Question: What is the process for transferring NPDES permits?

As an alternative to transfers under paragraph (a), any NPDES permit may be automatically transferred to a new permittee if: 1. The current permittee notifies the Agency at least 30 days in advance of the proposed transfer date; 2. The new notice includes a written agreement between the existing and new permittees containing a specified date for transfer of permit responsibility, coverage and liability between the existing and new permittees; and 3. The Agency does not notify the existing permittee and the proposed new permittee of its intention to modify or revoke and reissue the permit. If this notice is not received, the transfer is effective on the date specified in the agreement.

All Publicly Owned Treatment Works (POTWs) must provide adequate notice to the Agency of the following: - (a) Any new introduction of pollutants into POTW from an indirect discharger which would be subject to Sections 301 or 306 of the Clean Water Act if it were directly discharging those pollutants; and - (b) Any substantial change in the volume or character of pollutants being introduced into that POTW by a source introducing pollutants into the POTW at the time of issuance of the permit.


Question: ASME Code Case N-740-2 requirements

The preservice examination acceptance standards of IWB-3514 will be met for the weld overlay. In applying the acceptance standards to planar indications, the thickness, t₁ or t₂, defined in Figure 1(b), will be used as the nominal wall thickness in IWB-3514, provided the base material beneath the flaw (i.e., safe end, nozzle, or piping material) is not susceptible to SCC. For susceptible material, t₁ will be used. Planar flaws in the outer 25% of the base metal thickness will meet the design analysis requirements of 2(b).

If the life of the overlay is less than the length of the inspection interval, this requirement will ensure that the overlay is examined prior to the end of the life of the overlay. This requirement is more


Question: aircraft impact design requirements

Design features must be DEMONSTRATION and related functional design criteria defined such that, with limited use of operator actions, one or more physical barriers are maintained to limit the release of radionuclides from reactor systems, waste stores, or other significant inventories of radioactive materials assuming the impact of a COMMERCIAL REGIONAL COMMUTER SIZED large commercial aircraft.

The Federal government must provide for measures to prevent or minimize the likelihood of seizure of a large commercial aircraft as well as appropriately deal with any such seizures.


Question: The inspection plans designed in accordance with ASME Code, Section XI ensure compliance with established safety standards.

The examinations and tests meet the Inspection Plan as required by the ASME Code, Section XI; and the repair/replacement activities and evaluations supporting the completion of B2R22 conform to the requirements of Section XI.

The Byron Generating Station has continuously operated since its commercial service date in 1985.


Question: The NRC specifies minimum cooling times based on the enrichment of uranium in fuel assemblies.

Minimum Cooling Time (years) is a critical factor that is determined by the enrichment wt % 235U (E) of the fuel assembly, as outlined in the loading tables.

The documentation also outlines various safety measures that must be implemented during the construction and operation of the nuclear facility.


Question: The testing protocols must ensure that all safety measures meet the established criteria for plant operations.

1.2 These programs include criteria for determining when testing is required, such as proof tests before installation, pre-operational tests, post-maintenance tests, post-modification tests, in-service tests, and operational tests (such as surveillance tests required by Plant Technical Specifications), to demonstrate that performance of plant systems is in accordance with design.

1.5 The initial start-up test program is planned and scheduled to permit safe fuel loading and start-up; to increase power in safe increments; and to perform major testing at specified power levels.


Question: What is the purpose of a construction permit in nuclear power plants?

The purpose of the construction permit is to ensure that the proposed facility is designed and constructed in accordance with applicable regulations and safety standards.

In the report, the adverse effects of environmental factors on the construction process must be evaluated accurately to ensure compliance with the regulatory requirements.


Question: safety goals for nuclear power plants

The Commission’s policy statement on “Safety Goals for the Operations of Nuclear Power Plants,” issued in 1986 (Volume 51 of the Federal Register [FR], page 30028 [51 FR 30028]; August 21, 1986), presents a risk-informed philosophy for the NRC staff to use in its regulatory reviews of proposed actions that may affect commercial nuclear power reactors.

In February 1981, Executive Order (EO) 12291, “Federal Regulation,” directed executive agencies to prepare a cost-benefit impact analysis for all major rules and stated that cost-benefit actions should be based on adequate information about the need for and consequences of proposed actions.


Question: The U.S. Nuclear Regulatory Commission must evaluate the financial capacity for decommissioning as part of the licensing process.

Pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.75(e)(1)(iii)(C), “Reporting and Recordkeeping for Decommissioning Planning,” and Appendix E to 10 CFR Part 30, “Criteria Relating to Use of Financial Tests and Self-Guarantee for Providing Reasonable Assurance of Funds for Decommissioning by Nonprofit Colleges, Universities, and Hospitals,” the applicant shall demonstrate financial assurance for decommissioning.

The NRC staff expects to complete this review by September 30, 2025.


Question: The redundant reactivity control system is designed to offer an alternative mechanism for shutting down the reactor during scenarios where the reactor protection system fails, while continuously monitoring key reactor parameters.

The RRCS is designed to provide a redundant and diverse method of shutting down the reactor, in the unlikely event that the RPS does not scram the reactor as a result of an anticipated operating transient.

The electric Class 1E refers to systems and components that are important to safety but do not contribute to the control of the reactor during normal operations.


Question: What is the process for submitting a regulatory audit summary report?

Upon completion of the audit, the regulatory audit summary report will be placed into ADAMS within 90 days of the completion of the audit or before the regulatory action that the audit supports is completed, whichever is shorter.

If the NRC staff identifies information during the audit that is needed to support its regulatory decision, the NRC staff will issue requests for additional information to the licensee as soon as possible after the end date of the audit.


Question: The NRC monitors various aqueous samples to evaluate the presence and concentrations of specific chemicals.

Total Alkalinity determined in the sample was recorded at a concentration of 120 mg/L, indicating a measure of the sample's ability to neutralize acids.

The sample was received on 07/19/24 and was analyzed by the laboratory to determine compliance with relevant environmental standards.


Question: The identification of key assumptions is critical to the risk-informed decision-making process.

An assumption is labeled key when it may influence (i.e., have the potential to change) the decision being made. Therefore, a key assumption is identified in the context of an application.

The PRA assumptions and modeling uncertainties that could potentially impact the risk-informed GSI-191 application were identified based on a review of the relevant PRA notebooks.


Question: The analysis of data includes an assessment of minimum detectable concentrations (MDC).

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

The project encompasses environmental restoration at the LaCrosse Boiling Water Reactor (LACBWR) site.


Question: What are the overall transport fractions for debris types in a pressurizer compartment break?

Table 3.e.6-20: Overall Transport Fractions for a Pressurizer Compartment Break (PBN1) includes various debris types and their corresponding transport fractions under operational conditions for Train A and Train B.

The NRC requires extensive documentation and demonstration of compliance with safety regulations to ensure that the reactor design meets all technical specifications.


Question: The program is designed to ensure that components with potential thermal aging are effectively monitored and evaluated to maintain safety standards during extended operation.

The Thermal Aging Embrittlement of Cast Austenitic Stainless Steel (CASS) aging management program is a new condition monitoring program that will provide assurance that RCPB CASS components (i.e., pump casings) with the potential for significant thermal aging embrittlement continue to meet their specified intended functions.

The reactor coolant system components are inspected in accordance with the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Section XI.


Question: Hermes Reactor


Question: What does the Plant Protection System (PPS) entail?

The Plant Protection System (PPS) Performance Specification outlines the requirements necessary for ensuring the effective operation and safety of the nuclear facility's protective measures.

The reactor's thermal efficiency is determined by the temperature difference between the coolant and the core material.


Question: Reactivity control during tests may be done under normal or abnormal conditions leading to significant evaluations according to specific sections of the guidelines.

Reactivity manipulations may be conducted under normal or controlled abnormal conditions (refer to Section D.5.d) but must be significant per Section C.2.a of Appendix D.

For new reactor facility licensees that use the ATC operator primarily for monitoring plant parameters, the chief examiner may place SRO I applicants in either the ATC or BOP position to best evaluate the SRO I in manipulating plant controls.


Question: hydrostratigraphic classification

Hydrostratigraphic Classification for the PSEG Site

Track of PMH Producing Maximum Negative Surge


Question: What is the significance of the Zircaloy-Inconel eutectic reaction in nuclear safety?

Unfortunately, to this day, nearly 40 years later, NRC’s regulations still do not consider the Zircaloy-Inconel eutectic reaction that would, with high probability, occur at temperatures lower than 2200°F.

All current designs of water reactors contain various components made from high-nickel and/or high-iron content alloys.


Question: What challenges are associated with concrete aging in nuclear facilities?

The NRC requires licensees to conduct exercises to ensure readiness in emergency situations.


Question: What are the factors influencing hydrostatic and hydrodynamic loads in flood risk assessment?

Hydrostatic and hydrodynamic loads are critical factors to consider in flood risk assessments, as they impact the structural integrity of facilities during flooding events.

The NRC emphasizes the importance of maintaining effective communication strategies to ensure that all stakeholders are well-informed about safety measures.


Question: sample container authentication

6. Sample container labels present and legible

### Notification of Variance
- **Person Notified:**
- **Contacted By:**
- **Notify Date:**
- **Notify Method:**
- **Notify Comment:**


Question: noble elemental organic aerosol

Detailed model information at time (h) = 0.5000


Question: What is Neptunium-237 and its significance in nuclear regulatory filings?

Neptunium-237 (Np-237) is a synthetic element that is used in nuclear materials and has implications for radiological assessments in nuclear regulatory contexts.

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


Question: What guidelines are established for Anchor/Darling DDG valves regarding motor stall events?

BWROG guidance for Anchor/Darling DDG valves needs to be followed in evaluating stem-disk connection weak link under a motor stall event.

Some licensees have a weak link justification for the stem-disk connection for Anchor/Darling DDG valves that are determined to be non-susceptible, and do not require additional attention in response to the stem-disk connection failure issue.


