as the RSO for 10 CFR 35.100, 35.200 and 35.300 on License No. 21-1858501.

Thank you for your cooperation with this matter. If you have any questions, please contact our physicist, Michelle Kritzman, at (734) 662-3197 or by email at mkritzman@mpcphysics.com.


Question: The Plume Exposure Pathway EPZ is crucial for planning protective actions during a radiological emergency.

Protective Action Sections (PAS) within the established Plume Exposure Pathway (10-mile) EPZ for River Bend Station

The reactor's cooling system is designed to maintain safe operational temperatures and prevent overheating.


Question: separation between liner and moisture barrier

The SLRA states that a “separation between the liner and the moisture barrier 1.5-inches long was identified at approximately 145 degrees.

The staff notes that OE 7 states that after the moisture barrier “sealant was removed the inspector.”


Question: What is the ultimate tensile strength in nuclear engineering?

The ultimate tensile strength (UTS) is a measure of the maximum amount of tensile stress that a material can withstand before failure, and it is critical for assessing the mechanical properties of materials used in nuclear reactors.

The American National Standards Institute (ANSI) provides standards that are crucial for ensuring that nuclear facilities operate safely and efficiently.


Question: The weighted mean activity is an important metric in assessing nuclide contributions in regulatory analysis.

The confidence levels indicate the reliability of predictions in various nuclear scenarios.


Question: What are the control room doses in nuclear safety assessments?

Natural deposition - Powers' Model, Compartment 2
  - Deposition Lambda (1 / Hours)
    - Noble Elemental Organic Aerosol
      - 0.0000E+00  0.0000E+00  9.0078E-02
  - Deposition Net DF
    - Noble Elemental Organic Aerosol
      - 1.0000E+00  1.0000E+00  1.1609E+00


Question: average enrichment levels for PWR fuel

Minimum Initial Assembly Avg. Enrichment wt% 235U (E) | 30 < Assembly Average Burnup ≤ 32.5 GWd/MTU | Minimum Cooling Time (years)

The decay heat from the reactor must be adequately managed after shutdown to ensure safety.


Question: What is the activity of Eu-152 in pCi/grams?

EU-152 has an activity of 4.40E-02 pCi/grams.

The nuclide PA-228 has an energy of 39.95 keV and a yield of 22.00%.


Question: What are the verification requirements for the LMTC Neutron Shield during draining operations?

If draining the LMTC Neutron Shield is required to meet the plant architectural limits, the LMTC Neutron Shield shall be verified to be filled after completion of the critical lift.

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


Question: What is the significance of Cs-137 concentration in nuclear safety assessments?

Figure 16-3 shows the histogram for Cs-137 concentration, indicating the distribution of upper end values measured in picocuries per gram (pCi/g).

The upper end values for various measurements can reveal patterns in distribution but must be interpreted carefully.


Question: What is the significance of Nuclide Minimum Detectable Activity (MDA) in nuclear safety assessments?

Nuclide MDA (Minimum Detectable Activity) is an important metric that indicates the lowest amount of activity of a radionuclide that can be reliably detected in a sample. For example, K-40 has a Line MDA of 5.89E-01 pC/g.

The Peak Locate was performed on the specified channel at a given date and time, ensuring that the peaks are accurately identified and characterized.


Question: What is the purpose of a valve body in the turbine steam supply system?

The valve body is defined as a leakage boundary (spatial) which plays a crucial role in preventing leaks in the turbine steam supply system.

The aging management program for valve bodies includes routine inspections and ongoing monitoring to ensure functionality and safety compliance.


Question: What are the criteria to reopen a closed record in a regulatory filing?

To reopen a closed record, a petitioner must file a motion that demonstrates: 1. its new contention is timely based upon the availability of new information; 2. the contention addresses a significant safety or environmental issue; and 3. a materially different result would be or would have been likely had the newly proffered evidence been considered initially.

The petitioner must attach an affidavit from 'experts in the disciplines appropriate to the issues raised' or from 'competent individuals with knowledge of the facts alleged' that separately addresses each of these criteria, explaining how each criterion has been satisfied.


Question: Assessment of safety and control mechanisms is essential for reactor licensing.

Licensees must establish appropriate safety analyses to demonstrate compliance with regulatory requirements during the licensing process.

The cooling system is designed to remove heat from the reactor core and prevent overheating during normal operations.


Question: groundwater transfer

a. Groundwater Transfer       |   X  |     | (if yes, complete part 2)

### 2. LIST WELLS USED IN GROUNDWATER TRANSFER DURING THE MONTH:
*(No wells listed)*


Question: reserve auxiliary transformer failure

The inspectors reviewed NextEra's corrective action program to verify that the licensee identified, evaluated, and implemented corrective actions, as needed, for issues involving the reserve auxiliary transformer.

The inspectors reviewed the licensee's corrective action program for potential adverse trends that might be indicative of a more significant safety issue.


Question: What is negative reactivity in reactor kinetics?

When negative reactivity is inserted, the power drops quickly prior to reaching a stable reactor period.

The prompt neutron population is affected before the delayed neutron population.


Question: What is a Small Break Loss of Coolant Accident (LOCA)?

A Small Break Loss of Coolant Accident (LOCA) is characterized by a rupture in the piping system where the break area is less than one square foot, which can result in a significant decrease in coolant levels in the reactor and requires rapid emergency response actions to maintain safety.

The reactor's control systems continuously monitor various safety parameters to ensure that operations remain within established limits.


Question: The NRC conducted an evaluation of fire safety systems at the facility to ensure compliance with regulatory standards.

No findings or violations of more than minor significance were identified during this inspection.

No additional details about the operational metrics were provided in the report.


Question: What is the proposed 'dose limit' in nuclear regulatory filings?

The proposed 'dose limit'--really ANY dose limit--is a calculation done by applicant using unverified assumptions to claim the site will be acceptable.

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


Question: What are the requirements for proposing alternatives to ASME Code in nuclear regulatory filings?

Proposed alternatives to ASME Code, Section III, 2001 Edition with Addenda through 2003, other than the alternatives listed in Table 2.1-2 of the FSAR, may be used when authorized by the Director of the Office of Nuclear Material Safety and Safeguards or designee. The request for such alternatives should demonstrate that: 1. The proposed alternatives would provide an acceptable level of quality and safety, or 2. Compliance with the specified requirements of ASME Code, Section III, Subsections NB and NG, 2001 Edition with Addenda through 2003, would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

This section presents site-specific parameters and analytical bases that must be verified by the MAGNASTOR® SYSTEM user. The parameters and bases presented in Section 4.3.1 are those applied in the design basis analysis.


Question: The maximum still water level at the site is determined by factoring in the effects of wind waves and other hydrological parameters.

By adding the backwater effect of 0.2 ft to the PMF still water elevation of 264.07 ft msl at the dam, the PMF still water elevation at the site is 264.27 ft msl.

The fetch diagram used to determine an effective fetch length of 4700 ft with a maximum fetch of 10,600 ft is shown on Figure 2.4-12.


Question: average river water temperature

The average monthly river temperature at the intake structure shows a significant variation throughout the year, ranging from a low of 30.0 °F in January to a high of 80.0 °F in August.

The discharge canal is critical for ensuring environmental compliance during the operation of a nuclear power plant.


Question: What is the significance of channel independence in nuclear protection systems?

The criteria provided in IEEE 279-1971 establish the minimum requirements for the safety-related functional performance and reliability of protection systems for nuclear power generating stations. Specifically, Section 4.6 regarding channel independence states: Channels that provide signals for the same protective function shall be independent and physically separated to accomplish decoupling of the effects of unsafe environmental factors, electric transients, and physical accident consequences documented in the design basis, and to reduce the likelihood of interactions between channels during maintenance operations or in the event of channel malfunction.

Application of the regulation in this circumstance would not serve the underlying purpose of the rule and is not necessary to achieve the underlying purpose of the rule.


Question: What are the requirements for technical specification changes in a BWR license renewal application?

10 CFR 54.22 requires that each application for license renewal include any technical specification changes (and the justification for the changes) or additions necessary to manage the effects of aging during the period of extended operation as part of the renewal application.

There are no changes to technical specifications that are required to meet the requirements of the BWRVIP reports during the subsequent period of extended operation.


Question: total well depth measurement

The data collected over the years shows significant variations in groundwater levels.


Question: Data A failure in nuclear systems

- DATA A FAILURE                                                                                |     |       |                |

- PLACE the ACCURACY MODE Switch (S106) on the back of the DRPI display panel in the "B ONLY" position.


Question: annual weather data

The Annual Summary of Dry Bulb Temperature, Relative Humidity, Wind Speed and Precipitation from Tower D for 2018 provides detailed monthly climate statistics including maximum, minimum, and average temperatures, humidity levels, wind speeds, and total precipitation for each month of the year.

The reports included various updates to the safety protocols for the reactor, emphasizing the importance of regular inspections and compliance with the latest nuclear safety standards.


