Question: gamma spectrum analysis

The gamma spectrum analysis was performed using a 500 ml Marinelli sample identified as L1-010-107-AJGS-006C-SS, with the peak analysis executed on 6/11/2019 at 10:22:32 AM.

The sample size was recorded as 7.919E+02 grams and the sample was taken on 6/5/2019 at 10:00:00 AM.


Question: The licensing process for nuclear facilities involves detailed safety analyses and technical evaluations.

The Nuclear Regulatory Commission (NRC) mandates comprehensive safety assessments as part of the nuclear licensing process to ensure public and environmental protection.

The plant operates under specific thermal efficiency protocols to maximize energy output and minimize operational costs.


Question: The NRC specifies the allowable initial enrichment levels for various basket types in the 61BTH Type 2 Dry Storage Container.

Maximum Lattice Average Initial Enrichment and Minimum B-10 Areal Density for the 61BTH Type 2 DSC (Failed and Damaged Fuel)

Borated Aluminum/MMC Boral® is used as a minimum density material for certain configurations.


Question: operability determination

The inspectors evaluated the licensee’s justifications and actions associated with the following operability determinations and functionality assessments:

The inspectors evaluated 1R34 activities from September 20, 2024, to January 16, 2025.


Question: The functional design of the PCRVICS specifies that isolation valves are necessary for systems that interface directly with primary containment.

The PCRVICS consists of seven functions implemented using eight logical isolation groups. These functions and groups are largely divided by the interfacing systems which are isolated by actuation of the PCRVICS (e.g., Group IA provides main steam isolation, Group IIA provides isolation of the RHR system, etc.).

Pipes or vents that penetrate primary containment and communicate directly with the reactor vessel have two isolation valves: one inside primary containment (i.e., inboard) and one outside primary containment (i.e., outboard).


Question: Section 106 of NHPA compliance requirements

In accordance with the NHPA of 1966 (16 U.S.C. § 470-470w-6), federally funded, licensed, or permitted undertakings that are subject to the Section 106 review process must determine effects to significant historic properties.

The Eastern Shawnee Tribe is committed to protecting sites important to Tribal Heritage, Culture and Religion.


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

K-40 has a yield of 10.67% and an activity of 1.54E+00 pCi/grams.

Errors quoted at 2.000eigm are significant for determining the precision of our measurements.


Question: The criteria for assessing load combinations on components during emergency conditions include differential pressure and dead weight.

The applicable load combinations for the strainers are: (3) Emergency/Faulted DP + DW + WD + SSE.

Flood loads consist of the effects due to an earthquake in a submerged condition (sloshing and added mass). There are no hydrostatic pressure loads associated with flooding since the flood waters are present on all sides.


Question: What is the process for obtaining nuclear reactor licensing?

The licensing process for nuclear reactors involves a comprehensive review of safety procedures, engineering designs, environmental impact assessments, and regulatory compliance to ensure public health and safety.

The benefits of solar energy include reduced utility bills, decreased carbon footprint, and sustainability, which are critical in the transition to renewable energy sources.


Question: fuel fabrication licensing

Fuel fabrication (10 CFR Part 70) is a crucial part of the licensing process for advanced reactors, ensuring that facilities meet safety and regulatory standards during the construction and operation phases.

Material Control & Accounting (MC&A) (10 CFR Part 74) involves the implementation of procedures and practices to manage the inventory and use of nuclear materials, vital for maintaining safety and security standards.


Question: The NRC has established guidelines under 10 CFR 50.55a concerning reactor coolant systems.

10 CFR 50.55a(c) Reactor Coolant Pressure Boundary, ASME Section III, Code Case 1361-2 Parameter

The ASME Code is essential for maintaining safety standards across various engineering disciplines.


Question: Nickel-59 regulatory measurements

Rad Gamma Spec | Batch | | | | | | | | | | |


Question: The analysis of nuclides involves determining whether they are accepted or rejected based on specific criteria.

?  = nuclide is part of an undetermined solution


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

The safety analysis report (SAR) contains the information needed to demonstrate that the proposed facility will operate safely and in accordance with regulatory requirements.

The methodologies used in estimating the potential market impacts of the proposed changes include differential pricing models and elasticity of demand analysis.


Question: What are the acceptance criteria for analytical methods in nuclear regulations?

All sample data provided in this report met the acceptance criteria specified in the analytical methods and procedures for initial calibration, continuing calibration, instrument controls and process controls where applicable, with the following exceptions.

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


Question: What is the purpose of a regulatory audit in nuclear reactor reviews?

The NRC has determined that a regulatory audit would assist in the review of the TRISO Fuel Design Methodology Topical Report.

By letter EVR_LTR_240176 dated July 31, 2024, Westinghouse Electric Company, LLC requested a review of the TRISO Fuel Design Methodology Topical Report related to the eVinci™ microreactor.


Question: The NRC mandates that operators undergo rigorous requalification programs to maintain their licenses.

This scenario involves the validation of the simulator examination, ensuring that all systems operate correctly under various conditions and the required procedures are followed during the assessment of operator performance.

The examination includes a summary of various operational programs, including the initial license and other training aspects.


Question: nuclear reactor regulatory safety

In order to maintain nuclear safety, licensees must adhere to strict regulatory guidelines established by the NRC during their operation of registered reactors.

The effective date for the regulations listed in Chapter 2 is marked as 09 OCT 2015, which indicates the most recent updates to the documentation.


Question: What is SUNSI and its role in regulatory filings?

Deadline for submission of contentions whose development depends upon access to SUNSI.

If access granted: issuance of presiding officer or other designated officer decision on motion for protective order for access to sensitive information (including schedule for providing access and submission of contentions) or decision reversing a final adverse determination by the NRC staff.


Question: Under what conditions can an owner or operator be excluded from liability for hazardous substances?

As used in this subdivision (a)(4), "owner or operator" does not include a person who establishes, by a preponderance of the evidence, that they acquired the title to the hazardous substance site after the disposal or placement of the hazardous substance.

The person must have undertaken, at the time of acquisition, all appropriate inquiry into the previous ownership and uses of the property consistent with good commercial or customary practice in an effort to minimize liability.


Question: The unit operations shift crew requires advisory technical support in areas such as reactor engineering and thermal hydraulics for safe operations.

An individual shall provide advisory technical support to the unit operations shift crew in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit.

The operations manager or assistant operations manager shall hold an SRO license.


Question: The licensee must adhere to all regulations regarding the management and control of radioactive sources as per COMAR.

The licensee shall comply with all appropriate provisions of COMAR 26.12.01.01 "Regulations for the Control of Ionizing Radiation", and shall possess a copy of these regulations.

The licensee may use Calichek or Lineator device for doing linearity tests of his dose calibrator provided he follows the procedures for the respective device.


Question: The economic activities related to the TVA facility will notably affect property values and local business opportunities in Jackson County.

The location of a TVA facility at Bellefonte will have both a direct and indirect impact on the local economy of Jackson County.

Tentative plans call for railroad access to the site via approximately three miles of new roadbed from the Southern Railway main line at a point about one mile west of Hollywood entering the site from the southwest.


Question: What are risk-informed completion times in nuclear licensing?

The proposed amendment would modify the ANO-2 technical specification (TS) requirements to permit the use of risk-informed completion times (RICTs) in accordance with TSs Task Force (TSTF) Traveler TSTF-505, "Provide Risk-informed Extended Completion Times, RITSTF Initiative 4b," dated July 2, 2018 (ML181834493).

An audit is conducted to gain understanding, to verify information, and to identify information that will be necessary to support the basis of a licensing or regulatory decision.


Question: The various discharge sections of the auxiliary feedwater system may experience failures.