Question: NRC review standard RS-002

This section describes conformance to the published draft NRC review standard RS-002 (Reference 1). Draft RS-002, Attachment 2 incorporates and clarifies NRC guidance from the Standard Review Plan (SRP). For each applicable SRP section listed, the corresponding SSAR section(s), and a statement of conformance are provided.

The table includes sections on population distribution and evaluations, confirming specific criteria as necessary for the site assessment.


Question: groundwater quality degradation

Xcel Energy finds that impacts from groundwater quality degradation resulting from water withdrawals for the proposed SPEO are SMALL.

Impingement occurs when organisms are held against the intake screen or netting placed within intake canals.


Question: nuclide minimum detectable activity

The nuclide MDA (Minimum Detectable Activity) is a crucial metric that defines the lowest level of activity (in pCi/grams) of a radionuclide that can be reliably detected in a given sample.

The activity of K-40 is measured at 5.23E+00 pCi/grams, with a minimum detectable activity (MDA) of 4.74E-01 pCi/grams.


Question: The NRC is conducting an audit to review the license amendment request for the digital instrumentation and controls at the Limerick site.

The audit will consist of a series of open or partially closed public meetings using video conferencing tools and an online portal (also known as electronic portal, ePortal, electronic reading room, or eDocs web portal).

If you have any questions, please contact me by telephone at 301-415-2871 or by e-mail to michael.marshall@nrc.gov or Bhagwat “BP” Jain by telephone at 301-415-6303 or by e-mail to bhagwat.jain@nrc.gov.


Question: time-limited aging analyses

Each DNPS TLAA has been evaluated for the subsequent period of extended operation. Each evaluation contains the following information: TLAA Description, TLAA Evaluation, and TLAA Disposition.

This evaluation provides the basis for the disposition, which falls into one of the three disposition categories described below.


Question: The average enrichment of uranium-235 in the initial assembly is critical for reactor safety and efficiency.

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

The minimum initial assembly average enrichment is crucial for ensuring optimal fuel performance.


Question: What is the importance of debris margins in nuclear reactor safety evaluations?

The values in Table 2 are similar to the debris limits and margins for a deterministic design basis. These values can be used to perform a prompt operability determination following discovery of an unanalyzed debris source.

The maximum transported quantity of ThermoLag debris for breaks up to 10 inches is slightly greater than the tested quantity due to rounding differences and is reported as no margin.


Question: The table presents various characteristics of different nuclides, including their energy values and yields.

The reactor uses a series of complex safety systems to ensure operational integrity and minimize potential risks to the environment.


Question: The ultimate parent company will be US Athens SpinCo Corp., which will own a significant share of the joint venture related to the ConverDyn operations.

The indirect transfer of control will occur when Honeywell transfers its ownership of HES to NewCo Corp.

After the close of the transaction, US Athens SpinCo Corp. will be a publicly traded Delaware corporation.


Question: How do facility licensees ensure that their simulation models maintain accuracy over time?

Facility licensees that maintain a simulation facility must demonstrate continued assurance of simulator fidelity by conducting performance testing in a manner sufficient to ensure that simulator fidelity has been demonstrated and met (10 CFR 55.46).

If the re-evaluation reveals significant modeling problems with the previous model and the model had been used to negatively train operators, then reactor safety may have been impacted.


Question: What is a significant basic event in nuclear risk assessment?

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.

See the definition for 'current good practice.'


Question: What is the capacity of the Coneross Creek Wastewater Treatment Plant?

The Coneross Creek Wastewater Treatment Plant has a capacity of 7.8 million gallons per day (mgd) (0.3 cubic meters/second [m³/sec]) and an average daily flow of 3 mgd (0.1 m³/sec) (Oconee County 2020-TN967).

The Oconee County also collects taxes and their disbursement for the Keowee Key Fire District and the Oconee County School District.


Question: design-basis hazard event

A. The current design-basis hazard event has mean frequency less than \(10^{-5}/year\), and the mean value of the conditional core damage probability (CCDP) is assessed to be less than \(10^{-1}\).

Table E4-1 provides the external hazards evaluated, identifies the applicable screening criteria, summarizes the evaluation, and provides a disposition for the RICT Program.


Question: nuclear reactor licensing process

The nuclear reactor licensing process requires several detailed submissions to ensure compliance with regulatory safety standards and public health considerations.

The cooling system uses a combination of water and air to maintain optimal temperatures within the reactor core.


Question: What is ISG 06 in the context of nuclear licensing?

ISG 06 "Licensing Process," Revision 2 provides guidance on the regulatory processes related to digital instrumentation and controls in nuclear facilities.

The meeting will include introductions from NRC and NEI and will feature presentations by NEI on relevant topics.


Question: The proposal to extend the activation timeframes for emergency response ensures that more qualified personnel can be deployed during critical situations.

1) Extension of facility activation requirements for the TSC and OSC from 60 minutes to 90 minutes, and for the CERC from 75 minutes to 90 minutes, following declaration of an emergency event classified as an Alert or higher: The proposed changes revise the SPS license by extending augmentation response timeframes for the augmented ERO.

Figure 5.5.c, “Radiation Protection Organization,” deleted figure. Radiological Protection (RP) responders incorporated into TSC, OSC, and CERC figures to better reflect response in each facility.


Question: What is atmospheric stability classification in nuclear safety assessments?

Atmospheric stability classification is essential for understanding the dispersion of airborne pollutants and is categorized into different classes based on factors such as wind speed and direction.

The diagrams depict various environmental parameters aggregated over time, providing insights into general weather patterns but not specifically related to nuclear safety.


Question: Licensees are required to submit a financial assurance plan if their possession limit exceeds the maximum amount prescribed by Section 30.35.

If the license’s possession limit exceeds the quantity for which Section 30.35 prescribes the maximum $1,125,000 fixed amount of financial assurance, the licensee must submit a decommissioning funding plan (DFP) based on the expected actual cost of decommissioning.

The decommissioning process involves various steps, including site assessment, dismantling of structures, and disposal of radioactive waste.


Question: Stakeholder engagement is critical for ensuring compliance with regulatory standards during nuclear reactor developments.

Topics, such as considerations for environmental reviews, streamlining back-end administrative processes, and considerations for how design changes may be addressed.

To see information regarding previously held periodic advanced reactor stakeholder public meetings, please visit the NRC’s public website at https://www.nrc.gov/reactors/new-reactors/advanced/stakeholder-engagement/meetings.html.


Question: Licensees must evaluate their system performance under various loading conditions and power factors to ensure reliability.

The capability curves provide a graphical representation of the performance characteristics of generators at different power factors and kilowatt outputs.

The generator operates at 1800 RPM and requires a hydrogen pressure of 60 PSIG for optimal performance.


Question: reasonably foreseeable harm disclosure

Reasonably Foreseeable Harm. The FOIA Improvement Act of 2016 amended the FOIA as follows (5 USC 552(a)(8)): (A) An agency shall— (i) withhold information under this section only if— (I) the agency reasonably foresees that disclosure would harm an interest protected by an exemption described in subsection (b); or (II) disclosure is prohibited by law; and (ii) consider whether partial disclosure of information is possible whenever the agency determines that a full disclosure of a requested record is not possible; and (II) take reasonable steps necessary to segregate and release nonexempt information.

After the 2003 Columbia shuttle disaster, the National Aeronautics and Space Administration (NASA) created a dissent resolution process with a broad mandate: Any “substantive disagreement with a decision or action that an individual judges is not in the best interests of NASA” can qualify.


Question: regulatory guidance on radiation exposure limits

And so both of those things are being considered in the guidance that we put out, and 1 rem over a month is not a life-threatening dose. But it is a dose at which EPA, NRC, other federal regulators will start to consider relocating those populations. That criteria is 500 millirems a year.

But it was a first step in changing a policy that's been in place for, you know, sixty-some years.


Question: The activity levels measured in microcuries per milliliter are critical for understanding the safety of nuclear materials.

Activity (Units - 10^-14 µCi/ml) measures the radiation emitted from a sample, which is essential for assessing safety in nuclear operations.

The monthly breakdown of activities is recorded to ensure consistent monitoring of reactor performance.


Question: What is the identified leak rate for the Clinton nuclear facility?

Identified leak rate of Clinton: less than 3.5 gpm.

Assumed actuation due to: Inadequate overpressure condition.


Question: What are the requirements for conducting a safety review for a nuclear reactor license?

The NRC requires that applicants submit sufficient information to demonstrate that the proposed activities will not unduly endanger the health and safety of the public.

During initial testing, the temperature of the coolant fluid was measured at various intervals to ensure system efficiency.


Question: The maximum allowable concentration for each radionuclide in the discharge canal has been calculated considering constant discharge rates and the properties of the mixture flowing from the Waste Heat Treatment Facility.

The maximum permissible release rate based on MPC values from the waste disposal system of each radionuclide in the mixture was calculated using Equation 11.2-14.

Groundwater movement is described in Section 2.4.13.


Question: upgrade of instrumentation and control systems

This License Amendment Request proposes changes to the Technical Specifications related to the upgrade of the Instrumentation and Control System for the Armed Forces Radiobiology Research Institute TRIGA Reactor.

The Armed Forces Radiobiology Research Institute performs various research initiatives aimed at enhancing medical responses to nuclear threats.


Question: What is the role of the DWM System in fire protection for nuclear plants?

The DWM System provides makeup water to the ICS during certain fire scenarios.

The license renewal boundary drawings include numerous sheets that detail the physical layout and safety features of the facility.


Question: The NRC mandates inspections of ventilation components to ensure they function effectively during operations.

Heat Exchanger – (CCSW Pump Cubicle Cooler) Fins, which are made of Aluminum Alloy, serve the intended function of heat transfer, particularly in an environment subject to external condensation, where aging effects like loss of material and reduction of heat transfer must be managed through various programs including one-time inspections.