Question: The NRC facilitates pre-application discussions to foster clarity in the licensing process.

The NRC encourages preapplication engagements to establish a common understanding on the application to support regulatory stability and review timeliness.

The NRC is committed to reviewing a license application in an efficient and effective manner.


Question: What are the requirements for containment penetrations during refueling operations?

The containment penetrations shall be in the following status: a. The equipment hatch closed and held in place by four bolts, or if open, capable of being closed; b. One door in the emergency air lock and one door in the personnel air lock capable of being closed; and c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere either: 1. closed by a manual or automatic isolation valve, blind flange, or equivalent, or 2. capable of being closed by an OPERABLE Containment Purge Isolation valve.

The CORE ALTERATIONS involve detailed planning and execution to minimize the time that personnel are exposed to radiation.


Question: feedwater isolation valves testing

The feedwater isolation valves are routinely tested during refueling outages.

These functions are not directly related to safety, but are provided for precautionary and economical reasons and are not subject to the same testability guidelines as the protection system.


Question: Licensees must adhere to specific criteria when applying for renewals of operating licenses.

The license renewal process allows operators to apply for the continued operation of their facilities beyond the original licensing period, ensuring that safety and regulatory standards are continually met.

The transport of fuel rods is strictly monitored to prevent environmental contamination during the movement process.


Question: What is the function of the Control Room Emergency Ventilation System (CREVS)?

The Control Room Emergency Ventilation System (CREVS) processes outside air needed to provide ventilation and pressurization for the Control Room Habitability Zone (CRHZ) during isolated conditions.

There are several separate subsystems serving these areas.


Question: average burnup in BWR fuel

The assembly average burnup is a measure of the energy produced by the nuclear fuel in terms of gigawatt-days per metric ton of uranium (GWd/MTU).

The minimum cooling time prior to reloading the fuel assembly is critical to ensuring that the assembly has sufficiently cooled to meet regulatory safety requirements.


Question: nuclide activity measurement

The table lists various nuclides along with their respective weighted mean activity measured in pCi/grams, providing crucial data for assessing radiation levels in regulatory filings.

The interference analysis is used to determine which nuclides are valid for the measurement process, ensuring that the data collected is accurate and reliable.


Question: The NRC specifies rigorous testing protocols for debris management during performance assessments of systems.

During conventional debris addition, test tank water level was maintained within the acceptable limits by diverting a portion of the returning flow to the transition tank (instead of back to the test tank) via the three-way valve.

Both head loss tests were conducted under the same test conditions except for variations in debris batching and quantities.


Question: The system is designed to alert operators about valve status and potential issues related to alarm activations.

An alarm annunciator point is activated by both a valve motor operator limit switch and by a valve position limit switch activated by stem travel whenever an accumulator valve is not fully open for any reason with the system at pressure (the pressure at which the safety injection block is unblocked).

A separate annunciator point is used for each accumulator valve. This alarm is received at approximately one hour intervals to remind the operator of the improper valve lineup.


Question: What is a regulatory audit summary in the context of nuclear licensing?

A regulatory audit summary will be provided within 90 days of the completion of the audit.

Audit dates are Monday, October 21, 2024 – Thursday, October 24, 2024.


Question: Cause Failure Coping Analysis

The Cause Failure Coping Analysis is a fundamental aspect of the defense in depth and diversity philosophy, providing a systematic approach to identifying potential failure mechanisms and ensuring the safety of nuclear reactor operations.

The Digital Modernization Project at the Limerick Generating Stations involves upgrading the existing systems to improve reliability and efficiency across various operational domains.


Question: Regulatory frameworks demand that operators perform detailed safety assessments as part of their licensing process.

The review of the safety analysis report is essential to ensure compliance with regulatory requirements during the licensing of nuclear facilities.

Nuclear reactors are subject to various mechanical stresses that can impact their overall performance in different operating conditions.


Question: aging management program for structural components

Aging Management Program: Structures Monitoring Program (B.2.3.33) is designed to manage the aging effects requiring management, such as loss of material.

Penetration Seals are used as flood barrier shelters for structural protection.


Question: AmERICIUM-241 regulatory filings


Question: The distances considered in reactor siting include radial measurements to various critical boundaries and zones.

Power reactor siting has typically involved assessment of a variety of distances, most of which are depicted in Fig. 1

Siting criteria protect from societal impacts & provides DiD to minimize societal impacts should containment fail.


Question: non-seismic liquefaction potential

NUREG-0800 states that non-seismic liquefaction potential from erosion, floods, wind loads on structures, and area weight should be considered.

The excavation for the power block will be bounded by a structural support system located approximately 850 ft. from the centerline of the nuclear island structures depending on the technology selected.


Question: What is the nominal trip setpoint (NTSP) in nuclear relay systems?

To determine the NTSP for a decreasing setpoint: NTSP = AL + (Z* ) and ATSP = AL + (Z* + MAR) [Eq. 3.1.2.b]

The voltage drop from the VTs to the degraded voltage relays and ratio correction factor includes measurements for Div 1 and Div 2 Phase AB relays.


Question: The procedure AOP-1 outlines the necessary actions to take in the event of a failure to successfully scram the reactor, ensuring that key operational parameters are maintained.

AOP-1 requires maintaining Reactor water level in a band of 180-220°. AOP-1 also provides guidance to insert control rods using either RMCS or manual scrams in Attachments 4 and 5.

B. Incorrect – EP-3 use is not procedurally allowed in these conditions. Plausible because EP-3 would be correct if EOP entry was required and EP-3 provides more methods for attempting control rod insertion.


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

Nuclide MDA (Minimum Detectable Activity) is a critical measurement used in radiation detection and analysis that indicates the lowest level of activity that can be reliably detected by a given measurement system.

The yield percentages of various nuclides, such as EU-152 and I-131, provide insights into the efficiency of radioactive decay processes.


Question: What is thermal aging embrittlement in cast austenitic stainless steel?

Thermal Aging Embrittlement of Cast Austenitic Stainless Steel (CASS) is a phenomenon that can affect the integrity and performance of nuclear reactor components constructed from this material.

The Open-Cycle Cooling Water System is designed to circulate water to remove heat from the reactor and discharge it back to the environment.


Question: The NRC mandates that a complete visual inspection must be conducted on all submerged surfaces for Brunswick Units during the inspection cycle.

In accordance with ASME Section XI, 2007 Edition through 2008 Addenda, Table IWE-2500-1, Examination Category E-A, Item Number E1.12, Wetted Surfaces of Submerged Areas shall receive a visual (VT-3) examination on 100% of the submerged surfaces during the inspection interval.

Prior to continued service, protective coating restoration was applied to the submerged surfaces in all sixteen (16) torus bays to arrest further corrosive degradation.


Question: What measures are taken to ensure confidence in identified nuclides?

Nuclide confidence index threshold = 0.30


Question: source strength in nuclear physics

The source strength is defined as the measure of the energy emitted by a radioactive source, expressed in MeV/cm³-s, which indicates the intensity of radiation per unit volume over time.

The various energy groups listed indicate the range of energy levels of the emitted gamma radiation, categorized within specific MeV ranges.


Question: The assessment of safety systems is crucial for maintaining regulatory compliance in nuclear operations.

Diversity and Defense-in-Depth Evaluation is essential in understanding the safety features and their operational effectiveness within the nuclear facility's design.

The company undertook several initiatives to reduce operational costs and improve efficiency across its plants.


Question: What is the significance of NELAC certification in laboratory analyses?

Except as noted, the test results for the methods and parameters listed on ACZ's current NELAC certificate letter (#ACZ) meet all requirements of NELAC.

ACZ Laboratories, Inc. (ACZ) on October 05, 2024 and originally reported on November 04, 2024.


Question: The NRC expects that licensees will submit detailed safety evaluations based on relevant nuclear regulations.

The NRC requires that all licensees submit a comprehensive safety analysis report that demonstrates compliance with regulatory requirements.

The annual inspections of the reactor are performed to verify operational efficiency and maintenance schedules.


Question: nuclide minimum detectable activity

The Nuclide MDA (Minimum Detectable Activity) provides a threshold for the detection of specific radionuclides in samples, measured in pCi/grams.

The table lists various peak energies, sizes, and types of unidentified peaks observed during the analysis.


Question: X/Q values in atmospheric dispersion

Relative concentrations of released gases, X/Q values, as a function of direction for various time periods at the exclusion area boundary (EAB) and the outer boundary of the low population zone (LPZ), were determined by the use of the computer code PAVAN (Reference 43). This code implements the guidance provided in Regulatory Guide 1.145, "Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants," August 1979. The X/Q calculations are based on the theory that material released to the atmosphere will be normally distributed (Gaussian) about the plume centerline.

For this ESP application, calculations to determine diffusion estimates for both short- and long-term conditions were completed using data from the meteorological instrumentation in service prior to the most recent replacement in December 2000, as described in Section 2.3.3.1. Data recovery for the period evaluated in the calculations (Sections 2.3.4 and 2.3.5) is indicated in Table 2.3-131.