Train C discharge sections fail randomly

The check valve is an essential part of the system that prevents backflow of fluids.


Question: nuclear regulatory compliance requirements

The licensing process for nuclear facilities entails a comprehensive review of safety protocols, environmental impacts, and regulatory compliance requirements established by the Nuclear Regulatory Commission (NRC).

The reactor was tested under various operational conditions to ensure it operates within the designed thermal limits.


Question: What are the types of loading patterns for PWR fuel?

Loading patterns are referred to in the FSAR as follows: A – Uniform Loading Pattern, B – Preferential Three-Zone Loading Pattern, C – Preferential Four-Zone Loading Pattern (with Reduced Cool Times), D – Uniform PMTC Loading Pattern, E – Loading Pattern - X, F – Loading Pattern - Y, G – Loading Pattern - Z, H – Loading Pattern – Z-Prime.

The table lists various storage locations for fuel loads in a PWR reactor, along with their corresponding values for different loading scenarios.


Question: The NRC mandates that inspections of radioactive gaseous and liquid effluent treatment occur every three years to ensure compliance with safety standards.

Radioactive Gaseous and Liquid Effluent Treatment should be conducted every three years to assess the safety and effectiveness of the treatment measures in place.

Inspection results show that the performance indicators have met or exceeded the required standards for the year 2024.


Question: What is neutron fluence and how is it measured in reactor vessels?

Where: - f_{1T} = neutron fluence (neutrons/cm^2) at the outside surface of the reactor vessel (1T) - f_{0T} = neutron fluence at the inner surface of the reactor vessel (0T) = 1.69E+18 neutrons/cm^2

A plant specific calculation was performed to determine the gamma dose rates in the CBS concrete.


Question: Tritium distillation process in nuclear licensing

Product: LSC, Tritium Distillation, Solid

The samples in this SDG were analyzed on an "as received" basis.


Question: Ground Motion Response Spectrum

The horizontal Ground Motion Response Spectrum (GMRS) is developed from the horizontal soil URHS using the approach described in ASCE/SEI Standard 43-05 (Reference 2.5.2-4) and RG 1.208.

Figures 2.5.2-46 through 2.5.2-52 present seismic hazard curves for the seven spectral frequencies at which ground motion equations are available for the GMRS elevation.


Question: What are the results for Americium-241 and Americium-243 in a radiological assessment?

The report details the contact information for Mr. Jason Q. Spaide at LaCrosse Solutions, including his address and project related to the LaCBWR Site Restoration Project.


Question: requirements for transporting nuclear materials

NPC package marking shall be in accordance with 10 CFR §71.85(c) and Subpart D of 49 CFR 172. Package labeling shall be in accordance with Subpart E of 49 CFR 172. Packaging placarding shall be in accordance with Subpart F of 49 CFR 172.

Using a special lifting sling, attach a lifting eye to each of the four threaded holes located near the corners of the OCA lid.


Question: radiological consequences of a liquid radwaste tank failure

Table 15.7.2-2 provides the radiological consequences of a liquid radwaste tank failure, detailing doses in rem for the thyroid and whole body gamma exposure in various areas.

The setup for the radiation monitoring system includes multiple detector types that ensure comprehensive coverage across the facility, enhancing the safety protocols in place.


Question: nuclear reactor performance metrics

The review of the reactor performance metrics includes critical parameters such as thermal efficiency, fuel utilization, and coolant flow rates which are essential for determining compliance with operational safety standards.

The record of the public meeting indicates that several stakeholders expressed their concerns about environmental impacts and safety regulations, highlighting the importance of transparent communication between the NRC and the community.


Question: The NRC mandates that licensees modify their physical security strategies to meet updated DBT standards.

modified to require that the physical security plans, safeguards contingency plans, and the guard training and qualification plans required by 10 C.F.R. §§ 50.34(c), 50.34(d), and 73.55(b)(4)(ii) be revised to provide protection against this revised DBT.

the Commission will consider requests to relax or rescind, either in whole or in part, the requirements of the Order issued on February 25, 2002, imposing ICMs.


Question: containment pressure in nuclear safety assessments

Containment Pressure                            | Minimum                  | Consensus (Minimum)                       | Same as Base Case

pH                                            | Minimum or Maximum       | Maximum for release, Minimum for solubility | Same as Base Case


Question: lead-210 allowable concentration

The allowable concentration limit (ACL) for Pb-210 is set at 62 pCi/L.

Precision bar values may exceed the scale.


Question: total average annual daily traffic counts

The Total Average Annual Daily Traffic Counts on US Highway 20/North Ridge Road provide important data for assessing traffic patterns and impacts in the area.

The Office of Nuclear Regulatory Research conducts independent studies to support the regulatory framework for nuclear safety.


Question: What is an Integrated Safety Analysis (ISA) update?

Enclosed with this letter, in digital format, is the annual update of the ISA Summary for Framatome’s Richland, Washington nuclear fuel fabrication facility.

Framatome requests that these digital files in their entirety be withheld from public disclosure in accordance with 10 CFR 2.390 in that portions of all files are either commercially proprietary in nature or contain security-related information, respectively.


Question: The Commission ensures that the facility operates according to the mandated standards and regulations.

The application for amendment to the Edwin I. Hatch Nuclear Plant, Unit No. 2 (the facility) Renewed Facility Operating License No. NPF-5 filed by Southern Nuclear Operating Company, Inc. (the licensee), acting for itself, Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia (the owners), dated September 11, 2024, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission’s rules and regulations as set forth in 10 CFR Chapter I;

The Commission's regulations also address other types of energy production methods that do not involve nuclear reactions.


Question: The NRC mandates maintaining adequate levels of TSP-C in the storage baskets to ensure effective pH adjustment during post-accident scenarios.

Periodic determination of the amount of TSP-C in containment must be performed due to the possibility of leaking valves and components in the containment building that could cause dissolution of the TSP-C during normal operation.

The allowed Completion Time of 6 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.


Question: What are MD 8.3 deterministic questions in nuclear regulation?

The non-concurrying employee presents disagreement in how three MD 8.3 Deterministic questions were answered by RIII staff: - Question D - "Led to the loss of a safety function or multiple failures in systems used to mitigate an actual event," - Question E - "Involved possible adverse generic implications," and - Question G - "Involved repetitive failures or events involving safety-related equipment or deficiencies in operations."

As the DRP Director, I was in the process of reviewing the draft MD 8.3 document when I was informed that the employee was no longer concurring on the draft MD 8.3.


Question: What is the role of a Contracting Officer's Representative (COR) in nuclear regulatory processes?

The COR is in DORL, so we won’t be responsible for this part. This is just a heads up for your awareness.

I was concerned this was the one John B was COR on for John L, but thanks for the correction.


Question: What is the core reflood rate defined in nuclear regulatory filings?

The regulation at 10 CFR 50.46(b) does not include criteria for calculated ECCS cooling performance pertaining to the core reflood rate following postulated LOCAs.

The petitioner requested the NRC to amend RG 1.157 and appendix K to 10 CFR Part 50, to require that the rates of energy release, hydrogen generation, and Zircaloy cladding oxidation from the metal-water reaction of zirconium with steam considered in evaluation models used to calculate ECCS cooling performance be calculated based on data from cited experiments, instead of using the Cathcart-Pawel or Baker-Just equations.


Question: The limits for maximum initial enrichment of 235U and the corresponding absorber content are crucial for safety assessments in fuel assembly designs.

| Max. Initial Enrichment ( wt % 235U) | Absorber 0.0225 ¹⁰B g/cm² | Absorber 0.02 ¹⁰B g/cm² |

Nuclear reactors require comprehensive analyses for efficient fuel cycle management and operational safety.