The components in the open-cycle cooling water system are subject to loss of material and require specific aging management programs.


Question: What is a Seismic Margins Analysis in the context of nuclear safety?

The Seismic Margins Analysis (SMA) is one of the options for evaluating a plant's seismic capabilities and ensuring its safety under seismic events.

The Fire PRA is a critical component that assesses the potential risks and safety measures associated with fire hazards in nuclear facilities.


Question: construction permit licensing

The NRC staff is reviewing a construction permit (CP) application submitted by TerraPower, LLC (TerraPower), on behalf of US SFR Owner, LLC, a wholly owned subsidiary of TerraPower, for the proposed Kemmerer Power Station Unit 1.

The application and additional information is also available on the NRC’s public website at [Terrapower Project](https://www.nrc.gov/reactors/new-reactors/advanced/who-were-working-with/applicant-projects/terrapower.html).


Question: The submission includes various updates regarding safety analysis and licensee responsibilities.

The Updated Final Safety Analysis Report (UFSAR) is a comprehensive document that outlines the safety evaluations conducted for the McGuire Nuclear Station and includes critical information essential for regulatory compliance.

The NRC’s evaluation process also considers environmental impacts and public safety measures that are to be upheld during all operations.


Question: The NRC monitors the performance of safety systems to ensure they function correctly during accidents that could cause core damage.

Generic Safety Issue (GSI)-191 was raised by the United States Nuclear Regulatory Commission (NRC) to ensure that post-accident debris blockage will not impede or prevent the operation of the emergency core cooling system (ECCS) or containment spray system (CSS) in recirculation mode at pressurized water reactors (PWRs) during loss of coolant accidents (LOCAs) or other high energy line break (HELB) accidents that would require recirculation.

Acceptance criteria were developed to identify the postulated breaks that contribute to an increase in risk.


Question: What is the significance of the EALF in nuclear reactor safety assessments?

The Equivalent Anti-Lethal Dose Factor (EALF) is a measure used to assess the radiological safety of nuclear systems, allowing for comparisons between different fuel types and configurations.

Figure 12 illustrates the relationship between knorm and the weight percentage of uranium-235 in the fuel matrix.


Question: The instruments for primary containment isolation must be fully functional before implementing the RICT Program at Fermi 2.

One or more required channels inoperable.<br>SSCs are not modeled. The model will be updated to include these SSCs prior to exercising the RICT program for this TS. The PRA Success Criteria will match the Design Success Criteria.

All issues identified below will be addressed and any associated changes will be made, focused-scope peer reviews will be performed on changes that are PRA upgrades as defined in the PRA standard (ASME/ANS RA-Sa-2009, as endorsed by RG 1.200, Revision 2), and any findings will be resolved and reflected in the PRA of record prior to the implementation of the RICT Program.


Question: What is the purpose of the Faulted topical report ANP-10337?

The Faulted topical report (ANP-10337, Reference 8) defines the criteria and method for analyzing the fuel assembly structural response to externally applied dynamic excitations (i.e. operating basis earthquake (OBE), safe shutdown earthquake (SSE), and LOCA), and is explicitly referenced in the GAIA topical report (ANP-10342, Reference 1).

Reference 15 (ARITA topical report) and Reference 14 (AREA topical report) replace BAW-10231P-A, Revision 1, 'COPERNIC Fuel Rod Design Computer Code', for the overheating of fuel pellets SAFDL (see item 8.2.3 of Table 4-1).


Question: The flow of energy through the system is computed as a function of mass flow and enthalpy.

Energy Flow (BTU/sec) is calculated by multiplying the Mass Flow (lbm/sec) with the Average Enthalpy (BTU/lbm).

The chart indicates various measurements of time, mass flow, energy flow, and average enthalpy.


Question: What is the acceptable ECCS performance for LWRs?

the acceptable ECCS performance for LWRs.

CHAIR BALLINGER: Yeah, I think three we've done --


Question: What is the purpose of AMP XI.M29 in managing aluminum tanks?

Aluminum tanks (within the scope of AMP XI.M29, "Outdoor and Large Atmospheric Metallic Storage Tanks") will be used to manage loss of material due to pitting, crevice corrosion.

Not Applicable. There are no stainless steel, nickel alloy underground piping, piping components, or tanks in Steam and Power Conversion. See Subsection 3.4.2.2.3.


Question: What is the process of nondestructive examination (NDE) in nuclear work?

Each MISTRAS location performs a variety of Nondestructive Examination. NDE personnel are qualified as described in Section 14 of this checklist.

Are procedural controls adequately implemented? YES ☑ or NO ☐ (describe in Findings/Deficiencies current section above) or N/A ☐


Question: regulatory inspection reports for nuclear reactors

The Integrated Inspection Report for the Sequoyah Nuclear Plant, including the details of inspections conducted and their findings, provides a comprehensive assessment of compliance with regulatory requirements as of January 28, 2025.

The ADAMS accession number assigned to this document is ML25024A182, indicating it can be found in the NRC document repository for public access.


Question: pressurized thermal shock analysis

Unit 1 Reactor Vessel Values for Analysis of Potential Pressurized Thermal Shock Events provide critical parameters for evaluating the integrity of the reactor vessel under thermal stress conditions during operational transients.

The Steam Generator Quality Assurance Program outlines the procedures and standards for maintaining the reliability and safety of steam generators in nuclear power plants.


Question: The operational requirements and safety measures tailored for the facility are outlined in the updated Technical Specifications.

The Technical Specifications (Appendix A) and the Environmental Protection Plan (Appendix B), as revised through Amendment No. 326, are hereby incorporated in the renewed license.

This license amendment is effective as of its date of issuance and shall be implemented within 90 days from the date of issuance.


Question: boron recycle system

Figure 9-110 illustrates the flow diagram of the Boron Recycle System (NB), which outlines the connections and processes involved in managing boron within the reactor.

The document outlines various safety protocols that ensure the reliability and integrity of the reactor's containment structure during normal and emergency operations.


Question: Nuclear facilities are mandated to assess their groundwater monitoring programs regularly.

The groundwater monitoring program must ensure that any potential release of radioactive materials is detected and reported in compliance with NRC regulations.

Regular inspections are crucial to ensure the structural integrity of the facility and the safety of its operations.


Question: The NRC’s guidance specifies the proper handling of security-related sensitive information to maintain safety and security standards.

NRC's Regulatory Issue Summary (RIS) RIS 2005-31 provides criteria to identify security-related sensitive information and guidance for handling and marking of such documents.

A copy of this letter will be available electronically for public inspection in the NRC Public Document Room or from the NRC’s ADAMS, accessible from the NRC Web site at [NRC ADAMS](https://www.nrc.gov/reading-rm/adams.html).


Question: nuclear licensing terminology

Nuclear licensing refers to the process by which entities obtain permission from regulatory agencies, like the NRC, to build, operate, or modify nuclear facilities and ensure compliance with safety regulations.

The development of modern manufacturing techniques has led to the production of materials with enhanced properties that can meet the demanding specifications of various industries.


Question: The NRC evaluates potential consequences of accidents in nuclear plants before approving licenses.

In evaluating new information concerning severe accident consequences, the NRC considers potential consequences of extreme events that could result in significant radiation exposure to the public.

Xcel Energy has provided a response regarding the requirements for environmental analysis under 10 CFR 51.45(c) in their license renewal application.


Question: nuclide yield measurements

CS-137 has a yield of 85.12%, indicating the percentage of decay events that produce this specific nuclide in a nuclear reaction.

The table displays various measurements for numerous nuclides, each providing detailed characteristics such as energy levels and activity rates in the context of radiation detection.


Question: What constitutes NAC Proprietary Information in regulatory filings?

NAC makes this application for withholding of proprietary information based upon the exemption from disclosure set forth in: the Freedom of Information Act (“FOIA”); 5 USC Sec. 552(b)(4) and the Trade Secrets Act; 18 USC Sec. 1905; and NRC Regulations 10 CFR part 9.17(a)(4), 2.390(a)(4), and 2.390(b)(1) for “trade secrets and commercial financial information obtained from a person, and privileged or confidential” (Exemption 4).

Affiant has reviewed the information described in Item 2 and is personally familiar with the trade secrets and privileged information contained therein, and is authorized to request its withholding.


Question: What is the method for inspecting CEA shroud bolts under aging management?

Non-qualified visual inspection of the CEA shroud bolts for management of material due to wear and potential loss of preload due to ISI/RC no later than two refueling outages from the beginning of the license renewal period.

Comments are akin to those for the UIA fuel alignment plates in Item C10.


Question: The precision of the indication systems for control rods is critical to ensure safe reactor operations.

The RPI System provides an accurate indication of actual control rod position, but at a lower precision than the step counters. This system is based on inductive analog signals from a series of alternating primary and secondary coils spaced along a hollow tube. The normal indication accuracy of the RPI System is ± 6 steps (± 3.75 inches), and the maximum uncertainty is ± 12 steps (± 7.5 inches).

The acceptance criteria for rod position indication is that rod positions must be known with sufficient accuracy in order to verify the core is operating within the given sequence, overlap, design peaking limits, ejected rod worth, and minimum SDM (LCO 3.1.6, "Shutdown Bank Insertion Limits," and LCO 3.1.7, "Control Bank Insertion Limits").


Question: What is the importance of nuclear regulatory filings?

Regulatory filings are essential for ensuring compliance with safety standards and implementing safety measures in nuclear facilities.

McKinley County has a median household income of $40,108, which is significantly lower than the statewide median.


Question: emergency source acceptance light

An EO reports the 'EMERGENCY SOURCE ACCEPTED' light on the Unit 2 ESS ABT is NOT lit.


Question: Could construction activities lead to changes in bat behavior, affecting their feeding and sheltering routines?

Construction or refurbishment and other site activities, including site maintenance and infrastructure repairs, could prompt behavioral changes in bats.

The proposed action would not involve forest conversion or management and would generally not disturb existing forested habitat on the site.