Question: What is the purpose of a reactor protection system in nuclear reactors?

The reactor protection system (RPS) is designed to automatically initiate actions that ensure the safety of the reactor during abnormal conditions, helping to prevent accidents or mitigate their consequences.

The Idaho National Laboratory's report includes various statistical analyses related to component performance, without delving into the details of reactor safety systems.


Question: What is nuclide identification in nuclear regulatory filings?

The Nuclide Identification Report details the identified nuclides along with their corresponding confidence levels, energy, yield, and activity measurements which are critical for understanding the radiological characteristics of materials.

The identified nuclides include K-40, CS-137, and PB-212, each of which has distinct properties that can impact their handling and assessment in various safety contexts.


Question: residual heat removal system

Residual heat removal loop in service. At least one pump operating through an OPERABLE hx. (In service equipment)

Standard: Verifies requirements are met.


Question: What are the heavy loads requirements for the MAGNASTOR® TSC?

Each lift of a MAGNASTOR® TSC, transfer cask, or concrete cask must be made in accordance with the heavy loads requirements and procedures of the licensed facility at which the lift is made. A plant-specific safety review (under 10 CFR 50.59 or 10 CFR 72.48 requirements, if applicable) is required to show operational compliance with existing plant-specific heavy loads requirements.

The holder of this certificate who desires to make changes to the certificate, which includes Appendix A (Technical Specifications and Design Features) and Appendix B (Approved Contents), shall submit an application for amendment of the certificate.


Question: In the analysis of the subsurface materials, various types of soil were identified, including medium to coarse grained sand and stiff clay.

The boring log classifications reveal types of materials found at specified depths, such as gray, medium to coarse grained sand and stiff clay at depths of 20.0 and 35.0 feet.


Question: The seismic hazard results inform the selection of input motions for the site response analysis.

Rock input motions are selected for input to the site response calculations using the seismic hazard and deaggregation results.

The correlation coefficient between ln(Vs) in adjacent layers is estimated using the inter-layer correlation model from Toro (Reference 2.5.2-84) for USGS category A+B, which corresponds to Vs30 values of 360 m/s (~1,180 ft/s) or greater.


Question: Licensees must demonstrate adequate safety measures within their applications for operational compliance.

The nuclear licensing process includes a comprehensive safety assessment to ensure that the facility can operate without posing undue risk to the public and environment.

In the event of a breach of safety protocols, the facility must implement corrective actions to address the identified deficiencies.


Question: A requirement for inspections related to the operational functionality of the secondary cooling system.

Defect with Detroit Diesel/MTU Fuel Injectors


Question: What is Americium-243 and its significance in a nuclear context?

Americium-243 is a radioactive isotope that can be measured in soil samples, and its presence is significant for environmental monitoring and regulatory compliance in nuclear safety assessments.

The report lists various parameters and methods for analyzing soil samples, including Rad Alpha Spec Analysis and Liquid Scintillation Analysis.


Question: Residual Heat Removal System design requirements

Containment Spray and Suppression Pool Cooling are non-ECCS modes of RHR operation that are manually placed into operation to limit the post-accident temperatures of the primary containment atmosphere and suppression pool water.

The originally installed ECCS torus suction strainers were of a traditional witch’s hat design (i.e., cone-shaped design) with hydraulic head loss typically assigned assuming blockage of 50% of the available surface area.


Question: The safety-related raceway systems must remain functional during seismic events as per regulatory requirements.

The DCPP Class 1E raceway systems (safety-related) consist of conduits, cable trays, pull boxes, and supports.

Normal and upset condition operating temperatures less than 220°F in carbon steel components, or 270°F in stainless steel components, will not produce significant thermal stresses, and therefore will not produce significant geometric stress intensification factors or significant residual stresses.


Question: When effluent releases are made, grab samples must be taken and analyzed for noble gases at regular intervals.

effluent releases via this pathway may continue provided grab samples are taken at least once per 12 hours and these samples are analyzed for noble gases (gamma emitters) in accordance with plant procedures.

The performance step's criticality is determined by assessing the actions required when the system is compromised.


Question: Licensees must ensure compliance with the applicable regulations regarding the formulation of their safety analysis reports.

The safety analysis report is a comprehensive document that describes the analyses related to the safety of the facility and demonstrates compliance with regulatory requirements.

To ensure effective communication, stakeholders must participate in regular meetings and collaborate on project goals.


Question: What is the flexural strength of different coke types used in nuclear applications?


Question: What is the RIPE approach in nuclear regulatory filings?

RIPE is an approach that is acceptable to the NRC for developing a risk-informed application for an exemption request or license amendment request that applies risk insights, consistent with the guidance in Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 3 (Reference 10) and as applicable RG 1.177, "Plant-Specific, Risk-Informed Decision-making: Technical Specifications," Revision 2 (Reference 11).

The NRC Integrated Inspection report for HNP from November 10, 2022 (Reference 1) includes documentation of a green finding and associated non-cited violation of Title 10 of the Code of Federal Regulations (10 CFR) Section 50 Appendix B, Criterion III, "Design Control" for "Treatment of Class 1E Interfaces and Interlocks with the Turbine Trip System (TTS) Design."


Question: The aging management program for coatings and linings in piping and tanks assesses integrity through regular inspections and defined acceptance criteria.

The Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks program will include periodic visual inspections to verify the integrity of internal coatings designed to adhere to and protect the base metal.

Coatings/linings that do not meet acceptance criteria are repaired, replaced, or removed.


Question: Individuals with disabilities may request specific accommodations to engage fully 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: catch rates near Browns Ferry Nuclear Plant

Table 3.6-4 presents the Catch Rates (CPUE; in the Vicinity of Browns Ferry Nuclear Plant (Tennessee River Miles 292.5 and 295.9) from 1993 through 2020, detailing various fish species and their average CPUE.

The average annual rainfall in Limestone County is approximately 54 inches, which significantly affects the local ecosystems and wildlife.


Question: What is the formula for calculating the stud force in the reactor head?

The mathematical form of the force distribution in the studs consequently may be expressed as follows, where **f** is a constant: **Fi** = **F** + **f (xi - δ)**.

The coordinates **xi** can be written in terms of the bolt-circle radius (**Ro**) and **θi** as defined in Figure 1.


Question: What is a construction permit application in nuclear regulatory filings?

By letter dated March 28, 2024, TerraPower, LLC (TerraPower), on behalf of its wholly owned subsidiary US SFR Owner, LLC (USO), submitted a construction permit (CP) application and corresponding preliminary safety analysis report (PSAR) to the U.S. Nuclear Regulatory Commission (NRC) staff for a reactor facility pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Part 50, "Domestic Licensing of Production and Utilization Facilities," and section 103 of the Atomic Energy Act of 1954, as amended.

The NRC staff is conducting a detailed review of the Kemmerer 1 CP application. This regulatory audit is intended to assist the NRC staff in its review of the Kemmerer 1 CP application.


Question: nuclide minimum detectable activity (MDA)

A nuclide minimum detectable activity (MDA) is defined as the lowest level of a nuclide that can be reliably detected in a sample, often expressed in units of pCi/grams.

A fitted singlet is a result in spectroscopy that indicates the presence of a singular peak that has been adjusted for better clarity in the data.


Question: cyber security program requirements

Licensees must protect those digital assets that, if compromised, would adversely impact security functions.

In summary, it is expected that a licensee’s evaluation of necessary changes to their Security Plans would likely conclude: the change does not affect compliance with any regulatory requirement.


Question: What is the impact of varying fuel grain size on fission gas release and fuel densification?

A lower limit of [ ] has been defined for fuel grain size in Section 4.6 of Reference 1. This lower limit is supported by the range of fuel grain size in the GALILEO calibration and validation database.

A limit of 10 percent by weight for the gadolinium level in UO₂-Gd₂O₃ rods is specified in Table 4-1 of the V&V document, but not in ANP-10323P.


Question: fire protection systems in nuclear facilities

Automatic detection with local alarm and alarm and annunciation to the Control Room is provided.

Portable emergency lights are provided in the Control Room.


Question: The NRC's electronic hearing docket is essential for transparency in regulatory proceedings.

Documents submitted in adjudicatory proceedings will appear in the NRC’s electronic hearing docket, which is available to the public at https://adams.nrc.gov/ehd, unless excluded pursuant to an order of the Commission or the presiding officer.

Attorney for licensee: Sherry Quirk, Executive Vice President and General Counsel, Tennessee Valley Authority, 400 West Summit Hill Drive, WT 6A, Knoxville, TN 37902.


Question: The distribution of combined experimental data for the ORFEO-HMP correlation highlights relationships between critical heat flux and mass flux under varying pressure conditions.

The distribution of the combined data set for the ORFEO-HMP correlation (pressure vs. mass flux) 7-23

The axial geometry of CHF test CE400 was designed to investigate the effects of flow on heat transfer characteristics.