Question: In an acoustic survey conducted in 2015, the presence of the tricolored bat was confirmed at the Oconee Station site in relation to the independent spent fuel storage expansion.

As described previously in this section, the tricolored bat was identified in a 2015 acoustic bat survey to be associated with the Oconee Station independent spent fuel storage expansion.

Monarchs are associated with prairie, meadow, and grassland habitats. Within South Carolina, 13 native milkweed species provide habitat for the development of monarch eggs and larvae, the most prevalent of which is the common milkweed (*Asclepias syriaca*). It is unknown whether milkweed occurs on the Oconee Station site, although grasslands within the action area are undeveloped and would remain undisturbed during the proposed license renewal period.


Question: The evaluation of defense-in-depth strategies is crucial for ensuring safety across nuclear facilities.

Defense-in-depth is a fundamental safety concept that refers to the use of multiple layers of protection to prevent accidents and mitigate consequences should an accident occur.

The proposed digital modernization plan outlines several enhancements intended to improve operational efficiency and reliability for the Turkey Point Unit 3&4.


Question: isotopic analysis of reactor coolant

ENSURE Chemistry is performing Isotopic Analysis of the RCS for Dose Equivalent Iodine and Dose Equivalent Xenon.

MONITOR area radiation monitors and recorder 1-RR-90-1 NORMAL.


Question: nuclear reactor licensing process

The licensing process for nuclear reactors involves multiple phases, including pre-application, application, review, and approval stages as mandated by the Nuclear Regulatory Commission (NRC).

Cooling systems are essential for maintaining optimal temperatures within the reactor core, ensuring safe and efficient operation during normal and emergency conditions.


Question: The expected radiation levels in the vicinity of the reactor must be monitored during operation.

The unit of measurement used for radioactivity in these reports is counts per minute (cpm).

The logged readings for each sample are recorded in the specified format, providing a clear overview of the data collected.


Question: The NRC monitors and regulates the operation of emergency cooling systems to ensure reactor safety during unexpected events.

The Emergency Core Cooling Systems (ECCS) are designed to provide adequate cooling to the reactor core following a loss of coolant accident.

Reactor coolant system pressure and temperature limits must be maintained within specified ranges during normal operations.


Question: The analysis of energy peaks is critical when assessing the efficiency of nuclear detection systems.

Energy (keV) is a crucial parameter in evaluating the performance of nuclear measurement instruments, particularly in identifying specific isotopes.

The ambient background for the measurements must be recorded to ensure accurate data analysis during the nuclear assessments.


Question: specific conductance measurement in water quality reports

Specific Conductance is measured in microsiemens per centimeter (μS/cm) and is a key factor in determining water quality.

Aluminum had a measured value of 6 micrograms per liter (ug/L), indicating the presence of this element in the water samples.


Question: uranium production tax assessment

The assessed value of uranium production for tax purposes is 50% of the sales price.

The total school enrollment included the following: Crownpoint Elementary School (enrollment 331), Crownpoint High School (396), Crownpoint Middle School (259), all operated by the Gallup McKinley County School District.


Question: What are the IEEE-450 battery replacement criteria?

The proposed change revises the Battery Parameters SR acceptance criteria for verifying battery capacity following a performance test discharge or a modified performance test discharge for the 125 VDC batteries to align with the IEEE-450 recommended battery replacement criteria.

The proposed change revises the Monticello TS battery parameter surveillance requirement (SR) acceptance criteria for the 125 VDC batteries to adopt the IEEE-450 recommended replacement criteria following battery replacement that raised overall capacity.


Question: nuclear licensing process

The nuclear licensing process involves a series of regulatory steps that must be satisfied before a facility can be constructed or operated, including safety analysis, environmental review, and public participation.

The reactor's coolant system is designed to maintain safe operational temperatures, ensuring the integrity of the reactor core under various load conditions.


Question: What information is required for devices subject to nuclear registration?

Provide information about other devices you have that are subject to registration. Do not report specifically licensed devices.

Isotope (e.g. AM241) | Activity (e.g. 100) | Unit (e.g. mCi)


Question: What is the Nuclear Steam Supply System (NSSS)?

The NSSS, or Nuclear Steam Supply System, is the portion of a nuclear power plant that generates steam to drive the turbines and produce electricity.

The Clean Water Act (CWA) regulates discharges of pollutants into the waters of the United States and quality standards for surface waters.


Question: The NRC establishes specific safety requirements that licensees must follow for their operations.

The regulatory framework for nuclear facility licensing encompasses extensive safety protocols to ensure the protection of public health and safety during the operation of nuclear reactors.

Regular updates and maintenance of communication systems are vital for operational efficiency in all sectors.


Question: The NRC mandates that licensees must submit a report detailing the upgraded probabilistic risk assessment methods before their first application for calculating the Risk Informed Completion Time.

A report describing newly developed methods and their implementation must be submitted following a probabilistic risk assessment (PRA) upgrade associated with newly developed methods and prior to the first use of those methods to calculate a RICT.

The categories of items required to be in the TSs are then listed in 10 CFR 50.36(c).


Question: What measurement is used to quantify cladding diameter changes in relation to reactor burnup?

Figure 33-3 illustrates the relationship between measured cladding diameter change and burnup.

The cooling system is essential for maintaining a stable reactor core temperature during operations.


Question: Resources Databook in emergency planning

The Resources Databook (RDB) is organized in two volumes. Volume 1 contains generic information for all five Parishes and includes definitions of Protective Action Sections, decision making suggestions, and evacuation routes.

A scenario is a grouping of Protective Action Sections requiring a unified response (i.e., sheltering, evacuation, or combination of sheltering and evacuation) depending on the wind direction and the protective action recommendation by River Bend Station and/or the Louisiana Department of Environmental Quality.


Question: What is Strontium-90 and its significance in nuclear analysis?

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


Question: How are control displays arranged to ensure operator visibility and functionality?

Controls and displays (including annunciators and recorders) are arranged in functional groups and in a functional sequence with the displays above their corresponding controls to prevent the arm of the operator from obscuring the indicated response to control adjustments.

Maintenance accessibility is designed for minimum operator maintenance and technician interference by maintaining sufficient aisle clearance and by positioning panels so that there is adequate aisle space for panel back access.


Question: What are the qualifications for Radiation Protection personnel in emergency response?

The proposed emergency plan reduces the number and qualifications of minimum staff ERO positions for the Radiation Protection function. Specifically, this change reduces the overall number of minimum staff ERO positions used for the Radiation Protection function from six (6) in the current emergency plan and in NUREG-0654, to five (5) in the proposed emergency plan, consisting of three (3) RPTs and two (2) RP Qualified individuals that are task trained but not required to be ANSI equivalent RPTs.

During the plant operational phase, routine maintenance checks are performed on all safety-related systems to ensure operational readiness and regulatory compliance.


Question: What strategies are associated with 10 CFR 50.54(h)(2) for emergency response?

In this regard, NUREG-0654 R2 guidance on the demonstration of the use of equipment, procedures, and strategies developed in compliance with 10 CFR 50.54(h)(2) [10 CFR 50.155(b)(2)] are as follows:

The NRC noted that the conduct of drills or exercises was not included as a requirement in the Mitigation Strategies Order, instead being an element of an acceptable approach to meeting the order’s requirement for training.


Question: The NRC emphasizes the importance of understanding critical heat flux in reactor design and safety assessments.

The COBRA-FLX model provides a comprehensive analysis of critical heat flux correlations that are essential for ensuring fuel performance and safety in nuclear reactors.

Table B-3 presents a detailed account of local conditions derived from various simulations, specifically tailored for grid tests.