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

The Small Break Loss-of-Coolant Accident (SBLOCA) is a design basis accident that occurs when a small break in the primary coolant system allows water to escape and can affect the cooling of the reactor core.

The Reactor Coolant System (RCS) is designed to transport heat away from the reactor core, ensuring that the temperature remains within safe operational limits.


Question: source term in nuclear safety calculations

Source Term: - 1: 8.250E-01 - c:\program files (x86)\vadrad3.03\defaults\gfr11612.inp - D:\tsum\2020\LCA\BRW LOCA.rft - 0.00000E+00


Question: The regulations mandate that during normal operations, the oxygen levels within the primary containment must remain at or below 4 percent by volume.

Title 10 of the Code of Federal Regulations (10 CFR) Section 50.44, “Combustible gas control for nuclear power reactors,” states that for a plant with an inerted containment atmosphere, the oxygen concentration in the primary containment is required to be maintained below 4 percent by volume during normal plant operation.

During a loss-of-coolant accident, pressure increase in the drywell leads to a series of events where steam and gases are vented into the suppression pool, mitigating the pressure in the drywell.


Question: nuclear probe testing results

The logging mode used during the probe testing was set to '1', which indicates a specific operational state in the data collection process.

The serial number of the probe model used in the tests was 357439, ensuring traceability and compliance with regulatory standards.


Question: The NRC mandates that all licensees submit detailed safety evaluations for significant changes in their operational processes.

The regulatory framework requires that any alterations to a licensed activity, including but not limited to operational changes, must undergo thorough safety evaluations and be submitted to the NRC for review.

Advanced nuclear reactors utilize passive safety systems which rely on natural phenomena to ensure their safety during operational conditions.


Question: confidential commercial information in nuclear licensing

The basis for proposing that the information be withheld is that NuScale treats the information as a trade secret, privileged, or as confidential commercial or financial information.

The information is of a sort customarily held in confidence by NuScale and, to the best of my knowledge and belief, consistently has been held in confidence by NuScale.


Question: nuclear safety affidavit

I declare that the averments of fact set forth in this Affidavit are true and correct to the best of my knowledge, information, and belief.


Question: plume exposure emergency planning zone

...which includes participation by each State and local agency within the plume exposure EPZ that have not been tested in a previous exercise.

An exercise scenario, simulating a radiological release to the environment, was developed and conducted using procedure NMP-EP-303 (Reference 3).


Question: automatic pressurizer PORV actuation

The automatic PORV opening logic is satisfied when two-out-of-four (2/4) pressurizer pressure channels exceed their setpoint.

Automatic actuation logic and actuation relays consist of the same features and operate in the same manner as described for Function 1.b, except that the LCO is not applicable in MODE 4 as discussed below for Function 9.b.


Question: Nuclear reactors must be designed to withstand various environmental conditions and operational stresses to ensure safety.

The design of the reactor must account for potential natural and man-made hazards to protect the integrity of the plant.

The financial impact of regulatory changes can significantly affect the operational budget of a facility, leading to potential delays in project timelines.


Question: What measures are taken to mitigate noise impacts during construction for nuclear projects?

As a result of the remote location of the project areas and the low population density of the surrounding area, noise impacts are expected to be SMALL.

NuFuels will implement a hearing conservation program to ensure that proper PPE is worn and engineering controls are in place to protect workers from potentially damaging noise.


Question: The NRC mandates that when license conditions are not satisfied, corresponding corrective actions must be taken as outlined in the procedure.

This example represents a more complicated use of logical connectors. Required Actions A.1, A.2, and A.3 are alternative choices, only one of which must be performed as indicated by the use of the logical connector "OR" and the left justified placement.

The chemical composition of materials used in reactor construction is critical for ensuring structural integrity and safety during operation.


Question: What analytical methods are used for isotopes like Am241 and Pu238 in nuclear safety reports?

**Analytical Method:** DOE EML HASL-300, Am-05-RC Modified

The following samples were analyzed using the above methods and analytical procedure(s).


Question: nuclear decay levels

The decay level (pCi/M²) represents the radioactivity concentration measured for a specific nuclide, indicating the presence of nuclear materials in terms of picocuries per square meter.

K-40 is a naturally occurring isotope of potassium that is found in various concentrations in soil and other environmental materials.


Question: The NRC evaluates alternatives to the relicensing of the Oconee Station and assesses the environmental impacts of each option.

As part of its environmental review, the NRC is required to consider alternatives to SLR and evaluate the environmental impacts associated with each alternative.

Duke Energy’s environmental report was supplemented with additional data from various governmental consultations.


Question: nuclear licensing regulations

Room in accordance with Title 10 of the Code of Federal Regulations 2.390, “Public Inspections, Exemptions, Requests for Withholding.”

Sincerely, Signed by Blamey, Alan on 02/12/25


Question: What is the significance of safety-related systems in nuclear reactors?

Safety-related systems in a nuclear facility are designed to control and safely shut down the reactor in the event of an emergency, ensuring the protection of public health and safety.

The dimensions of components used in the construction of reactors must adhere to stringent standards to ensure structural integrity.


Question: energy peak analysis


Question: minimum detectable concentration

MDC refers to Minimum Detectable Concentration, which is the lowest concentration of a substance that can be reliably distinguished from the background noise level.

The report date for this analysis is August 15, 2019, and the sample was collected under Client Sample ID: L1-SUB-TNK-AJGS-001-SB.


Question: What is the significance of deposition lambda in nuclear safety assessments?

- Deposition Lambda (1 / Hours) - Noble | Elemental Organic | Aerosol 0.00000+00 | 0.00000+00 | 6.00000+00

Natural deposition - Powers' Model, Compartment 2 - Deposition Net DF - Noble | Elemental Organic | Aerosol 1.00000+00 | 1.00000E+00 | 1.00000E+00


Question: Strontium-90 measurement in soil samples

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


Question: activity concentration measurement

Activity Concentration is reported in units of pCi/g Dry, representing the amount of radioactivity per gram of dry sample.

The uncertainty associated with the measurements can significantly impact the interpretation of results, requiring careful analysis for regulatory compliance.


Question: The NRC evaluates the potential consequences of not renewing the operating licenses for Oconee Station which would lead to the cessation of operations and decommissioning.

The majority of nuclear power plant operations activities would cease with reactor shutdown.

Some activities (e.g., security and oversight of spent nuclear fuel) would remain unchanged, whereas others (e.g., waste management, administrative work, laboratory analysis, surveillance, monitoring, and maintenance) would continue at reduced or altered levels.


Question: What is the significance of a 6-mile radius in nuclear regulatory documents?

The 6-mile radius around the nuclear facility is critical for emergency planning and response, as it helps define the areas that may be directly affected by radiological releases during potential accidents.

The Mississipi River State Wild and Scenic Recreational District offers various activities for visitors, including fishing, hiking, and canoeing, which contribute to the ecological health of the area.


Question: risk-informed categorization for nuclear reactors

Risk-informed categorization is a methodology that allows for a more efficient prioritization of structures, systems, and components based on their safety significance and risk profile within a nuclear power plant context.

The availability of documents related to regulatory history can support stakeholders in understanding changes to safety regulations over time.


Question: What is the procedure for testing Core Spray A pump?

When you take the shift, perform Core Spray A pump testing per ST-3PA, starting at step 8.7.6.

The plant is operating at approximately 100% power.


Question: spent fuel pool cooling system

The Spent Fuel Pool Cooling System is designed to maintain the temperature of the spent fuel pool by removing heat generated from the spent fuel assemblies, ensuring safe storage conditions.

The Flexible Hoses are designed to act as a leakage boundary in the plant, preventing unwanted fluid loss and maintaining system integrity.


Question: peak cladding temperature in nuclear reactors

The 2200°F (1204°C) peak cladding temperature (PCT) criterion was selected on the basis of Hobson’s slow-ring-compression tests that were performed at 25-150°C. Samples oxidized at 2400°F (1315°C) were far more brittle than samples oxidized at <2200°F (<1204°C) in spite of comparable level of total oxidation.

It is significant that “The History of LOCA Embrittlement Criteria” states that “the 17%-ECR and 1204°C [PCT] criteria [of 10 C.F.R. § 50.46(b)] were primarily based on the results of post-quench ductility tests conducted by Hobson.”


Question: license renewal environmental review

The NRC’s license renewal environmental regulations in Part 51 are based in large part on the analyses in its GEIS for license renewal, which summarizes the findings of a systematic inquiry into the potential environmental consequences of license renewal.

Accordingly, the NRC promulgated a further requirement.


Question: aging management programs in nuclear licensing

The TLAA designation in the Aging Management Programs column indicates that cumulative fatigue damage due to fatigue for this component is evaluated in Section 4.3.

Consistent with NUREG-2191 item for component, material, environment, and aging effect.


Question: The NRC provides access to licensing documents through its online system known as ADAMS.

In accordance with 10 CFR 2.390 of the NRC's “Agency Rules of Practice,” a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the NRC’s document system (ADAMS).


Question: dedicated outside air supply fans in nuclear plants

DEDICATED OUTSIDE AIR SUPPLY FANS – 2 IN NUMBER - OPERATE ON HIGH/LOW/DISABLED SELECTOR SWITCH.

BULK FLOW CONTROL VALVE – LOCATED IN RETURN LINE.


Question: radiological baseline studies

Comprehensive radiological baseline studies were conducted at the Crownpoint and Unit 1 project areas in 1978 and 1980. The studies included determining the background dose from cosmic and terrestrial radiation via gamma scanning across the project areas.

Cleanup of the Quivira Mines was included in the Tronox Settlement in 2014.


Question: The systems responsible for regulating reactivity must ensure safe shutdown during both normal operations and foreseeable accidents.