Question: A summary of the meeting regarding nuclear regulatory discussions can be reviewed at ADAMS Package Accession No. ML25021A001.

The meeting notice is available at ADAMS Accession No. ML25003A086.

The enclosure includes a list of attendees that participated in the meeting.


Question: operating inverters regulation

Inverters - Operating: GGNS TS do not contain an LCO specific to inverters.

Distribution Systems - Operating: In this Condition, the remaining AC electrical power distribution systems can support the minimum safety functions necessary to shut down the reactor and maintain in a safe shutdown condition.


Question: radionuclide analysis

Rad Alpha Spec Analysis is conducted to measure the concentration of radioactive isotopes in a solid matrix, providing crucial data for safety and regulatory compliance.

The report includes detailed client sample information such as the location, collector's name, and specific dates associated with the sampling process.


Question: rate of water discharge from reactor cooling systems


Question: What is the unit of measurement used for gamma radiation in the context of nuclear sampling?

The measurement is reported in counts per minute (cpm), which quantifies the rate of gamma radiation detected in a given area.

The sampling location for the data collection was Pickard Hall, where various boring numbers were recorded.


Question: What is the significance of density in drilling reports for nuclear projects?

Density is categorized based on the number of blows per foot (BLOWS/FT) indicating the soil's consistency, which ranges from very loose to very dense, providing crucial information for construction and safety assessments.

The Westinghouse Columbia Fuel Fabrication Facility is a major project focused on the development of advanced fuel technologies for nuclear reactors.


Question: What is peak xenon in nuclear reactors?

Peak xenon occurs approximately 7-9 hours after a trip from 100% power, when Xe-135 adds 'net' positive reactivity post-trip, returning to pre-trip values around 24 hours after the trip.

Initially an increase in SDM as [Xe] increases.


Question: What does the term 'pCi/g' represent in nuclear licensing?

The unit 'pCi/g' refers to picocuries per gram, which is a measurement used to denote the concentration of radioactive isotopes in a sample of material.

LaCrosseSolutions has been involved in various projects to ensure compliance with environmental regulations in nuclear safety.


Question: What is the significance of License Amendment Requests in nuclear regulatory filings?

A License Amendment Request (LAR) is a formal request submitted to the NRC seeking changes to the existing licenses that permit the operation of nuclear reactors.

The reactor coolant system plays a crucial role in removing heat from the reactor core during operation and must maintain specific pressure levels to ensure safety.


Question: The NRC stipulates that accurate identification of nuclides is essential for ensuring safety in nuclear operations.

The table lists various identified nuclides along with their respective energies, yields, and activities.

The identification report also includes uncertainty assessments for the measurements taken.


Question: design of safety-related structures

The RXB incorporates SC walls designed according to AISC N690-18 using element- and panel-based approaches.

The simulations were performed using SASSI and SAP2000.


Question: permit modification process

This permit may be modified or revoked and reissued, in whole or in part, during its term for cause, including but not limited to, the following: If cause for termination under Provision II. E. 5. of this permit exists, the Director may choose to revoke and reissue this permit instead of terminating the permit;

Whenever the Director is notified of a change in name, ownership or control, he may decide not to modify the existing permit and require the submission of a new permit application.


Question: What is the Engineered Safeguards Actuation System (ESAS) in nuclear instrumentation?

Three ESAS analog instrument channels for each Parameter in Table 3.3.5-1 shall be OPERABLE.

Separate Condition entry is allowed for each Parameter.


Question: What are the licensing requirements for a nuclear reactor?

The licensing process for a nuclear reactor involves a series of comprehensive safety assessments and regulatory submissions to the NRC to ensure compliance with safety standards.

Helion Energy is focused on developing advanced fusion technologies that could revolutionize the energy sector.


Question: nuclear safety licensing requirements

The Nuclear Regulatory Commission (NRC) establishes licensing requirements for ensuring the safety and security of nuclear facilities, which includes evaluating safety reports and compliance with regulations before granting a license.

Many computer systems rely on the operating software to enhance performance and user experience, but this does not apply to regulatory frameworks in nuclear licensing.


Question: The analysis of the adiabatic temperature increase is crucial for understanding the behavior of the reactor during transient conditions.

**Adiabatic Temperature Increase in Wet Well** – BWR/3 ATWS PRFO SLCS Case initial 600 seconds

The reactor core is designed with safety features that ensure stable operation under normal conditions.


Question: adjusted gross OpDCGL

The term adjusted gross OpDCGL refers to the calculated value derived from gross activity measurements and is used to assess compliance with safety standards in nuclear regulatory filings.

Nuclear measurements are crucial for ensuring compliance with safety standards during operations.


Question: The use of Zircaloy rods in tests is limited due to their reaction with water, impacting the cooling process.

Only 5 of the 143 FLECHT tests utilized [Zircaloy] rods. The remaining 138 tests were conducted with stainless steel...rods. Since...[Zircaloy] reacts exothermically with water at elevated temperatures, contributing additional energy to that of the decaying fission products, the application of water to the core has the potential of increasing the heat input to the fuel rods rather than cooling them, as desired.

They could be used repeatedly in testing (for 30 or 40 individual tests) without substantial changes in response over the series.


Question: thermal hydraulic design reactor

The thermal and hydraulic design of the reactor core encompasses considerations such as critical heat flux ratios, coolant flow distribution, and thermal effects of operational transients.

The control rod patterns and reactivity worths are essential in achieving desired power distributions and maintaining reactor stability during operations.


Question: What are the procedures for evaluating Unit 2 containment ventilation?

When contacted as Unit 2, acknowledge request to evaluate Unit 2 containment ventilation status.

When contacted as AUO, acknowledge request to perform Attachment 1 of 1-E-0. Wait 5 minutes and report Attachment 1 is complete.


Question: The comparison of results between the GOTHIC and COPATTA models demonstrates the energy contained in the reactor coolant system.

Reactor Vessel Energy (Btu) is the amount of energy (relative to 32°F) contained in the RCS water inventory that remains in the RCS.

The GOTHIC results are graphically identical to the COPATTA results for the vessel energy and are considered valid.


Question: maximum flood elevation Mississippi River


Question: The procedure outlines the necessary steps for initiating an RCP to ensure proper reactor cooling and maintain pressurizer spray functionality.

Forced coolant flow is the preferred mode of operation to allow for normal RCS cooldown and provide PRZR spray.

Depressurization of the RCS may generate a steam bubble in the upper head region of the reactor vessel if no RCP is running.


Question: The projected growth in electrical demand is significant, showing that TVA anticipates higher peak loads in the coming years due to increased requirements.

Considering the increased AEC demand, TVA is projecting a growth rate of 6.1 percent for its peak load between 1970 and 1975.

Estimates of future electrical loads on the TVA system are prepared by the applicant considering past load trends along with factors which may have an impact on future growth.


Question: fuel assembly locations in a BWR

In a boiling water reactor (BWR), the fuel assembly locations are marked by blue circles (🔵) in the assembly basket pattern.

The designated nonfuel locations are indicated by a red cross (❌) in the layout diagram.


Question: nuclear quality assurance program

The quality assurance program maintains rigorous standards to ensure that nuclear components and systems are designed, manufactured, tested, and operated according to regulatory requirements and safety standards.

The reactor containment structure is designed to withstand operational pressures and provides a robust barrier against the release of radioactive materials.


Question: What happens when high voltage power loss occurs to SRM B?

Loss of high voltage power to SRM B causes an SRM INOP condition to be detected and provides input to the rod block circuitry. However, all SRM rod blocks are bypassed with IRMs on Range 8 or higher, so no rod block actually occurs.

Only 1 SRM can be bypassed at a time. Therefore, SRM B cannot be bypassed while SRM A is still bypassed.


Question: nuclear reactor regulatory requirements

The regulatory requirements for a nuclear reactor are detailed in 10 CFR Part 50, which outlines the licensing process, including construction and operation regulations mandated by the Nuclear Regulatory Commission (NRC).

Further analysis of site demographics and economic indicators must be conducted to assess the long-term viability of the proposed project.


Question: The assessment of safety parameters is crucial for effective nuclear licensing.

The safety analysis of a proposed design is an essential part of the licensing process.

Design modifications must also ensure compliance with environmental protection regulations.


Question: Control Room Ventilation Isolation Signal

A Safety injection signal initiates a Containment Isolation Phase A, which initiates a CRVIS.

The LCO requirements ensure that instrumentation necessary to initiate the CREVS is OPERABLE.


Question: The NRC guidelines recommend analyzing the distribution of the structural mass using consistent mass matrices within the design framework.

The structural mass of the lower support structure is represented automatically in the ANSYS program through the use of consistent mass matrices associated with each of the structural finite elements.

Displacement boundary conditions are not specified for the tops of the columns nor for other nodes contained in the column radial plane.


Question: The analysis of stress conditions during cyclic loading is crucial for understanding material behavior.