Question: non-LOCA DBA

We're effectively just being consistent with the other non-LOCA DBAs that we've assessed with the well within criteria 6.3 rem.

But I guess I'm still struggling. You are talking about a large break LOCA DBA. Isn't that a contradiction in terms now? A large break LOCA is a BDBA.


Question: The NRC mandates the evaluation of seismic effects for the safety of nuclear facilities.

Regulation 10 CFR Part 50, Appendix S, “Earthquake Engineering Criteria for Nuclear Power Plants” is applicable to applications for a construction permit or operating license pursuant to 10 CFR Part 50 on or after January 10, 1997.

This audit will focus on information provided by TerraPower in its online reference portal and during virtual meetings.


Question: nuclear licensing process

But you got it great. It can be for potential licensees coming to us for a license.


Question: digital instrumentation and controls license amendment

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). The end of the audit will be scheduled after the acceptance review of the planned Limerick digital instrumentation and controls license amendment request is completed.

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: significant reduction in margin of safety

The proposed change does not involve a significant reduction in a margin of safety.

The proposed change revises the inspection frequencies for SG tube inspections and associated reporting requirements.


Question: What is the Minimum Detectable Concentration in regulatory filings?

The LaCrosseSolutions site is located along the Mississippi River, providing a unique scenic environment for restoration activities.


Question: reactor coolant system

The inspectors evaluated system configurations during partial walkdowns of the following systems/trains: Reactor coolant system level instrumentation during reactor coolant system drain down and evacuation and fill on October 28, 2024.

Samples were declared complete when the IP requirements most appropriate to the inspection activity were met consistent with Inspection Manual Chapter (IMC) 2515, “Light-Water Reactor Inspection Program - Operations Phase.”


Question: This letter serves to inform the NRC about the transfer of tritium exit signs to an authorized disposal facility as outlined in regulatory guidelines.

Per the requirements of 10 CFR 31.5 (c) (8), this letter is to provide notification that the self-luminous tritium exit signs listed on the attached documents were transferred to Curie Environmental Services, 4020 Vassar Drive NE, Suite D, Albuquerque NM 87107 (State of New Mexico Specific License RS500-03) on Friday, September 20, 2019.

These signs were originally installed at: Home Depot #6002, 1740 Fleischli Pkwy, Cheyenne, WY 82001 (307) 632-2897.


Question: What is the function of the RCIC system in a nuclear reactor?

The Reactor Core Isolation Cooling (RCIC) system is designed to provide cooling to the reactor core during certain emergency conditions by injecting water into the reactor pressure vessel (RPV).

The grid operator confirmed that an overload condition was occurring on Bus 11, and measures were taken to redistribute loads across the available buses.


Question: reactor thermal power procedures

2-OHP-4030-214-029 REACTOR THERMAL POWER


Question: general license registration

NRC will use this information to track general licenses and their devices to ensure a higher level of device accountability.

The NRC may not conduct or sponsor, and a person is not required to respond to, a collection of information unless the document requesting or requiring the collection displays a currently valid OMB control number.


Question: What is the significance of the SRO licensing structure?

The SRO licensing structure is designed to ensure that individuals responsible for supervising operations at nuclear facilities possess the necessary qualifications and training to uphold safety standards.

The table indicates various ratings for the performance metrics of different operators across several operational criteria.


Question: ABWR ATWS MSIVC SLCS Case

Figure D.7-2 illustrates the reactor pressures during the Advanced Boiling Water Reactor (ABWR) Anticipated Transient Without Scram (ATWS) Main Steam Isolation Valve Control (MSIVC) and Standby Liquid Control System (SLCS) case.

The reactor noise analysis was conducted to ensure that the vibrations are within the acceptable limits during normal operations.


Question: commercial grade dedication in nuclear licensing

Because the application logic and hardware configuration will be plant specific, the scope of the dedication activities is limited to the RadICS platform hardware, as well as platform Function Block Logic Library (FBL) and ED logic. Application hardware configuration and logic development will be performed by a licensee and Radics LLC under a 10 CFR Part 50, Appendix B compliant QA program.


Question: What is the population distribution within a ten-mile radius of a nuclear facility?

As seen in Figure 2.1-4, the population distribution within ten miles varies significantly across different directions, with specific numbers indicated for each directional segment.

The table for Population Inside One Mile shows that the total population is much smaller, with only a handful of people recorded in the northern and northeastern sectors.


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

The nuclear operating license is obtained through a structured process involving the submission of various regulatory filings that demonstrate compliance with safety and operational standards set by the NRC.

This document may contain attachments that provide additional information on reactor design specifications and configurations.


Question: The assessment of large break loss-of-coolant accident (LBLOCA) size ranges is critical for safety analysis in nuclear reactor operations.

LBLOCA Size Range Discretization (6-15, 15-25, and 25-31 inches) is essential for evaluating the associated uncertainty in model predictions.

The results of the model uncertainty case are presented in a table format, which highlights the changes in ΔCDF between various scenarios.


Question: The operability of Main Feedwater Isolation Valves is crucial for ensuring safety in certain operational modes.

The MFIV, MFRV, and MFRVBV in each of the four main feedwater lines shall be OPERABLE.

During the reactor's operational phases, numerous tests are conducted to assess the performance of various systems.


Question: What is the role of the Corrosion and Metallurgy Branch in nuclear regulatory research?

The Corrosion and Metallurgy Branch is part of the US Nuclear Regulatory Commission's Office of Nuclear Regulatory Research, focused on materials engineering and ensuring the safety and integrity of components within nuclear systems.

The attached draft announcement for the harvesting workshop mentions various date options for the workshop but suggests a focus on March 16-17.


Question: aging management programs for fire protection systems

The Fire Protection (B.2.1.15) program will be used to manage the aging effect(s) applicable to this component type, material, and environment combination.

The review of 10-years of plant specific operating experience has not identified cracking in copper alloy with 15% zinc or more in air environments due to potential exposure to ammonia or ammonium compounds.


Question: What is a temporary package sewage treatment facility?

A temporary package sewage treatment facility will be installed to treat domestic wastes from a peak construction force of approximately 2,500 persons.

Feasibility, costs, and the minimization of turbidity and siltation in the reservoir will be taken into consideration when determining what type of dredge will finally be used.


Question: The facility must obtain different permits and certifications to ensure compliance with environmental regulations during construction and operation.

Tennessee Department of Environment and Conservation, Division of Water Resources | Federal Clean Water Act | Notice of Coverage under the General National Pollutant Discharge Elimination System (NPDES) Permit for Stormwater Discharges associated with Construction Activity | Required for clearing, grading, or excavation that disturbs one or more acres.

However, two channels and one stream were identified within the project survey area, outside the site boundary, with the potential to be regulated by either the USACE and/or TDEC.


Question: What is the aging management program for bolting integrity?

Loss of Material and Loss of Preload are aging effects requiring management for bolting closures, specifically addressed under the Bolting Integrity program (B.2.3.9).

In addition to the mechanical closure applications, the filters are designed to maintain the integrity of the system by removing particulate contamination from the process flow.


Question: strainer failure criteria in nuclear regulation

The strainer failure definition of 1/8" of fiber represents a common threshold associated with the onset of thin-bed formation, which makes it suitable for a baseline risk evaluation against which variations can be compared.

Designation of a 1/8" fibrous debris strainer failure threshold encompasses many failures that might result from large particulate debris loadings in combination with loss of strainer area, provided that at least a minimum amount of fiber arrives at the strainer to form a filter medium.


Question: lift-off force in nuclear safety analysis

Lift-off Force (YDP) is calculated as a function of time, represented by the trend line equation: F_T = -1.5735 - 414.65 ⋅ log_{10}(T).