A minimum of two reactivity control systems or means shall provide: 1. A means of inserting negative reactivity at a sufficient rate and amount to assure, with appropriate margin for malfunctions, that the specified acceptable system radionuclide release design limits for the fission product barriers are not exceeded and safe shutdown is achieved and maintained during normal operation, including anticipated operational occurrences.

Information incorporated into SFR-DC 26 per RG 1.232.


Question: financial criteria for nuclear licensing

The NRC staff reviewed CBR’s Crow Butte surety estimate in accordance with requirements in 10 CFR Part 40, Appendix A, Criterion 9, “Financial Criteria” and in License Condition 9.5 of Materials License SUA-1534.

CBR estimated the cost to decommission and decontaminate the Crow Butte facility by an independent party at $51,383,364. This constitutes a decrease of $389,366 below the previous...


Question: What is the COBRA-FLX Critical Heat Flux correlation?

The COBRA-FLX: ORFEO-HMP Critical Heat Flux Correlation provides a methodology for calculating the critical heat flux in nuclear reactor conditions, facilitating the evaluation of safety margins in reactor design.

Table B-1 outlines several test assembly conditions that are crucial for understanding the performance of fuel assemblies under various operating scenarios.


Question: What are the requirements for volumetric examinations according to ASME Code Case N-729-1?

The visual and volumetric/surface examinations will use acceptance criteria set forth in ASME Code Case N-729-1 subject to the conditions in 10CFR Part 50.55a.

Inspections will be controlled by site specific procedures. Engineering evaluations are performed in accordance with the station Corrective Action Program.


Question: What is required for routine inspections of nuclear licensees?

Routine inspections of licensees include important information such as Licensee Name, License Number, Program Codes, Priority (IMC 2800), Last Inspection Date or License Issuance Date (if initial inspection), Date Due, Date Performed, Amount of Time Overdue, and Date Inspection Findings Issued.

14 candidates for Reciprocity Inspections (Priority 1, 2, & 3) were recorded for the period between Feb. 6, 2021 - Dec. 31, 2021, but no inspections were completed.


Question: What are Lead Test Assemblies (LTAs) in nuclear licensing?

This amendment request proposes to add a License Condition to Appendix D, "Additional Conditions," of the Vogtle Unit 1 and Unit 2 Operating License that authorizes use of four Lead Test Assemblies (LTAs) to be placed in limiting core locations.

According to 10 CFR 50.92, "Issuance of Amendment," paragraph (c), a proposed amendment to an operating license involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not:


Question: What constitutes NRC licensable material?

Are any samples NRC licensable material?

Yes - Wet ice was present in the shipment container(s).


Question: The NRC evaluates whether changes to technical specifications support the safety of nuclear reactors.

NRC Regulatory Guide 1.200, "Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 3, December 2020 (ADAMS Accession No. ML20238B871).

NRC Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 3, dated January 2018 (ADAMS Accession No. ML17317A256).


Question: What is the Auxiliary Feedwater system?

In addition, given the worst case failure, the AFW is designed to supply sufficient makeup water to replace SG inventory loss as the RCS is cooled to less than 350°F at which point the Residual Heat Removal System may be placed into operation.

At MPS 3, only two of the three available steam supplies are required to establish an OPERABLE steam supply system.


Question: What is the function of the sump strainer in nuclear containment systems?

The strainer modules (Tophats) are made of stainless steel perforated plate for straining debris from the water.

The types of insulation used in the Containment are reflective or mass insulation as listed on Table 6-140.


Question: nuclear material safety regulations

This document outlines the regulatory framework established by the NRC to ensure the safety and control of nuclear materials.

The office emphasizes the importance of public transparency and community engagement in all decision-making processes.


Question: gamma spectrum analysis

Peak Analysis Performed on: 1/15/2019 8:14:23 AM


Question: risk-informed categorization and treatment of structures

The proposed LARs would adopt Technical Specifications Task Force (TSTF) Traveler TSTF-505, Revision 2, 'Provide Risk-informed Extended Completion Times, RITSTF [Risk-informed Technical Specifications Task Force] Initiative 4b' and adopt the provisions of Title 10 of the Code of Federal Regulations (10 CFR), section 50.69, 'Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors.'

The NRC staff will conduct the audit virtually via Teams using a licensee-established electronic portal available to NRC staff from the date of this letter through March 31, 2024, with formal audit meetings to be scheduled during this period, as needed.


Question: high drywell temperature

Knowledge of the interrelations between HIGH DRYWELL TEMPERATURE and the following: Drywell spray: Mark-I & II

The importance rating of the knowledge and ability necessary for this examination is 3.7.


Question: average enrichment of Uranium-235 in PWR fuel

The table lists the Initial Assembly Avg. Enrichment wt % ²³⁵U (E) for WE 14x14 Fuel, which is critical for understanding the fuel's performance in a Pressurized Water Reactor (PWR).

The loading table also includes burnup rates for various classifications of fuel assemblies, which describe the amount of energy extracted from the fuel during its operation.


Question: Liftoff forces in nuclear safety analysis

The unit of measure for the force is in kilopounds per square inch, abbreviated as kIP, which is commonly used in engineering contexts.


Question: heat transfer mechanism differences between Zircaloy and stainless steel

that the heat transfer correlations derived from stainless steel clad heater rods are suitable for use with Zircaloy clad fuel rods

Run 9573 was one of the four tests conducted with Zircaloy cladding in the PWR FLECHT test program; the assembly used in run 9573 incurred autocalytic (runaway) oxidation.


Question: annual storage loss in reservoirs

The annual storage loss is calculated based on the initial and final storage capacities of the reservoirs over a specified timespan, indicating the amount of water lost each year due to evaporation, seepage, or other factors.

The Gilboa Dam is located in Schoharie County, New York, and supports various recreational activities for the community.


Question: What is the minimum initial assembly average enrichment of uranium-235?

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

The NRC requires a review of the proposed technical specifications during the licensing process.


Question: What is the Level 1 Probabilistic Risk Assessment (PRA) in the context of SAMA analysis?

As part of its review of the Perry Plant SAMA analysis, NRC staff evaluates the applicant’s treatment of internal events and calculation of core damage frequency in the Level 1 PRA model.

Additionally, Table G1.3-2 shows that the seismic PRA peer review and the PRA modeling of offsite power recovery were performed to the ASME/American Nuclear Society (ANS) PRA Standard RA-Sb-2013.


Question: The NRC mandates that licensees must conduct evaluations of safety systems...

The Nuclear Regulatory Commission (NRC) has specific guidelines for the licensing process of nuclear reactors that ensure safety and compliance with regulatory standards.

TerraPower, LLC has proposed a new approach to integrate advanced monitoring systems in its operations.


Question: procurement document requirements

The procurement documents specify the GEH requirements for the Supplier’s reporting of nonconformances, including invoking the requirements of 10 CFR Part 21 if applicable.

The program is applied to all phases of procurement and may require verification of activities of suppliers below the first tier.


Question: The NRC mandates that licensees monitor contamination levels to ensure decontamination effectiveness.

Frequent surveys of the decontaminated areas were essential to tracking the progress made to reduce contamination and to identify recontamination sources.

Problems in establishing a rigorous database included: heterogeneity of containment building surfaces and coatings, irregular geometry of equipment surfaces, and heterogeneity of the contamination pathway and resulting deposition patterns.


Question: The Surry Power Station's plan to classify the turbine building as resistant to tornadoes reflects a commitment to enhanced safety measures.

NRC approval is requested to designate the SPS Units 1 and 2 Turbine Buildings as tornado-resistant structures in the SPS Updated Final Safety Analysis Report (UFSAR) under a different methodology and acceptance criteria than those defined for the other SPS tornado resistant (i.e., Tornado Protection "T") structures.

The LAR has been reviewed and approved by the SPS Facility Safety Review Committee.


Question: The monitoring of environmental conditions is essential for assessing the impact of effluent releases.

The station's Meteorological Monitoring System provides the capability for predicting atmospheric effluent transport and diffusion.

Installed in the Control Room are the necessary instrumentation readouts to assess station status under all conditions.


Question: nuclear reactor licensing process

The licensing process for a nuclear reactor involves multiple stages, including site approval, construction permits, and operational licensing, each regulated by the NRC to ensure safety and compliance with federal standards.

The cooling system of a nuclear reactor must effectively remove heat to prevent overheating and maintain safe operational conditions during all phases of reactor operation.


Question: The Makeup Water System is essential for supplying demineralized water required for various safety functions and normal reactor operations.

The MWS supplies demineralized makeup water necessary for normal reactor coolant services, secondary system makeup, firewater, and miscellaneous plant uses including domestic and drinking water.

The MWS includes the safety-related condensate storage tanks (CSTs).


Question: The NRC oversees the continuous improvement of regulatory processes in response to changing safety standards.

The NRC is responsible for ensuring that all regulatory requirements are met regarding nuclear facility operations and safety protocols.

OI adopted new software and investigative tools necessary to help the office successfully address evolving challenges in a post pandemic setting.


Question: What are the key components of a nuclear reactor's cooling system?

The primary cooling system at the Watts Bar Nuclear Plant consists of a closed loop system that circulates coolant through the reactor core to remove heat generated during nuclear fission.

In the context of administrative processes, the NRC exam focuses on key regulatory measures that ensure the safe operation of nuclear facilities, including training and qualification of plant personnel.


Question: The NRC mandates that all licensees must follow safety protocols during operations to protect public health and safety.

The regulation of nuclear materials and the licensing of nuclear power plants are crucial to ensuring the safety and security of the nuclear industry.

The power level during reactor operations is monitored continuously, and any fluctuations must be addressed by the control room staff.


Question: nuclear safety analysis

Nuclear safety analysis involves a systematic evaluation of potential risks and hazards associated with the operation of a nuclear reactor, focusing on prevention and mitigation strategies to protect public health and safety.

The cooling system of a reactor operates by transferring heat away from the reactor core to prevent overheating and maintain safe operational conditions.


Question: What is the significance of peak analysis in gamma spectrum analysis?