Cyclic deviator stress, a critical parameter in geotechnical engineering, is defined as the difference between the maximum and minimum effective stresses that occur in soils during cyclic loading conditions.

Geotechnical Engineers Inc. provides a range of services related to soil analysis and foundation design, ensuring safe and efficient engineering practices.


Question: license transfer application

The December 8, 2020 public teleconference was held to discuss the pre-submittal of a license transfer application from Entergy Nuclear Operations Inc. to Holtec International for the Palisades Nuclear Plant.

The discussions included various topics related to technical specifications and operational procedures of the plant, including maintenance schedules and operational readiness assessments.


Question: The applicant must specify the Class 2 piping that will be examined for compliance during the Preservice Inspection phase.

The phrase, “repair and replacement of Class MC vessels,” means “repair and replacement of Class MC vessels and components (systems).” Acceptance of this Code Case in no way provides/constitutes NRC approval to violate the technical specification or any NRC requirements with regard to breaching containment during repair and replacement procedures while the plant is in operation.

Table 5 lists Code Cases that have been superseded by revision.


Question: ARITA methodology

Absent justification for the insignificance of primary system and secondary system metal heat capacity on applicable FOMs, licensees implementing the ARITA methodology shall [ ] irrespective of evaluation model variant, where the primary system significantly cools down relative to its initial condition, [ ]

In the absence of plant-specific data or other information conclusively demonstrating the ability to determine the reactor pressure vessel upper head fluid temperature to within [ ] by the ARITA methodology, licensees adopting the ARITA methodology shall [ ]


Question: The performance of the Zr-4 M/P alloy in hydrogen absorption significantly impacts reactor safety assessments.

Figure 3-7 illustrates the comparison between the zircaloy-4 material properties (M/P) and the measured hydrogen levels in the reactor system.

The operational procedures for refueling must comply with strict safety guidelines and regulatory reviews, ensuring minimal risk during operations.


Question: nuclear reactor safety analysis

The safety analysis report is essential for assessing the performance and safety of the reactor under normal and postulated accident conditions, ensuring compliance with regulatory safety standards.

The technical specifications for materials must include detailed descriptions of the chemical composition and physical properties of the alloys used in the construction.


Question: Emergency core cooling system requirements for licensing

Refers to Technical Requirements Manual (TRM): - TRM 8.1.1, Boration Systems – Operating Condition A (Enters action statements to restore E CCP to operable status within 72 hours.)

The applicant’s actions included adjusting the CRV-470 controller to 50% and checking charging flow to maintain above 75 gpm.


Question: What is the purpose of the exemption from 10 CFR Part 26 for nuclear plants?

The U.S. Nuclear Regulatory Commission (NRC) has approved the requested exemption from specific requirements of Title 10 of the Code of Federal Regulations (10 CFR) 26, “Fitness for Duty Programs,” Section 26.205, “Work hours,” for Edwin I. Hatch Nuclear Plant, Units Nos. 1 and 2 (Hatch).

This action is in response to the application submitted by Southern Nuclear Operating Company Inc. (SNC, the licensee), dated January 6, 2021 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML21006A227).


Question: The transfer of devices with General Licenses must adhere to regulations outlined in 10 CFR 32.52.

Schlumberger Technology Corporation (STC) made the following transfers of the Schlumberger Vx Spectra Series and PhaseWatcher Vx Series multiphase flow meters (General License TX-586-D-114-G) during the period of October 01, 2024 to December 31, 2024.

Please contact me if you require any additional information.


Question: What are the effects of radiopharmaceutical extravasations in nuclear medicine?

I have seen no evidence in practice that the diagnostic radiopharmaceutical extravasations can cause harm to the patients.

I was never aware of any immediate or deterministic effects as a result of a prior diagnostic radiopharmaceutical extravasation.


Question: The licensing process for nuclear facilities mandates comprehensive safety evaluations to ensure public health and environmental protection.

The NRC reviews and approves applications for new nuclear reactor licenses, which includes detailed assessments of the applicant's safety and environmental plans.

Filings related to the construction of new nuclear plants often include economic analyses that consider market viability and financing options.


Question: nuclear reactor safety systems

Nuclear reactor safety systems are designed to prevent accidents and mitigate the consequences of incidents, ensuring the safety of both the public and the environment.

The cooling system in a nuclear plant is essential for maintaining optimal temperature levels, but it does not directly influence the reactor's safety systems.


Question: What is head loss testing in nuclear safety?

Wolf Creek performed head loss tests in 2016 to measure the head losses caused by conventional debris (fiber and particulate) and chemical precipitate debris generated and transported to the sump strainers following a LOCA.

The closed test loop included a metal test tank that housed a test strainer submerged in water.


Question: The NRC mandates that all operators must adhere to strict regulations regarding the safety and operation of nuclear reactors.

Nuclear licensing involves a detailed evaluation of the safety documentation that licensees must submit to the NRC prior to the operation of a reactor.

The weather conditions can significantly affect the performance of solar panels, impacting energy production.


Question: MAGNASTOR SYSTEM Criticality Control

3.2 MAGNASTOR SYSTEM Criticality Control for PWR Fuel

1.1 Definitions contain a detailed explanation of phrases used throughout the document.


Question: Both organizations must maintain the independence of their verification and validation capabilities.

Independence of V&V functions is a critical attribute for both organizations.

The document outlines several key organizational attributes that are essential for effective operations.


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

The energy measurements in nuclear spectroscopy are crucial for identifying and quantifying the isotopes present in a sample, as they provide specific peaks corresponding to the energies of gamma rays emitted by the isotopes.

The analysis of the reactor coolant system is vital for ensuring the integrity and safety of a nuclear plant's operations.


Question: The NRC mandates that water chemistry be controlled to ensure the safety of reactor operations...

The ABWR design incorporates a specific water chemistry control program to maintain the boron concentration within acceptable limits under all operational conditions.

The energy release calculations for the containment structure are critical to ensure that safety standards are met during operational transients.


Question: socioeconomic impacts of ISR projects

However and as described in the following sections, from a local perspective, any economic activities that can provide jobs and employees who can purchase local goods and services can have positive impacts on the local standard of living.

At each project area, aquifer restoration will commence as soon as the initial well field is depleted of economic reserves and will be completed within about 2 years after production ends at each location.


Question: The use of Risk Informed Completion Time Program provides a framework for determining the necessary time allowances during safety-related system failures.

Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time Program.

In this situation (loss of automatic initiation capability), the 96 hour or 8 day allowance, as applicable, of Required Action F.2 is not appropriate, and all ADS values must be declared inoperable within 1 hour after discovery of loss of ADS initiation capability in both trip systems.


Question: National Register of Historic Places recommendation

the licensing board determined the recommendation was “a sufficiently significant change since the time the construction permit was issued that it merits present consideration.”

After the licensing board reconsidered that decision and dismissed the contention, the Appeal Board reversed the dismissal, agreeing with the original decision that the recommended NRHP listing was “a significant change in circumstances that … warrant[ed] consideration in the … proceeding.”


Question: The NRC mandates that any usage of byproduct material in sealed sources must detail the manufacturer and model number as documented with regulatory bodies.

An application for a specific license to use byproduct material in the form of a sealed source or in a device that contains the sealed source must either identify the source or device by manufacturer and model number as registered with the Agency, NRC, or with an Agreement State.

The general license in subsection (H) does not authorize the manufacture, assembly, disassembly, repair, or import of products containing radium-226, except that timepieces may be disassembled and repaired.


Question: initial flaw size for crack growth assessment

For repair overlays, the initial flaw size for crack growth in the original weld or base metal will be based on the as-found flaw or postulated flaw, if no pre-overlay examination is performed.

The flaw is at or near the boundary of two different materials, evaluation of flaw growth in both materials is required.


Question: nuclear reactor licensing process

There were two members of the public in attendance for this meeting.


Question: What does irradiation embrittlement (IE) mean in the context of nuclear reactors?

The staff has amended "Aging Effect/Mechanism" column entry for GALL-SLR Report item IV.B2.RP-296 to read as "Loss of material due to wear; loss of fracture toughness due to thermal aging embrittlement (CASS only)."

EPRI wrote the following in relation to the staff’s proposed revision of “Aging Management Program (AMP)ITLAA” column entry for GALL-SLR Item IV.B2.RP-355 in the ISG.


Question: general requirements for Technical Specifications

12 Describe the general requirements associated with Specifications 3.0.1 through 3.0.6 relating to implementation of the Technical Specifications

The Salem ILOT 17-01 SRO NRC Exam has been approved following extensive review and regulatory compliance checks.


Question: What is Strontium-90 used for in nuclear regulatory processes?

Strontium-90 is measured in picocuries per gram (pCi/g) as it is a byproduct of nuclear fission and is important for assessing radiation levels in regulatory filings.

The QC summary table indicates the batches collected and analyzed on specific dates and times, with various parameters related to quality control.


Question: The evaluation of potential internal flooding must adhere to the requirements laid out in the probabilistic risk assessment.