The trend line clearly indicates the correlation between the variables measured in the experiment, allowing for further analysis on the stability of the reactor components.


Question: environmental qualification in nuclear regulation

Part of the LAR included the U.S. Nuclear Regulatory Commission (NRC) review and approval of the licensee’s information related to environmental qualification (EQ) of relevant electrical equipment.

The NRC staff will review the licensee’s information related to EQ of relevant electrical equipment to determine if any additional information will require docketing to support or develop conclusions for the NRC staff’s safety evaluation of the requested license amendment.


Question: geotechnical instrumentation monitoring

Instrumentation around excavation support systems is typically installed to monitor deflections during the excavation.

An instrumentation plan and monitoring schedule will be presented in the COLA.


Question: The components of the Sampling and Water Quality System require assessment for their aging effects due to functional importance.

The following aging effects associated with the Sampling and Water Quality System components require management: - Cracking - Hardening or loss of strength - Flow blockage - Loss of fracture toughness - Loss of material - Loss of preload - Reduction of heat transfer

The Building Heating System components are exposed to the following environments: - Air - Air - indoor uncontrolled - Air, condensation - Any - Closed-cycle cooling water - Treated Water.


Question: The table provided includes various nuclides and their corresponding energy levels, yields, and activities.

The reactor's cooling system is designed to handle heat removal effectively while maintaining safety protocols.


Question: service water system operation

The three 8-inch SW flow paths to the MCR/ESGR AC subsystem (referred to in UFSAR Section 9.9.2) are the 1D, 2A, and 2C SW supply headers; two of the three 8-inch SW headers are normally in service.

Significant defense in depth for the chillers and the flow paths to supply SW to the MER 3 and MER 5 chillers is available; however, manual action would be required to restore chiller operability.


Question: What is the minimum initial assembly average enrichment for PWR fuel?

Minimum Initial Assembly Avg. Enrichment wt % 235U (E) specifies the lower limits of enrichment for fuel assemblies based on the enrichment range for various configurations in a pressurized water reactor (PWR).

The NRC establishes guidelines for the general operation of nuclear reactors, ensuring that all safety measures are adequately addressed during the regulatory review process.


Question: regulatory measurements for surface area

The survey conducted in TRIGA Room 112 reported an area of 36.9 m² (397 ft²) for Survey Unit 13, indicating the spatial dimensions relevant for compliance with safety regulations.

The symbols used in this survey document include a red circle for a random starting point, a black circle for systematic locations, and an orange circle for judgmental locations.


Question: nuclear emergency planning procedure

The Revised Emergency Plan Implementing Procedure provides detailed guidelines for responding to nuclear emergencies, including evacuation protocols and communication strategies with local authorities.

The reactor's cooling system is designed to maintain safe operating temperatures and prevent overheating during normal and abnormal conditions.


Question: The requirement for estimating the water use for the circulating water system is crucial for operational efficiency.

Plant Well Water Demand is quantified as 752 gpm under normal conditions and can peak at 3,140 gpm.

Demineralized Water System Feed relies on a supply of treated water to ensure the proper functioning of various plant systems.


Question: What is required for registration of devices used in nuclear applications?

Devices subject to registration must be listed with an NRC Device Key and include relevant details such as distributor information and mechanism of license number.

The manufacturer name recorded in the document is TN TECHNOLOGIES, INC., known for producing various electronic devices.


Question: What is a fire protection system in nuclear facilities?

The term “fire protection system” refers to the integrated complex of components and equipment provided for detection and suppression of fires.

Total flooding carbon dioxide suppression systems are provided in the auxiliary electric equipment room (AEER), three diesel generator rooms, and diesel day tank rooms.


Question: aging management programs

The scope of the AMP, AMR, and TLAA review is to: Examine the applicant’s AMPs, AMR items, and TLAAs; Verify the applicant’s claims of consistency with the corresponding GALL-SLR Report AMPs, and AMR items; and Assess the adequacy of the TLAAs.

The three Aging Management Audit parts include: Part #1 – In-Office Technical Review Audit During the In-Office Technical Review Audit, technical staff will review the SLRA, documentation on the applicant’s electronic portal (ePortal), and conduct an OE review.


Question: The NRC safety evaluation for NEI 06-09-A outlines requirements for conducting peer reviews and self-assessments of probabilistic risk assessment models.

Section 4.0, Item 3 of the Nuclear Regulatory Commission’s (NRC’s) Final Safety Evaluation (Reference 1) for NEI 06-09-A, Revision 0, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines" (Reference 2), requires that the license amendment request (LAR) provide a discussion of the results of peer reviews and self-assessments conducted for the plant-specific probabilistic risk assessment (PRA) models that support the RMTS, including the resolution or disposition of any identified deficiencies (i.e., facts and observations [F&Os] from peer reviews).

This enclosure provides information on the technical acceptability of the Diablo Canyon PRA internal event, internal flood, fire, and seismic models that support the Risk-Informed Completion Time (RICT) Program, in support of the LAR to revise the Technical Specifications to adopt Technical Specifications Task Force (TSTF) traveler TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times – RITSTF Initiative 4b" (Reference 6).


Question: emergency planning procedures

An annual review of the plan, certified to be current.

Supervising the development and implementation of training and public information and education programs within the Parish.


Question: what is the continuity equation in nuclear engineering?

The continuity equation is a mathematical expression of the conservation of mass principle.

For a control volume that has a single inlet and a single outlet, the mass flow rate into the volume must equal the mass flow rate out.


Question: The radiation levels recorded at various scan points are essential for ensuring safety compliance.

Scan Logged Result (cpm) indicates the amount of radiation detected at various detector locations, which is critical for monitoring safety levels.

The average background radiation is measured in counts per minute (cpm) to establish a reference for normal environmental radiation.


Question: The system requires multiple channels of process control equipment to manage unit parameters effectively.

Generally, three or four channels of process control equipment are used for the signal processing of unit parameters measured by the field instruments.

The actual number of channels required for each unit parameter is specified in Reference 1.


Question: Zr-4 cladding performance under neutron irradiation

Figure 30-15: Zr-4 Cladding, P/M Strain versus Fast Fluence illustrates the relationship between plasticity/market strain and the amount of fast neutron fluence in Zr-4 cladding material used in nuclear reactors.

Figure 30-16: Zr-4 Cladding, P/M Strain versus Temperature depicts how temperature variations affect the strain characteristics of Zr-4 cladding, providing insights into material behavior under thermal conditions.


Question: The conditions for test assemblies were meticulously documented to ensure accurate results in critical heat flux studies.

Test assembly conditions for non-mixing grid tests were established to evaluate the performance of the COBRA-FLX thermal-hydraulic code.

The performance of a nuclear reactor is influenced by numerous factors including fuel type, coolant flow, and moderator characteristics.


Question: The report provides identified nuclides and their corresponding confidence levels along with measurements such as activity.


Question: nuclear construction inspection reports

NRC CONSTRUCTION INSPECTION REPORT(S)  | NRC VENDOR INSPECTION REPORT(S)

Federal Register Notice Published  | 85FR6001


Question: What are the requirements for a one-pager in nuclear regulatory filings?

I’d prefer to keep it to a one-pager (or 1+ pager) and not go to the effort making slides.

I scheduled this for 30 minutes at Rob’s suggestion. He didn’t think they would need longer than that, although perhaps this is new material for Chris.


Question: radiological dose from Sr-90

The average dose per ROC for Sr-90 is 1.34E-04, indicating a significant measure of the radiological effects associated with this radionuclide.

The on-site laboratory analyzed the three (3) judgmental soil samples using the on-site gamma spectroscopy system.