Peak Analysis Performed on: 5/2/2019 5:07:25 PM


Question: what is NPCH in nuclear safety analysis?

NPCH Reference Elevation (2) is 1972.07 ft.

Suction line losses @ 4,760 gpm is 3.945 ft.


Question: onset of nucleate boiling

The onset of nucleate boiling (ONB) is the critical point at which small vapor bubbles begin to form on heated surfaces in the reactor coolant channel, influencing the heat transfer characteristics significantly.

The axial flux distribution within the reactor core primarily depends on the position of the fuel assemblies and the design of the reactor itself.


Question: aging management evaluations

Consistent with NUREG-2191. The Structures Monitoring (B.2.1.33) program will be used to manage loss of sealing of elastomers in expansion joints, roofing, seals, gaskets, and moisture barriers (including caulking, flashing and other sealants) and supports associated with emergency diesel generator, HVAC system components, and other miscellaneous mechanical equipment (including support members, welds, bolted connections, and support anchorage to building structure) exposed to air - indoor uncontrolled, air - outdoor, and groundwater/soil in the Component Supports and Structural Commodity Group.

Not Applicable. There are no sliding support bearings or surfaces for cable trays, conduit, HVAC ducts, tube track, instrument tubing, non-ASME piping and components, emergency diesel generator, HVAC system components, and other miscellaneous mechanical equipment exposed to air - indoor uncontrolled or air - outdoor in the Structures and Component Supports Grouping.


Question: What are the implications of using alternative source terms in nuclear safety analyses?

Revision 8 adds the GAIA source term as well as changes the ECCS leakage pathway files to correctly model the transport of iodine contained in the ECCS sump.

Independent review complete, all comments resolved.


Question: What is the E-Filing rule for NRC adjudicatory proceedings?

All documents filed in NRC adjudicatory proceedings, including a request for hearing and petition for leave to intervene (petition), any motion or other document filed in the proceeding prior to the submission of a request for hearing or petition to intervene, and documents filed by interested governmental entities that request to participate under 10 CFR 2.315(c), must be filed in accordance with the NRC’s E-Filing rule (72 FR 49139; August 28, 2007, as amended at 77 FR 46562; August 3, 2012).

To comply with the procedural requirements of E-Filing, at least 10 days prior to the filing deadline, the participant should contact the Office of the Secretary by e-mail at hearing.docket@nrc.gov, or by telephone at 301-415-1677, to (1) request a digital identification (ID) certificate, which allows the participant (or its counsel or representative) to digitally sign submissions and access the E-Filing system for any proceeding in which it is participating; and (2) advise the Secretary that the participant will be submitting a petition or other adjudicatory document (even in instances in which the participant, or its counsel or representative, already holds an NRC-issued digital ID certificate).


Question: What is a primary containment air lock in nuclear facilities?

The primary containment air lock is a safety feature designed to prevent the release of radioactive materials from the containment structure, ensuring that the conditions within the reactor remain stable and secure.

To ensure the safety of the overall reactor system, it is important to monitor and control the temperature of the coolant used throughout the heat exchange processes.


Question: What are the aging effects requiring management for the Equipment Access Lock components?

The following aging effects associated with the Equipment Access Lock components require management: Cracking; Cracking; loss of bond; loss of material (spalling, scaling); Increase in porosity and permeability; cracking; loss of material (spalling, scaling); Increase in porosity and permeability; loss of strength; Loss of material (spalling, scaling) and cracking; Loss of material; Loss of preload.

The materials of construction for the Vacuum Pipe Building components are: Concrete.


Question: What is the chemical effects evaluation process for nuclear plants?

Figure 3.o.2-1 highlights the Wolf Creek chemical effects evaluation process using the flow chart in Figure 1 of the March 2008 guidance (Reference 50).

The applicable items in Figure 3.o.2-1 are blocks 2 through 5, 7 through 10, 12, 13, and 15 through 17.


Question: nuclide identification report

The Identification Report provides a comprehensive list of nuclides along with their respective identifiers, energy levels, yields, and activity measurements.

The activity uncertainty is calculated based on statistical methods and reflects the confidence in the measurement values.


Question: What are the critical conditions during a core offload operation in a nuclear plant?

The following Annunciators alarm: 128-A, SFP LEVEL HI/LO; 184-B, SFP 0-RM-102/103 RAD HI; AUO reports Spent Fuel Pool Level is 747’ 6” on Ultrasonic level indicator and slowly LOWERING.

The VAPORSTREAM is NOT available due to IT patch update.


Question: What is the standard process for referencing Technical Evaluation Reports in a Safety Evaluation?

The NRC evaluates the quality of Technical Evaluation Reports (TERs) to determine their sufficiency for inclusion as references in Safety Evaluations (SEs).

The Division of Operating Reactor Licensing oversees various licensing projects across different reactor types to ensure compliance with established safety regulations.


Question: uranium compound material specifications

Uranium Compound Material Specifications


Question: The candidate should be familiar with the criteria for initiating reactor trips and safety injection mechanisms.

State the setpoints, coincidence, blocks and permissives for all Reactor Trips and Safety Injections actuations.

Technical References include documents like S2.OP-AB.NIS-0001 and S2.OP-SO.RPS-0001.


Question: All devices must comply with the stipulations under 10 CFR Part 31, which governs generally licensed devices.

Common generally licensed devices include self-luminous exit signs and fill level and flow gauges.

You are receiving this notification, which includes the attached table(s) of applicable sections of the U.S. Nuclear Regulatory Commission's (NRC's) regulations, because you possess one or more generally licensed devices authorized under Title 10 of the Code of Federal Regulations (10 CFR Part 31).


Question: loss of feedwater flow

The initiating event being modeled is a loss of feedwater flow.

EC-116269 provides the NRELAP5 analysis supporting final safety analysis report (FSAR) Section 15.2.7 for the US460 standard design approval application (SDAA).


Question: 316L stainless steel applications in nuclear engineering

316L stainless steel is essential to U.S. Naval applications from ship parts to weapon systems.

Additive manufacturing (AM), the stepwise construction of a part layer by layer, is used extensively with 316L and shows promise for use in the Navy.


Question: compressive strength in nuclear construction

E1, General Provisions, states that compressive strength is the lowest limit state of flexural buckling, torsional buckling, and flexural-torsional buckling. Q1.5.1.3 of N690-1994 (R2004) addresses the same items.

The provisions in D5, Pin-Connected Members, have been updated compared to the provisions in Q1.5.1.1 and Q1.14.5.


Question: What is a criticality analysis checklist in nuclear regulatory filings?

The criticality analysis checklist is completed by the applicant prior to submittal to the NRC. It provides a useful guide to the applicant to ensure that all the applicable subject areas are addressed in the application, or to provide justification/identification of alternative approaches.

The checklist also assists the NRC reviewer in identifying areas of the analysis that conform or do not conform to the guidance in NEI 12-16.


Question: What is the HPCI system and its role in reactor safety?

The High Pressure Coolant Injection (HPCI) system is a safety system designed to provide coolant to the reactor core in the event of a loss of coolant accident, ensuring that the core remains adequately cooled during an emergency situation.

Reports RPV water level has stabilized at 30 inches and controlled by DFWLC.


Question: proprietary information

The information in this Document is considered proprietary for the reasons set forth in paragraph 6(c), 6(d) and 6(e) above.

Framatome policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.


Question: The NRC mandates that certain inspections are conducted to guarantee compliance with design specifications.

NRC construction inspection reports are essential to ensure that the reactor is built in accordance with approved designs and safety standards.

The Federal Register documents provide a historical account of the regulatory changes affecting various industries.


Question: What is the function of the MDAFW Pump in a nuclear plant?

The Main Design Auxiliary Feedwater (MDAFW) Pump is critical for providing cooling water to the steam generators during emergency situations, ensuring that the reactor remains safe and operational under adverse conditions.

When directed by Lead Examiner proceed to Event 6.


Question: Will expected climatic changes have an impact on the environmental effects from the generation and disposal of nonradioactive and mixed wastes?

NR-1 Will expected climatological changes affect environmental impacts resulting from the generation and disposal of nonradioactive and mixed wastes from the operating facility?


Question: nuclide activity measurement

Nuclide confidence index threshold = 0.30.


Question: What are the effluent characteristics that must be reported for nuclear power plants?

The following effluent characteristics should be updated to match the permit fact sheet: - Temperature, Water Deg. Fahrenheit 4 / Report should be moved from the Daily Maximum column to the Monthly Average Column - Temperature, Water Deg. Fahrenheit 5 / The Daily Maximum should be 93 F and the Monthly Average should be 90 F. - For DSN0011 only, Temp Diff. Between Up/Down Stream Deg F 6 / The Daily Maximum should be Report and the Monthly Average should be 10 F.

The permit will be issued effective July 1, 2018.


Question: stream channel diversions

The potential for stream channel diversions and their potential effects on the safety-related SSC at the new plant location are discussed in this subsection.

The river channel is 2.5 mi. wide at the PSEG Site.


Question: The responsibilities of the Emergency Preparedness Department are critical for ensuring the safety of plant operations.

The Emergency Preparedness Department is responsible for formulating and implementing the emergency response plan, including training and drills for emergency scenarios.

Waterford 3 SES includes various administrative procedures for testing sirens, which are vital components of alert systems in the facility.


Question: What is reciprocity in the context of nuclear licensing?

Performing licensed activities in other jurisdictions is possible through reciprocal recognition of specific licenses (i.e., reciprocity). Agreement States have reciprocity provisions that permit NRC licensees to perform licensed activities under circumstances when an Agreement State is the regulatory authority.

A current list of Agreement States (including names, addresses, and telephone numbers of responsible officials) is available at the NMSS public Web site, https://scp.nrc.gov.


Question: Gland Seal Exhauster trip instrumentation

The main turbine GSE trip instrumentation initiates a trip of the GSE breakers following events in which radiation in the main steam lines exceeds a predetermined value.