Options include: - Seismic PRA - Seismic Margins Analysis (SMA) - EPRI Tier 1/2/3 method**


Question: containment instrument air compressors

The containment instrument air compressors are normally kept in a standby mode.

The inspection methods that could be used to specifically identify the source of leakage into the containment are those specified by ANSI/ANS-56.8-1994, *Containment System Leakage Testing Requirements*, for Type B and Type C leak-test methods.


Question: The NRC emphasizes the importance of stakeholder engagement in regulatory discussions.

The Commission's Policy Statement on "Enhancing Public Participation in NRC Meetings" outlines strategies to increase stakeholder engagement in regulatory proceedings.

Gerond George submitted the report on February 12, 2025, concerning various operational aspects of the plant.


Question: What are the potential impacts of climate change on operational activities and socioeconomic factors in nuclear energy?

Will expected climatological changes affect the socioeconomic building or operational impact levels assigned in Chapters 3 and 4? Potential changes to the building impact level could occur if there are long-term, persistent impacts from building activities.

Are expected climatological changes likely to alter the impacts of plant activities on local transportation infrastructure?


Question: What is the recurrence rate for mafic eruptions in the BRVF?

The recurrence rate of 5.3x10⁻⁵/yr results from the eruption of 85 mafic events since 1.6 Ma in the BRVF.

The calculated column heights for the 16,000 realizations were binned into the corresponding altitude ranges used in the ASHPLUME2 model, and used to calculate a weighted value for each altitude range (Table 22).


Question: What are the measured forces for the tendons in nuclear reactors?

Table 4 presents a summary of Unit 1 Dome Tendon Forces, including measured force values recorded over several years.

The reactor's cooling system is designed to maintain optimal temperatures and prevent overheating during operations.


Question: how are emergency conditions terminated in nuclear facilities?

Steps will be taken to terminate from the event, either directly or following a transition period (prior to entering a state of recovery operations). Usually, the Unusual Event and Alert classification levels will be directly terminated (no entry into recovery).

This element is not applicable to the licensee emergency plan.


Question: What is the significance of the USGS National Hydrology Dataset in environmental reports?

The USGS National Hydrology Dataset (NHD) is crucial for mapping water bodies and understanding hydrological characteristics in environmental assessments.

The Horizon Center Site is marked with a red dot to indicate its specific location in Figure 1.3-1, which serves as a reference point in the environmental report.


Question: What are the boundaries defined for the scram region in the data tables?

The Slope of ABSP APRM flow-biased trip linear segment is represented by the symbol mTRIP.


Question: hydroelectric power dam

Georgia Power Hydroelectric Power Dam: Nacoochee has a capacity of 4.8 MW and is located in Lakemont, Georgia, approximately 35 mi (56 km) west of Oconee Station.

Baxter Manufacturing is an engineering, design and development firm specializing in injection molding, located in Westminster, South Carolina, approximately 11 mi (18 km) southwest of Oconee.


Question: The NRC conducts inspections to ensure compliance with technical safety regulations related to nuclear facilities.

The inspection was an examination of the activities conducted under your license as they relate to radiation safety and to compliance with the Nuclear Regulatory Commission (NRC) rules and regulations and the conditions of your license.

The inspection findings are as follows: 1. Based on the inspection findings, no violations were identified.


Question: What is Strontium-90 in nuclear analysis?

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


Question: Identify if the pressure in the reactor coolant system is below the specified limit of 300 psig.

The examiners will mark the performance as satisfactory or unsatisfactory based on the observations made during the evaluation.


Question: What is the role of the Ultimate Heat Sink in nuclear power plants?

Division 1 and 2 SW subsystems and UHS shall be OPERABLE.


Question: nuclear regulatory compliance

The Nuclear Regulatory Commission (NRC) is responsible for ensuring the safety and security of nuclear reactors by enforcing compliance with federal regulations.

The advances in renewable energy technologies are impacting the energy landscape and shaping future investments.


Question: What are invalid water level data measurements in nuclear regulatory filings?

Water level data for 802A (168.1 ft msl) measured on 13-Nov-85 considered invalid.

Water level data for Jun-86 to Jan-87 provided in Piezometer Weekly Readings Report for VEGP Units 1 and 2 (Georgia Power January 1987).


Question: What is the minimum detectable concentration in nuclear analysis?

The following Analytical Methods were performed:


Question: what is channel calibration in nuclear safety?

A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel so that it responds within the required range and accuracy to known input. The CHANNEL CALIBRATION shall encompass the entire channel, including the required sensor, alarm, interlock, and trip functions. Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an in-place qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel.

SNC proposes to change TS 1.1 'Use and Application Definitions' definition of ENGINEERED SAFETY FEATURE (ESF) RESPONSE TIME.


Question: What is the iodine partition factor in steam generators?

Primary-to-secondary leak rate (gal/min): 1.0.


Question: County Emergency Operations Plans

Under the provisions of the Wisconsin Statutes 22.16 and the Manitowoc and Kewaunee County Emergency Government Ordinances, authorities of both counties have the responsibility and authority to coordinate offsite emergency activities in the event of a radiological incident.

Selected department personnel receive training in ingestion sampling procedures. May ban the sale of foods which have harmful levels of radioactivity and advise on the use, sale, or disposal of animal feeds containing harmful levels of radioactive contamination.


Question: A contributor to risk that has significance is identified when its Fussell-Vesely importance measure exceeds 0.005.

A basic event (i.e., equipment unavailabilities and HFEs) or risk contributor whose Fussell-Vesely importance measure value is greater than 0.005 or the risk-achievement worth importance measure value is greater than 2.

The state-of-practice is effectively summarized in the definition for 'current good practice.'


Question: The test evaluates the tightness of the RCS and its associated systems under high pressure to ensure there are no leaks during operation.

To verify the integrity and leak tightness of the RCS and the high-pressure portions of associated systems by performing a hydrostatic pressure test in conformance with Section III of the American Society of Mechanical Engineers (ASME) Code.

The required spray flow is obtained during design conditions. (See NSSS Startup Manual, SU-2.1.4.)


Question: In support of regulatory compliance, the licensee is required to analyze the impact of maximum SIT pressure adjustments on key safety margins, such as peak cladding temperature.

In LAR enclosure, attachment 6, sections 3.1.1, “CE16STD Large Break LOCA,” and 3.1.2, “CE16NGF Large Break LOCA,” the licensee states CE16STD and CE16NGF large break LOCA analysis of record (AOR) limiting peak cladding temperature (PCT) cases has been reevaluated for an increase in its modeled maximum SIT pressure to 675 psia.

The documentation must specify the protocols used in assessing fuel performance under normal operating conditions to ensure safety guidelines are met.


Question: What are aging management programs in nuclear plants?

Aging Management Programs (AMPs) are designed to ensure that the effects of aging on systems, structures, and components are understood and adequately managed throughout the period of extended operation of a nuclear power plant.

The reactor coolant system provides a flow of water that absorbs the heat produced in the reactor core and transfers it to the steam generator.


Question: The NRC Facilities Security Branch Chief has a process for notifying individuals about decisions related to their access qualifications.

When the NRC Facilities Security Branch Chief makes a determination to deny or revoke unescorted access for an individual nominated as a reviewing official, the individual shall be provided with a written explanation of the basis for this determination.

The individual may make the appeal either in writing or in person.


Question: The NRC mandates that licensees conduct assessments of isolation systems for containment integrity and safety performance.

4. Fuel Channels (Liquid System) See Note 7 b


Question: loading category in nuclear fuel management

The LOADING CATEGORY defines allowable combinations of maximum total canister decay heat and maximum fuel assembly decay heat for the CY-MPC. They are used to determine operational time limits during LOADING OPERATIONS or TRANSFER OPERATIONS.

A stainless steel container that confines a LACBWR DAMAGED FUEL ASSEMBLY, DAMAGED FUEL ROD or FUEL DEBRIS, while allowing gaseous and liquid media to escape and minimizing the dispersal of gross particulates.


Question: Licensees must ensure that the digital upgrades meet the safety and reliability requirements established by the NRC.

The NRC’s approach to evaluating digital systems is rooted in the principles of defense-in-depth and diversity which are intended to manage the risks associated with digital technology.

The replacement of the steam generator is scheduled for summer 2024 and will be performed during the refueling outage.


Question: The regulatory evaluation for the proposed change includes a detailed assessment of potential hazards and the overall impact on safety systems.

The implementation of a risk-informed approach for addressing Generic Safety Issue 191 involves a comprehensive evaluation of the adequacy of the existing emergency core cooling systems and any enhancements that may be necessary.

The assessment of environmental considerations focuses on the potential impacts on local wildlife and natural resources, emphasizing compliance with environmental protection laws.


Question: What are the requirements for nuclear reactor licensing?

The process of nuclear reactor licensing involves a comprehensive review of the safety, environmental impact, and design specifications by the regulatory authority, ensuring that all safety standards are met before operation can begin.