Question: PWR reactor vessel internal components

Stainless steel, nickel alloy, stellite PWR reactor vessel internal components or LRA/SLRA-specified reactor vessel internal component exposed to reactor coolant, neutron flux

Plant-specific aging management program or AMP XI.M16A, "PWR Vessel Internals," with an adjusted site-specific or component-specific aging management basis for a specified reactor vessel internal component.


Question: Cs-137 detection in soil samples

Of the total number of surface soil samples taken and analyzed, Cs-137 was identified at concentrations greater than the MDC in 132 samples.

Using Type I and Type II error values (α and β) of 0.05, a relative shift of four (4) equates to a sample size of nine (9). The PSR Project elected to use a conservative sample size of fourteen (14) for each survey unit.


Question: What is the purpose of SAMA in nuclear regulatory assessments?

The Stage 1 assessment involves "demonstrating that SAMA implementation would not reduce the CDF or any of the Level 2 release category frequencies by 50 percent or more".

Please clarify this difference and confirm that all aspects of the ER make use of "the latest risk models that are available for internal events (including internal flooding) and for each of the external events contributors...", as discussed in Section 3.1 of NEI 17-04.


Question: What is the status of the SLR guidance document in regulatory filings?

Please cover the latest SLR guidance document status and plans: we're basically in the internal review and concurrence portion of the process.

Also, please provide information on the latest plans for the material harvesting workshop during the RIC.


Question: What is the VIPRE computer code in nuclear reactor safety analysis?

The computer code VIPRE (Versatile Internals and Components Program for Reactors - EPRI) was developed for the Electric Power Research Institute (EPRI) by Battelle Pacific Northwest Laboratories to perform detailed thermal-hydraulic analyses to predict CHF and DNBR of reactor cores.

Dominion Energy submitted to the USNRC in Reference 6, and received approval from the USNRC in Reference 7, a request to adopt the Dominion Energy's Core Design and Safety Analysis Methods at MPS3.


Question: What does a safety analysis report entail for nuclear reactors?

A safety analysis report must identify and analyze the potential hazards associated with the operation of the nuclear reactor, including the evaluation of the reactor design and safety systems.

The plant management team is responsible for overseeing daily operations and ensuring compliance with standard operating procedures.


Question: What is a fiber penetration model in nuclear safety analysis?

The penetration models from the previous step can then be used to determine the prompt fiber penetration fraction and shedding fraction for a given time and amount of fiber accumulated on the strainer.

For each time step, the fiber penetration rates and quantities were calculated.


Question: What is the definition of the Practical Quantitation Limit?

Preparation method: 'Method' indicates preparation defined in analytical method.


Question: non-cited violation (NCV)

We are treating this violation as a non-cited violation (NCV) consistent with Section 2.3.2 of the Enforcement Policy.

The results of this inspection are documented in the enclosed report.


Question: What is direct fracture toughness in reactor pressure vessel evaluation?

Direct fracture toughness is used for the evaluation of reactor pressure vessel integrity, allowing for the assessment of material performance under service conditions.

The NRC Audit Report Summary Package provides insights into the review of regulatory documents related to nuclear operations.


Question: pitting and crevice corrosion in aluminum alloys

Loss of material due to pitting and crevice corrosion is not an aging effect requiring management for aluminum alloys if: 1. Plant-specific OE does not reveal a history of loss of material due to pitting or crevice corrosion, and 2. A One-Time Inspection demonstrates that the aging effect is not occurring or is occurring so slowly that it will not affect the intended function of the components. The applicant documents the results of the plant-specific OE review in the SLRA.

An alternative strategy to demonstrating that an aggressive environment is not present is to isolate the aluminum alloy from the environment using a barrier to prevent loss of material due to pitting and crevice corrosion.


Question: What are the parameters inspected in the Pressurizer Spray Head Examination?

The parameter inspected by the Pressurizer Spray Head Examination is cracking of the pressurizer spray head due to thermal embrittlement.

The Non-EQ Insulated Cables and Connections Aging Management Program will be controlled by an engineering support program.


Question: The containment structure is designed to prevent the release of radioactive materials.

CONTAINM is a term used to refer to containment that is essential in nuclear reactors to prevent the escape of nuclear materials during normal operations and accidents.

The U.S. Department of Labor, Bureau of Labor Statistics provides important labor data but does not directly influence nuclear safety regulations.


Question: What is the FEIS in nuclear regulatory filings?

The FEIS is available through the ADAMS Public Documents collection at NRC website under ML25031A307.

If you have any questions regarding this matter, please contact Jean Trefethen by phone at 301-415-0867 or via email at Jean.Trefethen@nrc.gov.


Question: The aging management program for electrical cable connections will include sampling and testing for integrity over time.

The BFN Fuse Holders aging management program is a new condition monitoring program. The Fuse Holders program will apply to fuse holders outside of active equipment within the scope of subsequent license renewal and susceptible to the following aging effects: increased resistance of connection due to chemical contamination, corrosion, and oxidation or fatigue caused by ohmic heating, thermal cycling, electrical transients, frequent removal and replacement, or vibration.

The program will utilize thermography and visual inspection and testing to identify age-related degradation for both fuse holder electrical insulation material and fuse holder metallic clamps.


Question: nrc radioactive materials license termination

We decided to terminate our NRC Radioactive materials license effective on December 31, 2024.

I wanted to inform you that our RSO Lino T. Vera is back to his office, and we are ready to be inspected any time.


Question: nuclear device registration requirements

Certain devices that do not present a significant risk to public health and safety are not subject to registration with the NRC, which streamlines regulatory processes for manufacturers of low-risk equipment.

The manufacturer must ensure that all equipment is properly calibrated and maintained to prevent any operational anomalies during testing.


Question: The NRC mandates the quantification of radioactive isotopes to ensure compliance with safety standards.

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

The results of the analytical methods indicate that the measurements were performed using EPA 905.0 Modified.


Question: The potential impacts of pollutants from nuclear plant operations on marine life were previously assessed due to the presence of heavy metals in infrastructure materials.

This issue concerns the potential effects of nonradiological contaminants on aquatic organisms that could occur as a result of nuclear power plant operations.

With respect to gas supersaturation, Duke Energy has not reported any instances of fish kills at Oconee Station or any other information indicating that fish may have experienced symptoms of gas bubble disease.


Question: The site has been evaluated for the possibility of tectonic activity, and it has been determined that no capable tectonic sources exist nearby.

There are no capable tectonic sources within 5 miles of the ESP site.

Subsequent air photo interpretation, aerial reconnaissance, and field reconnaissance of the fault trace carried out for this ESP application confirms the conclusion reached by Dames & Moore (Reference 9), and no further investigation of this fault is considered necessary.


Question: geotechnical exploration for nuclear sites

MACTEC Engineering and Consulting, Inc., “Geotechnical Exploration and Testing, PSEG Site Application, Lower Alloways Creek Township, New Jersey”, Rev. 0, Jul 10, 2009.

ASTM, 2007a, “Standard Practices for Preserving and Transporting Soil Samples,” ASTM D 4220-07, 2007.


Question: What are limitations and conditions in nuclear regulatory filings?

The NRC staff reviewed and performed independent calculations to assess Framatome's additional information regarding limitation and condition (L&C)-18, presented at the May 4, 2023, meeting.

NRC staff will also modify the supporting language in the SE to be consistent with the L&C changes.


Question: What defines the ordinary high-water level in relation to watercourses and reservoirs?

Ordinary High-Water Level (OHWL) means the boundary of waterbasins, watercourses, public waters, and public waters wetlands, and shall be an elevation delineating the highest water level which has been maintained for a sufficient period of time to leave evidence upon the landscape, commonly that point where the natural vegetation shifts from predominantly aquatic to predominantly terrestrial. For watercourses, the ordinary high water level is the elevation of the top of the bank of the channel. For reservoirs and flowages, the ordinary high water level is the operating elevation of the normal summer pool.