Loss of Voltage and Degraded Voltage - two out of three logic.


Question: The NRC emphasizes the need for assessing the total core mass flow during operational scenarios.

The core net mass flow is measured in pounds per second and is critical for understanding reactor performance during various operational conditions.

Measurements taken during the operational phase must comply with federal safety standards to ensure compliance with regulatory frameworks.


Question: reactor elevation profiles

From Reactor 3 to Reference Point (Lat. 38°-07'-12", Long. 77°-04'-20") Elevation (ft) 100| 200 | 300 | 400 |       |       |

The neighboring community's engagement in safety discussions is crucial for maintaining transparency and public trust.


Question: INTERMEDIATE RANGE MONITOR (IRM) SYSTEM

Ability to predict and/or monitor changes in parameters associated with operating the INTERMEDIATE RANGE MONITOR (IRM) SYSTEM controls including: Detector position

A plant startup is in progress with the following: - The Reactor is critical. - The Reactor Mode Switch is in START & HOT STBY. - All IRMs are fully inserted and indicating mid-scale on Range 1.


Question: The surveillance requirements for systems such as the Control Room Emergency Ventilation System are critical in ensuring operational safety during emergencies.

The surveillance requirements for the Essential Service Water System (ESW) include routine checks and maintenance to ensure system availability and reliability during operation.

The final report must include an analysis of the environmental impact of the proposed facility on local wildlife and vegetation.


Question: total transient population within 80 km radii

In order to determine the total transient population, the following categories of transient population were estimated: Seasonal Population, Transient Business Population, Hotel/Motel Population, Special Population (Schools, Hospitals, Nursing Homes, and Correctional Facilities).

The area between 40 and 60 km (25 and 37 mi) of the site is the most densely populated, with a population of 267,376 and an average population density of 110 per mi² (U.S. Census Bureau, 2001).


Question: The NRC governs the procedures for reviewing applications related to obtaining a nuclear license.

Processing an Application for License involves a thorough review of information submitted by the applicant to ensure it meets the regulatory requirements.

The importance of accurate record-keeping is vital in various operational aspects of the facility, particularly when managing radioactive materials.


Question: What are the acceptance criteria for plenum cover assembly inspections?

40 – 60 Methods and Conditions: One time physical measurement from the top of the PC rib pads and support ring to the top of the RPV seating surface, followed by periodic VT-3 visual inspections thereafter. The relevant condition for the physical measurements is confirmation that the differential height between the top of the PC rib sections and support ring and the RPV seating surface shall average less than a 0.004 inch difference from the height in the as-built condition:

EPRI MRP is assuming that the physical measurements of the five operating B&W units were completed during the first renewed license operating period.


Question: What is the function of the Auxiliary Decay Heat Removal System in nuclear safety?

The Auxiliary Decay Heat Removal (ADHR) System can be used to remove residual heat from the spent fuel pool and reactor cavity during outages.

The component types that require aging management review are indicated in the below Table 2.3.3-23, Reactor Building Closed Cooling Water System.


Question: What is gamma spectroscopy in the context of nuclear safety assessments?

The ground/homogenized samples were then divided into aliquots for gamma spectroscopy, isotopic thorium and isotopic uranium.

All duplicate result(s) were within acceptance limits, unless otherwise noted.


Question: The Nuclear Service Water System must be able to support the continued operation of safety systems during various emergency scenarios.

The Nuclear Service Water System is designed to withstand a safe shutdown earthquake and to prevent any single failure from curtailing normal station operation or limiting the ability of the engineered safety features to perform their functions.

Blackout causes startup of all emergency diesel generators and system alignment as described under Engineered Safety Features Actuation for a containment high pressure signal with the exception that the Auxiliary Building branch of the non-essential header is not isolated.


Question: The steps outlined in AMS-TR-0720R2-A guide the calibration checks for transmitter drift monitoring.

Calculation of OLM limits must be based on combining uncertainties of components of each instrument channel from the transmitter in the field to the OLM data storage.

The NRC requires that operators conduct regular safety drills to ensure readiness in emergency situations.


Question: control rod scram procedures

As the SRO, the instruction was given to insert the control rod to position 00, with emphasis on not recommending raising Drive Water Pressure above 370 psid due to issues with the CRD mechanism.

The EO reported that the 1-0305-117 SCRAM Solenoid fuse was blown and was replaced with a like for like, stating no other issues had been found.


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


Question: The process for obtaining a license to operate a nuclear facility involves detailed safety assessments and compliance with regulatory standards.

The Nuclear Regulatory Commission (NRC) evaluates license applications for construction and operation of nuclear power plants to ensure compliance with safety and regulatory requirements.

Dockets provide crucial information about the status of various nuclear power plants that are undergoing licensing processes and modifications.


Question: What materials are used for the CO2 Storage, Fire Protection/Purge System?

The materials of construction for the CO2 Storage, Fire Protection/Purge System components are: - Aluminum - Copper alloy - Copper alloy (>15% Zn or >8% Al) - Elastomer - Glass - High-strength steel - Nickel Alloy - Stainless steel - Steel

The CO2 Storage, Fire Protection/Purge System components are exposed to the following environments: - Air - indoor uncontrolled - Air, condensation - Any - Condensation - Gas


Question: What is required for ground support during excavation of nuclear construction?

To ensure worker safety during excavation and facilitate construction activities, the initial ground support system is expected to be composed of the following components, installed prior to and during excavation: rock bolts to secure unstable rock blocks, as required; wire mesh and a non-structural sprayed-gunite lining to stabilize and protect exposed rock walls; horizontal gravity drains to manage groundwater, as required; and pressurized grout to seal any notable areas of water entry, as required.

Any person may request an exemption permitting the conduct of activities prior to the issuance of a construction permit prohibited by § 50.10.


Question: What is a Level 2 PRA model review?

Documentation associated with the Level 2 and 3 PRA model reviews.

Please provide additional information as to why a new nuclear SMR alternative was determined not to be feasible for detailed analysis.


Question: The analysis of time versus concentration plots is essential to understanding the stability of the reactor's operation under various conditions.

Time vs. concentration plots are utilized to indicate the stability of the reactor conditions over the operational period.

The facility will perform regular inspections to ensure that all systems are functioning within acceptable parameters.


Question: What are the aging effects on elastomer seals exposed to raw water?

Hardening and loss of strength due to elastomer degradation; loss of material due to erosion.

Not applicable. There are no concrete or cementitious material piping or piping components exposed to raw water in the Auxiliary Systems.


Question: zone of influence in nuclear safety

In some cases, if the ZOI for a particular material is very large (i.e., the material has a low destruction pressure or is located on a large pipe), it may extend beyond robust barriers located near the break.

Using the ZOIs listed in the Response to 3.b.1, the breaks selected in the Response to 3.a.1, and the size distributions provided in the Response to 3.c.1, the quantities of generated debris for each break case were calculated for each type of debris.


Question: nuclear licensing application review process

The license application review process for 10 CFR 50.69 involves evaluating the safety and effectiveness of the proposed reactor design and operational plans in accordance with regulatory standards.

The workshop will address various aspects of risk assessment and provide insights into the evolving regulatory landscape in nuclear energy.


Question: The concentrations of radionuclides found in various soil types must be closely monitored.

H-3 has a residential soil concentration of 228 pCi/g and an industrial/commercial concentration of 423 pCi/g.

Regulatory compliance requires adherence to strict guidelines for waste management and disposal.


Question: The NRC stipulates that licensees must evaluate control rod withdrawal dynamics when operating within specific power levels.

Figure 15-278 displays the uncontrolled RCCA bank withdrawal from 10% power, highlighting the relation between reactivity insertion rate and thermal performance.

The high flux trip serves as a critical safety mechanism designed to protect against excessive thermal power output in nuclear reactors.


Question: technical specifications in nuclear licensing

The NRC does not expect operators to memorize the TS, nor does it endorse operating the plant from memory. However, the NRC does expect operators to recognize TS entry conditions, immediate actions, and (in the case of senior operators) bases when presented in a multiple-choice format on the written examination.

If, for example, the systematic/random outline for Tier 2 ends up with 7 items under Category K1 and only 1 item under Category K4, you can balance the coverage by randomly deselecting one of the items in Category K1 and then randomly selecting a replacement item for the same system from Category K4.


Question: Sodium Hydroxide in nuclear facilities

Sodium Hydroxide is listed with the CAS No. 1310-73-2 in the chemicals stored at GGNS Unit 1.

The unit also stores a variety of gases, including Oxygen and Nitrogen, which are critical for various processes.


Question: fog statistics for nuclear site assessment

According to U.S. observing practice, ground fog is a fog that hides less than 60 percent of the sky, and does not extend to the base of any clouds that may lie above it.

During winter months, cold air passing over the relatively warmer water surface of Clinton Lake can become saturated with respect to water vapor.


Question: Technical Specifications in nuclear licensing

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

This license amendment is effective as of its date of issuance and shall be implemented within 90 days from the date of issuance.


Question: The MNGP site is now over 50 years old, qualifying it for evaluation under the National Register of Historic Places.

Therefore, we recommend that the MNGP be subject to intensive level survey and evaluation to determine eligibility for listing in the NRHP.

We recommend that, at a minimum, an archaeological survey be completed for areas within the MNGP Site which may be subject to ground disturbance as part of the relicensing.


Question: What is the purpose of drinking water certification in nuclear facilities?

Drinking water certification ensures that the water used in nuclear facilities meets all safety, health, and regulatory requirements.

Maryland is known for its picturesque landscapes and vibrant cultural heritage.


Question: gamma spectrum analysis

Gamma spectrum analysis is performed using a 500ml Marinelli sample, which is identified by sample number L-1-010-104-AJGS-011-SS and weighted at 759.8 grams.

The facility Dairyland_NPP is equipped with various operational systems to ensure compliance with safety regulations and operational efficiency.