The thermal efficiency of a power plant is significantly influenced by factors such as temperature differential and heat exchange mechanisms.


Question: reactor coolant system operational requirements

One residual heat removal (RHR) loop shall be OPERABLE and in operation, and either: a. One additional RHR loop shall be OPERABLE; or b. The secondary side water level of at least two steam generators (SGs) shall be greater than or equal to 6% narrow range.

No reactor coolant pump shall be started with one or more RCS cold leg temperatures less than or equal to the COMS arming temperature specified in the PTLR unless the secondary side water temperature of each SG is ≤ 50°F above each of the RCS cold leg temperatures.


Question: What is a Mitigating System Performance Index (MSPI)?

WCNOC-163 Mitigating System Performance Index (MSPI) Basis Document  | 14

Calibration of Converters is a critical component in ensuring accurate measurements within the reactor systems.


Question: What are steady-state model parameter uncertainties in nuclear safety assessments?

Steady-state model parameter uncertainties arise from the assumptions and approximations made during the establishment of a steady-state operational condition, affecting the reliability of calculations in safety assessments.

Transient model parameter uncertainties are crucial for understanding how a reactor behaves under changing conditions, which can significantly impact operational safety.


Question: What is the significance of enrichment levels in nuclear fuel?

Initial Assembly Avg. Enrichment wt % 235U (E) refers to the percentage of Uranium-235 isotope present in the fuel assemblies, which is a critical factor in determining the performance and safety of nuclear reactors.

The cooling system must maintain the temperature of the reactor within safe limits to prevent overheating and ensure efficient operation.


Question: nuclear licensing safety analysis

The safety analysis for the proposed reactor design outlines how the design meets the regulatory criteria set forth by the NRC, including a comprehensive assessment of potential hazards and risks associated with the operation of the reactor.

The report also includes a detailed examination of the economic impacts of the nuclear energy project, discussing potential job creation and the benefits of reliable energy supply.


Question: disposal of nuclear devices

Provide information about devices listed in Section 2 or 6, but no longer in your possession.

- **License Number of Recipient (if transferred to a specific licensee):**  Company Name:  ____________


Question: Section 106 of the National Historic Preservation Act

Pursuant to Section 800.6(b)(1)(iv), you will need to file the final Section 106 agreement document (Agreement), developed in consultation with the Wyoming SHPO and any other consulting parties, and related documentation with the ACHP at the conclusion of the consultation process.

We do not believe our participation in the consultation to resolve adverse effects is needed.


Question: public comment period for environmental impact statements

The public comment period continued for 45 days and ended on April 1, 2024.

The U.S. Environmental Protection Agency (EPA) issued its Notice of Availability on February 16, 2024 (89 FR 12335-TN10059).


Question: The completion of the data quality review confirms that the measurements taken were sufficient to categorize the survey unit appropriately.

The analytical results indicated that all systematic measurements were less than a SOF of one (1) when compared to the OpDCGLB.

No anomalies were observed during the performance or analyses of the survey.


Question: The regulation of airborne emissions from nuclear facilities requires precise calculations of dispersal factors for different weather conditions.

Long-Term Average Chi/Q (sec/m³) Calculations (Depleted and 8-Day Decayed) for Routine Releases


Question: nuclear licensing requirements

The nuclear licensing process involves several regulatory steps to ensure that all safety and technical requirements are met before a facility can operate.

The construction of a cooling tower is critical for the efficient function of the reactor, as it helps to dissipate heat generated during operation.


Question: What is a High Radiation Area (HRA) in nuclear facilities?

Operator refers to Radiological Survey Map and determines that the work area is in a High Radiation Area (HRA) and Contaminated Area (CA).

Evaluator's Note: RP-AA-300, Attachment 2 provides definitions of survey map symbols and abbreviations.


Question: Provide details about the whereabouts of the devices listed in your submittal.

Enter the name of the individual responsible for this device:


Question: steam generator repair program

The major factors in the repair program for the Unit 1 steam generator lower supports are shown on this slide. These include an inspection of the excavation prior to re-welding, control of welding pre-heat, buttering, weld bead sequencing, weld joint sequencing, peening, welding filler metal selection and control, post-heating of completed welds, non-destructive examinations and verification of welding parameters.

A review of the repair operations available for repair of the structures indicated that the in-place repair of the structures was the most advantageous solution.


Question: soil classification in nuclear regulatory documents

SOIL CLASSIFICATION: USCS - SILTY SAND (SM), brown, fine grained, well graded, damp (FILL). - SILTY SAND (CSM), as above with bricks at 31.5'. - SAND (SW), tan, fine to coarse grained, well graded, trace clay, damp to moist.


Question: The aging management of concrete structures involves monitoring for conditions such as cracking, which may arise from various environmental interactions.

Cracking due to expansion from reaction with aggregates

Increase in porosity and permeability; loss of strength due to leaching of calcium hydroxide and carbonation


Question: The NRC assesses the effects of thermal-hydraulic conditions on fuel damage experiments to evaluate metal-water reactions.

Therefore, the NRC concludes that the petitioner did not provide sufficient information to support the revision of RG 1.157 and appendix K to 10 CFR Part 50, to require that the rates of energy release, hydrogen generation, and Zircaloy cladding oxidation from the metal-water reaction of zirconium with steam considered in evaluation models used to calculate ECCS cooling performance be calculated based on data from cited experiments, instead of using the Cathcart-Pawel or Baker-Just equations.

Isothermal tests provide this necessary temperature control.


Question: What is the purpose of the Diesel Generator and HPCI Building in nuclear plants?

The Diesel Generator and High Pressure Coolant Injection (HPCI) Building houses the Units 2 and 3 (swing) emergency diesel generator, HPCI system components and other safe shutdown equipment.

The new diesel generator building, and the Unit 1 HPCI building are not included in the evaluation boundary of the Diesel Generator and HPCI building and are addressed in the Miscellaneous Not In Scope Structures and Switchyard Structures, respectively.


Question: ASME Code implications for nuclear licensing

NRC requirements and long term policy has allowed licensees to use Section XI to address structural integrity versus construction code.


Question: nuclear reactor licensing process

The nuclear reactor licensing process involves several stages, including pre-application consultations, submission of the application, detailed technical reviews by the Nuclear Regulatory Commission, and issuing of the operating license once compliance with safety and environmental standards is verified.

The cooling system is designed to remove heat from the reactor core during normal and emergency conditions, preventing overheating and maintaining safe operational temperatures.


Question: nuclear regulatory examination

The NRC Exam aims to assess the knowledge and proficiency of personnel in nuclear regulatory practices and safety protocols.

The reactor core must be carefully monitored to ensure optimal performance and preemptively address any potential operational issues.


Question: The NRC emphasizes the importance of radiation safety measures in nuclear facilities.

The NRC requires licensees to implement safety measures to ensure the protection of personnel and the public from radiation hazards during reactor operations.

SRB Technologies, Inc. offers a variety of products including exit signs and markers, designed for improved visibility and guidance in emergency situations.


Question: Steam Generator Blowdown valves operation

The SD054 valves are normal Steam Generator Blowdown throttle control valves. An additional function of the Unit 1, (A through H valves) and the Unit 2, (B train valves [B, D, F, & H]) is to isolate Blowdown in the event of a High Energy Line Break (HELB) in the SD system.

Full stroke exercising the valves is a Unit operation concern in that closure of these valves during normal operation presents a thermal transient to the downstream piping and components including the blowdown condenser.


Question: What is the purpose of the Commercial Grade Dedication process in nuclear regulatory procedures?

IP 71111.21.N.03 – Commercial Grade Dedication – 50.69 alternative procurement treatment.

PRA Configuration Control Smart Sample – Performed six times – no violations identified; one minor issue and one observation.


Question: The placement of fuel assemblies within the EOS-37PTH Dry Storage Canister is critical for maintaining safety and compliance with regulatory standards.

Figure 3 illustrates the Peripheral (P) and Inner (I) Fuel Locations for the EOS-37PTH DSC.

The operational efficiency of the coolant system is vital to the overall performance of the reactor, impacting the thermal dynamics significantly.


Question: The NRC mandates that the Emergency Ventilation System must effectively filter potential airborne contaminants before they are released to the atmosphere.

Filtered by ESF Emergency Ventilation System Iodine Adsorber to Atmosphere

Radiological Control is essential for maintaining safety protocols in nuclear facilities.


Question: radiation effects on foam materials

As described in FSAR Section 5.2.3.2, irradiation tests have shown that no change in the resilient characteristics will take place for exposures up to 1.0x10^8 Rads.

The effects of aging on the intended function of the reactor vessel vertical weld V-3-A will be adequately managed through ongoing periodic ASME Code Section XI Inservice Inspections, per 10 CFR 54.21(c)(1)(iii).


Question: The RCIC system must function properly during specified operational modes in order to meet safety standards.

TS 3.5.3 LCO requires the RCIC system to be operable in Mode 1, and Modes 2 and 3 with reactor steam dome pressure greater than 150 psig.