Stabilized means staked sod, riprap, wood fiber blanket, or other material that prevents erosion from occurring has covered the exposed ground surface. Grass seed is not stabilization.


Question: What is peak analysis in gamma spectrum analysis?

Peak analysis is performed on the data collected on 9/15/2017 at 1:33:11 PM, where the analysis identifies peaks based on the channel range from 100 to 4096.

The sample taken on 9/15/2017 was a 500 ml Marinelli type, which contains a total sample size of 8.417E+02 grams.


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

Defense-in-depth is a fundamental concept in nuclear safety, which ensures that multiple layers of protection are in place to prevent accidents and mitigate their consequences.

The evaluation of the digital modernization project at Turkey Point Unit 3&4 also considers the improvement of operational efficiency and reliability of the reactors.


Question: During Condition B, the RPS can only perform its function if one channel in each trip system is operable, limiting its capability to withstand a single failure.

Condition B exists when, for any one or more Functions, at least one required channel is inoperable in each trip system.

Completing one of these Required Actions restores RPS to a reliability level equivalent to that evaluated in References 9 and 13, which justified a 12 hour allowable out of service time as presented in Condition A.


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

Peak Centroid refers to the energy value at which the maximum intensity of a peak occurs in a spectrum, representing the energy of the emitted radiation.

The Ambient Background is subtracted from the Original Area to calculate the Subtracted Area, which is crucial for obtaining accurate measurements in spectroscopic analysis.


Question: The NRC conducts reviews and grading of various examination components to ensure compliance with licensing requirements.

Reviewed and incorporated any necessary facility and applicant written examination comments. NRC grading of written examination completed. (ES-4.4)

Mailed licenses, preliminary results, and pass letters.


Question: environmental impacts of nuclear licensing

The environmental impacts of the proposed action and reasonable alternatives are designated as SMALL, MODERATE, or LARGE, which represent three established significance levels for potential impacts, presented in a footnote of Table B–1 in Appendix B to Subpart A of 10 CFR Part 51 (TN250), and defined as follows:

Table ES-1 lists the environmental issues applicable to Oconee Station SLR and the findings related to these issues.


Question: The Commission's decision on Powertech's license is under review due to potential NEPA violations.

The Tribe relies on 5 U.S.C. § 706, which generally provides the standard for judicial review of agency action, and related federal court precedent to argue that, unless an analysis undertaken pursuant to Allied-Signal warrants rebutting the presumption of vacatur, the Commission should vacate Powertech’s license based on the finding of a NEPA violation.

The parties have provided their views in response to that order, and the Tribe, Powertech, and the Consolidated Intervenors have also filed responses to those initial filings.


Question: What is included in nuclear licensing submittals?

The licensing submittals typically include information about equipment specifications, regulatory compliance documentation, and safety analysis reports.

The report details various part numbers and vendor information for nuclear components.


Question: The ability to assess the reactor coolant system is fundamental in ensuring the steam generator operates safely and efficiently.

Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions on the steam generator system:

Failure to close the MSIV to isolate the faulted steam generator, given a steam generator tube rupture event (PRA related) 4.4.


Question: What is the Alert and Notification System in nuclear safety?

The Alert and Notification System (ANS) Siren Warning System (SWS) was updated as part of a siren upgrade project.

The design report was rewritten and reformatted to meet FEMA design report guidelines.


Question: What are the testing procedures for sealed sources in nuclear licensing?

Unless a sealed source is accompanied by a certificate from the transferor that shows that it has been leak tested within 6 months before the transfer, it may not be used by the licensee until tested for leakage.

Each licensee or registrant shall conduct a quarterly physical inventory to account for all sources of radiation, and for devices containing depleted uranium received and possessed under the license.


Question: What actions should the crew take in response to a loss of control air on Unit 1?

In accordance with S1.OP-AB.CA-0001, the crew's NEXT action will be to TRIP the Reactor and perform 1-EOP-TRIP-1 (Reactor Trip Or Safety Injection) CONCURRENTLY with S1.OP-AB.CA-0001 Attachment 12 (LOCAL CONTROL OF SG PRESSURE AND LEVEL).

The 1A Control Air Header Pressure is 79 psig and the 1B Control Air Header Pressure is 77 psig.


Question: online portal security requirements

The following conditions associated with the online portal must be maintained throughout the duration that the audit team has access to the online portal: - The online portal will be password-protected, and separate passwords will be assigned to the NRC staff and contractors who are participating in the audit. - The online portal will be sufficiently secure to prevent the NRC staff from printing, saving, downloading, or collecting any information on the online portal.

The NRC project manager will coordinate any changes to the audit schedule and location with the licensee.


Question: What is a 10 CFR 71.95 report?

10 CFR 71.95 report addressing an uninstalled leak test port bolt found in a heeled Liqui-Rad transport unit package at the customer site.

Shipping Low-Level Radioactive Waste training documents


Question: leakage boundary definition

Pump Casing (Equipment Drain Receiver Pumps) | Leakage Boundary (Spatial)

Tank (Floor Drain Receiver Tanks) | Leakage Boundary (Spatial)


Question: The NRC's conduct regarding public comments can influence subcommittee actions based on committee obligations.

In fact, Mark Edward Leyse first learned about the extended deadline when the ACRS Thermal Hydraulics Phenomena Subcommittee briefly discussed the matter on Monday, October 18, 2010.

Those usually go to the full committee. I don't know what your obligation is with respect to those.


Question: What is the purpose of estimating the PMF at the VEGP site?

A stage-discharge relationship or 'rating' curve is required to estimate the water surface elevation of the Savannah River near the VEGP site associated with the PMF discharge.

A logarithmic plot of the power curve fit to these values is presented in Figure 2.4.3-1.


Question: canister heat loading category

Establish LOADING CATEGORY by determining the total heat load for all of the fuel assemblies to be loaded into a CANISTER and then determining the maximum individual fuel assembly heat load for these assemblies.

Certificate of Compliance No. 1025 B2-16 RENEWED AMENDMENT 8 REVISION 1


Question: The NRC mandates that all relevant safety systems are designed to handle postulated accidents effectively.

The Subcommittee has reviewed the proposed specifications on the Standby Liquid Control System (SLCS), which provides critical safety function during accident scenarios in boiling water reactors (BWR).

The licensee submitted a request for an extension of the operational timeframe for Unit 4, detailing the planned maintenance schedule for the next three years.


Question: What are the quality assurance requirements for the installation of nuclear instrumentation?

Regulatory Guide 1.30, August 1972, Quality Assurance Requirements for the Installation, Inspection, and Testing of Instrumentation and Electric Equipment.

Regulatory Guide 1.34, December 1972, Control of Electroslag Weld Properties.


Question: The analysis measured the presence of americium-241 in the solid matrix with a result of 0.00519 pCi/g.

Mr. Jason Q. Spaide can be contacted for inquiries regarding the LaCBWR Site Restoration Project.


Question: The ND2A valve is critical for the removal of residual heat in a nuclear reactor system.

The ND2A valve is specifically designed for the residual heat removal system to ensure the safety of the nuclear reactor by removing excess heat during shutdown.

The inspection process for the reactor containment structure ensures its integrity under various operational conditions.


Question: severe accident mitigation alternatives

The Severe Accident Mitigation Alternatives (SAMAs) are strategies that can be implemented to reduce the consequences of severe accidents in nuclear power plants.

The Generic Environmental Impact Statements (GEIS) evaluate the environmental effects of nuclear plant operations and license renewals.