Question: The NRC evaluates the effectiveness of corrosion protection through measurements like base number.

CPM reserve alkalinity, and the corresponding degree of protection that it provides against corrosion is defined by a base number (also referred to as a neutralization number) determined using the procedure specified in Subsection IWL Table IWL-2525-1.

As can be seen in the above table and on Figure 5, base number test results exhibit a relatively high degree of scatter which is typical of tests performed on samples collected at other plants.


Question: gamma radiation measurement

The Gamma Result column indicates the detected gamma radiation levels from the measurements.


Question: The NRC establishes specific limits for the identification and quantification of radioisotopes in various nuclear materials.

PB-212 is a nuclide part of an undetermined solution and has a weight mean activity of 3.60E-02 pCi/grams.

Errors quoted at 2.000sigma can result in statistical uncertainties that affect calculations in nuclear safety analysis.


Question: What is channel calibration in nuclear safety systems?

CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to a measured parameter within the necessary range and accuracy.

The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.


Question: What is HEC-RAS and its significance in hydraulic modeling?

U.S. Army Corps of Engineers (USACE), Hydrologic Engineering Center River Analysis System (HEC-RAS) Computer Software, Version 4.1.0, Davis, California, January 2010.

The study analyzed wave patterns and their transformations along the coastline, focusing on the impact of human activities on local marine ecosystems.


Question: The NRC mandates that licensees must conduct a thorough evaluation of criticality safety for all transportation packages containing radioactive materials.

The process of validating nuclear criticality safety calculational methodology is essential to ensure the safe transportation of radioactive materials.

The document outlines guidelines for the design and operation of molten salt reactors, focusing on enhancing safety through innovative material use.


Question: L&C modifications

The NRC staff plans to modify both L&C #18 and 19, to make the language more general.

The NRC will share the changes with Framatome.


Question: What is Revision 2 of RG 1.200 and its importance in risk-informed licensing?

NEI understands that NRC does not intend to sunset Revision 2 of RG 1.200 with issuance of Revision 3. This is important to licensees, as many have conducted PRA technical acceptability evaluations for ongoing and upcoming risk-informed licensing applications using Revision 2, and there is little value in repeating these evaluations given that the same standard is endorsed in both revisions.

The NRC staff notes and agrees with the comment on Revision 2 of RG 1.200.


Question: The sensitivity of monitoring for radionuclides is influenced by factors such as measurement techniques and environmental conditions.

Minimum detectable dose to man that can be determined by a radiological monitoring program is a function of the measurement sensitivity and natural background variations.

Data gathered in accordance with the above outlined program are compiled and maintained on file by the NFCS.


Question: The time taken for the reactor protection system to respond after a monitored parameter exceeds its setpoint is critical for safety.

The **RPS RESPONSE TIME** shall be that time interval from when the monitored parameter exceeds its RPS trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for transmitters in the Online Monitoring Program provided that the methodology for verification has been previously reviewed and approved by the NRC.

The PTLR is the unit specific document that provides the reactor vessel pressure and temperature limits, including heatup and cooldown rates, for the current reactor vessel fluence period. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 5.6.7.


Question: What is AADT in traffic studies?

AADT, or Average Annual Daily Traffic, refers to the estimated average number of vehicles that pass a specific point on a roadway over the course of a year.

The percent of trucks on U.S. 491 in 2023 is reported as 13%, indicating a significant amount of commercial traffic on this route.


Question: The applicant uses various unverified methods to determine acceptable levels of exposure to radiation.

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: acute exposure scenario

In PAs for both sites, direct exposure to drill cuttings accounts for most of the acute dose (INL: 86%, SRS H-Area Tank Farm (HTF): 95%).

However, from discussions with well drillers, drill cuttings remain in the vicinity of the well head during drilling and are spread out to some unknown area by the owner or landscape contractor after the driller leaves the site.


Question: gamma spectrum analysis

The Gamma Spectrum Analysis utilizes the 500ml Marinelli geometry to detect gamma radiation from the sample with identification energy tolerance of 1.000 keV and includes peak analysis performed between channels 100 and 4096.

The sample identification for the Marinelli geometry was noted as B1-010-001-CIGS-BH02-SB, with a total sample weight of 755.1 grams taken from the Dairyland Nuclear Power Plant.


Question: The NRC's guidelines for evaluating the impacts related to reactor system failures are crucial for maintaining safety.

Regulatory Guide 1.24 focuses on the assumptions used for evaluating the potential consequences of a Pressurized Water Reactor radioactive gas storage tank failure.

Regulatory Guide 1.29 outlines the seismic design classification requirements for nuclear power plants.


Question: What is an Integrated Assessment in the context of flood safety analysis?

An Integrated Assessment will be performed.

PMF combined with maximum wind effects are bounded.


Question: What is the function of the Residual Heat Removal (RHR) system?

Knowledge of system purpose and/or function.

Correct. IAW S2.OP-AB.RHR-0001 (Loss Of RHR) CAS, going to S2.OP-AB.LOCA-0001 (Shutdown LOCA) is correct since the Unit is in MODE 4 when the RCS Inventory was lost.


Question: The identified nuclides assist in understanding the radioactive elements present in samples.

Energy Tolerance: 1.000 keV
Nuclide confidence index threshold = 0.30
Errors quoted at 2.00 sigma.


Question: National Historic Preservation Act compliance

The Board concluded that the Staff had fulfilled its NHPA obligations with respect to identification of historic properties.

Consolidated Intervenors have requested that we set a briefing schedule for any issues that we accept for review.


Question: The phenomena identified through sensitivity studies are essential for boundary conditions in evaluating the model's development.

The highly-ranked phenomena identified in the combined PIRT and shown in Table 2-5 may be divided into two groups dependent on whether additional experimental data are required or not.

The remaining highly-ranked phenomena are input for consideration of available data from vintage experimental data sets and the design of the TerraPower IET and SETs that are tasks within Element 2.


Question: The components that prevent loss of structural integrity in reactors include various elements like the PZR Safe Ends and PZR Shell, which maintain pressure boundaries.

The PZR Safe Ends are components designed to serve as a pressure boundary and are made of stainless steel, intended to withstand the reactor coolant environment, where cumulative fatigue damage may occur.

The aging management program for the PZR seismic lug includes boric acid corrosion assessment as part of routine maintenance.


Question: What is the role of the Reactor Projects Branch in nuclear regulatory safety?

Elba M. Sanchez Santiago, Chief  Reactor Projects Branch 2  Division of Operating Reactor Safety  is responsible for overseeing regulatory activities related to the safe operation of nuclear reactors.

This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at NRC Reading Room and at the NRC Public Document Room in accordance with Title 10 of the Code of Federal Regulations 2.390, “Public Inspections, Exemptions, Requests for Withholding.”


Question: What is the purpose of a subsequent license renewal for nuclear plants?

The FEIS includes the NRC staff’s site-specific evaluation of the environmental impacts of the subsequent license renewal (SLR) for Oconee Station and alternatives to the SLR.

In accordance with Title 36 of the Code of Federal Regulations Section 800.8(c), the NRC staff elected to use the National Environmental Policy Act of 1969, as amended, process to comply with its obligations under Section 106 of the National Historic Preservation Act of 1966, as amended.


Question: Primary Containment Isolation System

The Primary Containment Isolation System is designed to prevent leakage and maintain the integrity of the containment structure under abnormal conditions.

The table lists different aging management programs relevant to various components and materials used within reactor systems.


Question: The report indicates that the nuclide library is crucial for accurate measurements of radionuclide activities.

- **Nuclide Library Used**: C:\Canberra\Apex\Root\Dairyland_NPPLibrary\HOTLAB.NLB


Question: What are the requirements for loading damaged fuel assemblies in a 61BTH Type 2 DSC?

When loading up to 4 damaged or 4 failed assemblies, these must be placed in corner 'A' locations, and the remaining locations 'B' and 'C' shall be loaded with intact fuel.

The maximum lattice average initial enrichments of assemblies is limited to the 'Remaining Four Intact Assemblies' or 'Remaining Four Damaged Assemblies' column of Table 12.


Question: The NRC mandates the evaluation of temperature limits within high-pressure core injection systems for safety assurance.

The conditions specified must be documented per trip and validated with appropriate surveillance.


Question: What is total alkalinity in water analysis?

Total Alkalinity                      | Calculation     | 1        | 572    | H         | mg/L   | 2        | 20       | 08/17/24 0:00 | jck     |


Question: For accurate safety assessments, it is crucial to know the activity levels of various nuclides present in the reactor core.

The activity of K-40 is measured at 10.67 pCi/grams, indicating its presence and potential impact in nuclear processes.

The reactor core must be regularly inspected to ensure that all components are functioning within safe operational limits.


Question: What is the significance of using creep specimens for assessing uncertainties in commercial fuel applications?

Creep specimens are usually much better characterized (in terms of both operation and fabrication) than commercial fuel rods, therefore, having a much lower uncertainty than creep data from the latter.

Please provide the distributions for each cladding type in terms of P-M and ratio of P/M excluding any closed gap data (similar to Figures 5-8, 5-11, 5-14, and 5-17 of the methodology document) and include any new data comparisons requested (identifying this data by different colors). Identify those rods with a positive stress (strain). Please provide the one sided upper and lower 95/95 bounds on these distribution plots.


Question: The NRC evaluates the effects of uncertainties in plant parameters during licensing reviews to ensure safety margins are adequately addressed.

In plant-specific reviews, the uncertainty treatment for plant parameters will be considered acceptable if plant parameters are [ ] as appropriate.

If second-cycle fuel is identified in a plant-specific analysis, whose [ ] the NRC staff reviewing the plant-specific analysis may request technical justification or quantitative assessment, demonstrating that [ ].


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

Activity (pCi/grams) refers to the amount of radioactive decay of a nuclide per unit mass, which is important for assessing radiological safety in nuclear regulatory filings.

The nuclides listed in the table include isotopes with varying yields and energies.