The inspectors presented the emergency action level and emergency plan changes inspection results to Rachael Lorch, Emergency Preparedness Manager, and other members of the licensee staff.


Question: What is the significance of average discharge canal water temperature in nuclear safety evaluations?

The average discharge canal water temperature is monitored to ensure that it stays within regulatory limits, as excessive temperatures can impact aquatic life and violate environmental standards.

The yearly data displayed in various colors indicates the performance of the facility over different years, providing a visual representation of operational stability.


Question: The NRC's guidance on the management and design of heavy loads is critical for ensuring safety in nuclear facilities.

RG 1.244, Revision 0, 'Control of Heavy Loads at Nuclear Facilities', provides criteria for the lifting and handling of heavy loads in order to prevent accidents that could compromise safety.

Temporary Containment Opening Review Guidance outlines the procedures for assessing openings in containment structures, ensuring they meet safety standards during maintenance.


Question: What is the purpose of the Emergency Removal Valve (ERV) in nuclear reactors?

An ERV (Emergency Removal Valve) is designed to prevent overpressure in the reactor by allowing the removal of excess steam and reducing pressure in emergency situations.

The table outlines the torus water temperature at various times, which indicates a consistent rise in temperature, pointing out the need for monitoring in the event of a leak.


Question: What are the qualifications needed for sample preservation when conducting chemical analysis?

Sample received with inadequate chemical preservation. Additional preservation performed by the laboratory.

Target analyte detected in calibration blank at or above acceptance limit. Sample value was > 10x the concentration in the calibration blank.


Question: The analyzed data for the Zr-4 material is essential for ensuring the accuracy of the hydrogen pickup model used in reactor safety assessments.

Tables 34-1 and 34-2 provide the requested data for the Zr-4 data points used for the hydrogen pickup model calibration and validation, respectively.

Appendix C contains markup corrections for each of these items with the updated information.


Question: The safety analysis documents play a crucial role in establishing the fire protection measures at nuclear facilities.

The DNPS Updated Final Safety Analysis Report (UFSAR), which is updated regularly in accordance with the requirements of 10 CFR 50.71(e), provided significant input for system and structure descriptions and functions.

The Maintenance Rule Database documents the results of Maintenance Rule scoping for DNPS systems and structures.


Question: What is the full spectrum LOCA methodology?

The proposed amendment would add the Westinghouse Electric Company LLC (Westinghouse) Topical Report WCAP-16996-P, Rev.1, "Realistic LOCA [loss of coolant] Evaluation Methodology Applied to the Full Spectrum of Break Sizes (full spectrum LOCA methodology)," to the list of approved analytical methods used to determine the core operating limits as listed in Technical Specification 5.6.3, "Core Operating Limits Report (COLR)."

The U.S. Nuclear Regulatory Commission (NRC) staff has determined the need for a regulatory audit to be conducted in accordance with the Office of Nuclear Reactor Regulation, Office Instruction LIC-111, “Regulatory Audits” (ML082900195), for the NRC staff to examine the licensee’s non-docketed information with the intent to gain a better understanding of the license amendment request (LAR), to verify information, and to identify information that would require docketing to support the basis of the NRC staff’s licensing decision.


Question: What happens during a loss of external electrical load in a nuclear power plant?

A major load loss on the plant can result from loss of external electrical load due to some electrical system disturbance. Offsite ac power remains available to operate plant components, such as the reactor coolant pumps; as a result, the onsite emergency diesel generators are not required to function for this event.

It has been determined that the most severe radiological consequences will result from the loss of ac power accident discussed in subsection 15.2.6.


Question: Americium-241 in nuclear analysis

The project involves various stages including environmental assessments and community engagement initiatives.


Question: The concentration of chemicals in the control room must be monitored to ensure safety standards are met.

Worst Case Concentration in C.R. specifies the maximum expected chemical levels in the control room based on the types of chemicals stored and their toxicity limits.

The thorough analysis of compliance with electrical safety standards is essential for the safe operation of nuclear reactor systems.


Question: The new Risk Informed Completion Time (RICT) Program is being introduced for compliance with updated guidelines.

The RICT Program Upgrade Report is associated with the implementation of the Risk Informed Completion Time (RICT) and addresses the regulatory requirements outlined in TSTF 591.

The table summarizes various technical specifications and their disposition, highlighting the differences between TSTF submissions.


Question: What is the Diesel Generator Starting Air System and its function?

The Diesel Generator Starting Air System normal alignment allows air from both air tanks to enter both the left and right starting air headers. When one air tank is isolated, both left and right starting air headers are supplied from the remaining air tank.

The Diesel Generator Lubricating Oil System is designed to supply lubricating oil to all friction surfaces of the diesel engine.


Question: internal coatings aging management program

Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks aging management program is a new condition monitoring program that manages degradation of internal coatings/linings exposed to raw water, treated water, air, and condensation where loss of coating or lining integrity could prevent satisfactory accomplishment of any of the component's, or downstream component's, current licensing basis intended functions.

If the acceptance criteria are not met in any of the expanded samples, an analysis will be conducted to determine the extent of condition and extent of cause.


Question: What is the minimum detectable concentration in radioactive waste analysis?

The result for Strontium-90 is measured at -0.0295 pCi/g with an uncertainty of +/-0.00766.


Question: gamma radiation energy measurements

The cooling system uses water as a coolant to transfer heat away from the reactor core, ensuring safe operating temperatures and preventing overheating.


Question: The NRC has specific requirements for conducting inspections of large-scale radiation devices to ensure compliance with license conditions.

the NRC did not substantiate that CBP staff deliberately caused CBP to be in violation of the license condition requiring annual inspections of large-scale radiation devices.

Requests under the FOIA should be made in accordance with 10 CFR 9.23, Requests for Records.


Question: Operators must comply with rigorous safety standards set forth by the NRC to ensure reactor licensing.

The NRC requires licensees to analyze potential hazards and submit a safety analysis report as part of the licensing process.

The reactor's cooling system is designed to efficiently transfer heat away from the core to prevent overheating.


Question: What are the potential aging effects on BWR reactor vessel internal components?

The NRC staff’s guidance in Table IV.B1 of the GALL Report lists two potentially applicable aging effects (i.e., in addition to cracking) for generic BWR reactor vessel internal components (including BWR core shroud and core shroud repair assembly components) that are made from either stainless steel (including CASS) or nickel alloy: (1) loss of material due to pitting and crevice corrosion (Refer to GALL AMR IV.B1-15), and (2) cumulative fatigue damage (Refer to AMR Item IV.B1-14).

Cumulative fatigue damage analyses for the limiting ASME Section III Class 1 components have been identified as TLAA as discussed in Section 4.3.2.


Question: design basis events in nuclear power plants

The LGS design basis events are unchanged as a result of the PPS upgrade.

The PPS upgrade does not change the minimum criteria for each protective function whose operation may be controlled by manual means initially or subsequent to initiation.


Question: PWSCC susceptibility in Alloy 600

Evaluations of Alloy 600 susceptibility to PWSCC, or analyses in support of repairs to affected locations, can be TLAAs.

Westinghouse performed an assessment of PWSCC susceptibility for Alloy 600 components and Alloy 82/182 welds in DCPP Units 1 and 2.


Question: What is the significance of tritium concentration in nuclear safety assessments?

Tritium Concentration pCi/L is a critical parameter in assessing the environmental impact of nuclear operations, particularly regarding groundwater contamination.

Elevations in Feet are often used in construction and engineering to measure the height of structures or terrain, unrelated to nuclear regulatory processes.


Question: Gamma spectrum analysis in nuclear testing

Gamma spectrum analysis involves measuring the energy spectrum of gamma rays emitted by radioactive materials to identify the isotopes present.

The sample size recorded was 9.242E+02 grams, which is essential for determining the precision of the results obtained from the analysis.


Question: What is the variation of annual minimum daily-mean flow in the Savannah River?

Table 2.4.11-3 (cont.) Variation of Annual Minimum Daily-mean Flow in the Savannah River at Augusta, Jackson, and Burtons Ferry Gages includes data about the annual minimum daily mean flow at specific river miles.

The J. Strom Thurmond Dam is located on the Savannah River and serves multiple purposes including flood control and hydroelectric power production.


Question: The NRC mandates that licensees must maintain safety protocols when handling systems associated with the reactor operations.

The system involves the use of boron as a neutron absorber, which is maintained at a concentration of 1018 ppm during reactor operations to control reactivity levels.

The event description for the steamline break indicates a significant incident that requires immediate attention from the operators under the established safety protocols.


Question: licensing basis events

The PRA being developed is design- and site-specific. It will address all modes and hazards, including seismic events, using the guidance in RG 1.233 and NEI 18-04 as one acceptable method for non-LWR designers to use when selecting licensing basis events (LBEs).

The methodology in NEI 18-04 includes plotting event sequence families on the frequency-consequence (F-C) target and assessing margins based on event frequency and estimated 30-day dose at the EAB.