Question: Cultural Resources Report

Pursuant to 10 CFR 2.390, "Public inspections, exemptions, requests for withholding," Enclosure 2, the Class III Cultural Resources Report, is withheld from public disclosure and may not be made publicly available.

The NRC staff has scheduled a virtual workshop for consulting Tribes on February 25, 2025, from 2:00 p.m. to 5:00 p.m. Eastern Time, to provide information on Kemmerer Power Station Unit 1, discuss the undertaking and potential impacts to historic and cultural resources, and solicit Tribal input on the MOA process and any other concerns related to the undertaking.


Question: The inspectors conducted an inspection related to Transportation and Radiation Protection.

On October 10, 2024, the inspectors presented the Transportation and Radiation Protection inspection results to Roy Stutts and other members of the licensee staff.

The inspectors verified no proprietary information was retained or documented in this report.


Question: corrective action documents

Inspection Procedure 95001 outlines the requirements for managing corrective action documents (CADs) to ensure timely identification and resolution of issues that may affect safety.

The 2J EDG K1 Relay Failure was documented in CA12944602, which highlighted operational challenges faced during routine inspections.


Question: Strontium-90 measurement

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


Question: nuclear license application process

The nuclear license application process involves detailed documentation and analysis to ensure compliance with regulatory requirements set forth by the Nuclear Regulatory Commission (NRC).

During the reactor's operational phase, maintenance checks are crucial to ensure that all safety systems function adequately under normal and emergency conditions.


Question: What is a common cause failure coping analysis in nuclear regulatory contexts?

The regulatory audit plan in support of review of defense in depth and diversity common cause failure coping analysis is a systematic approach to evaluate the robustness of safety systems against simultaneous failures.

The Limerick Generating Station operates under stringent regulatory oversight to ensure the continuous safe generation of electricity for the region.


Question: What are the potential consequences of dam failures on nuclear safety?

Dam breaches can lead to significant flooding and water level increases at nuclear facilities, impacting the structural integrity and safety systems of the plant.

Flood hazard assessments are critical in determining the appropriate flood mitigation strategies for facility operations.


Question: regulatory framework for fusion energy

U.S. policymakers should establish a broad legislative and regulatory framework that explicitly and permanently removes fusion energy from the regulatory approaches that the federal government has taken towards fission power plants.

The development of fusion energy has the potential to revolutionize the global energy landscape, providing a clean and abundant alternative to fossil fuels.


Question: reactor coolant system pressure-temperature limits

Revised the reactor coolant system pressure-temperature limits, LTOPS setpoints, LTOPS T_enable values, and administrative cooldown rate limit.

Made under the provisions of 10 CFR 50.59 except where indicated in brackets.


Question: gamma spectrum analysis

Gamma spectrum analysis involves the examination of gamma rays emitted from a sample to determine its radioactive composition and activity levels, typically performed using specialized detectors and calibration techniques.

The sample number is recorded as 7266, which helps in tracking the specific analysis and results associated with that sample.


Question: completion times for required actions in nuclear regulation

Each Completion Time is referenced to the time that Condition B is entered. The Required Actions of Condition B are to complete action B.1 within 12 hours AND complete action B.2 within 36 hours.

When a system is determined to not meet the LCO, Condition A is entered. If the system is not restored within 7 days, Condition B is also entered and the Completion Time clocks for Required Actions B.1 and B.2 start.


Question: Periodic bathymetric surveys are conducted to assess the need for maintenance dredging of canals at nuclear power facilities.

As described in Sections 3.6.1.2.4 and 4.5.1.4, periodic maintenance dredging of the Units 2 and 3 intake and/or discharge canals may be required during the SLR term.

As a result, a site-specific evaluation of applicable Category 1 environmental issues is required to be included in SLR ERs.


Question: The NRC mandates that all devices containing isotopes must be properly registered for safety and regulatory compliance.

The manufacturer, TN Technologies, Inc., produces various types of electronic devices for commercial use.


Question: What is the significance of the licensing process in nuclear regulatory compliance?

The licensing process for nuclear facilities involves a series of detailed evaluations and consultations with the public and various stakeholders to ensure that the facility operates safely and within the regulatory framework established by the NRC.

The operational efficiency of a reactor is often measured by its thermal output and the effectiveness of heat exchangers in transferring energy.


Question: What is the calculated maximum fuel element cladding temperature limit?

On June 7, 2010, Mark Edward Leyse, on behalf of the New England Coalition, submitted a petition for enforcement action under 10 CFR 2.206, "Requests for Action under This Subpart." The petitioner requested the NRC to order the Vermont Yankee Nuclear Power Station to lower its licensing basis peak cladding temperature to provide an adequate margin of safety in the event of a LOCA.

The NRC identified three main issues in the two petitions. The remaining paragraphs of Section I summarize the following information for each main issue: (1) relevant background information; (2) arguments in the petitions; and (3) specific requests the petitioner made to address each issue.


Question: groundwater contamination management

Oconee Station has implemented a groundwater protection program to identify and monitor leaks through the installed monitoring well network and adheres to the appropriate State pollution prevention permits.

The NRC staff has not identified new and significant information during the audit, scoping process, and review of available information cited in this EIS.


Question: The NRC assesses strategies for enhancing the capability of existing systems, such as the AC cross-tie capability.

SLR SAMA 22 addresses SAMAs related to improving AC cross-tie capability.

The Dresden FLEX mitigating strategy includes an additional 480V generator to satisfy FLEX's 'N+1' approach.


Question: Manning's roughness coefficient in channel routing

Average Manning's roughness coefficient for the whole length of the main channel, exclusive of the overbank areas (0.02-0.065)

Automatic Adaption (1-minute time step for control specification).


Question: What is the purpose of a hydrogen skimmer system in nuclear safety?

The Hydrogen Skimmer System is designed to remove hydrogen gas produced during a potential loss-of-coolant accident, thereby preventing the risk of hydrogen explosions in containment areas.

The Crane Rail Assembly is essential for ensuring that heavy loads can be moved safely within the facility, allowing for maintenance and construction activities to proceed without incident.


Question: What is the methodology used for EPZ sizing in NuScale's report?

NuScale EPZ Sizing Methodology Topical Report (TR) Revision 2 (ML20217L422) uses PRA accident sequences and Design Basis Source Term to size the EPZ for Advanced Light Water Reactors (ALWRs) and Non Light Water Reactors (NLWRs) as applicable.

Discuss staff technical concerns with the NuScale EPZ TR Revision 2.


Question: The testing of PIVs is done to ensure they meet the leakage rate acceptance criteria at the designated operating pressures.

OM Code, Subsection ISTC-3630(a) requires PIV leakage rate testing to be conducted at least once every two years.

To repair or replace a failed PIV, the plant would have to reverse startup activities and cooldown, depressurize, reduce RCS water level, and remove fuel, as required, from the reactor vessel to perform repair or replacement of the failed PIV.


Question: Nominal Trip Setpoints

The Nominal Trip Setpoints are the values at which the bistables are set. Any bistable is considered to be properly adjusted when the 'as left' value is within the two-sided tolerance band for rack calibration accuracy.

To allow for calibration tolerances, instrumentation uncertainties, instrument drift, and harsh environment errors for those RTS channels that must function in harsh environments as defined by 10 CFR 50.49 (Ref. 4), the Nominal Trip Setpoints specified in Table B 3.3.1-1 are conservatively adjusted with respect to the analytical limits.


Question: What is the average boron concentration in the reactor core over time?

The core average boron concentration is measured in parts per million (ppm) and its values at different time intervals indicate the reactor's cooling and chemical control.

The table represents the elapsed time in seconds and the respective core average boron concentration, showing how the system responds to various operational states.


Question: What is the significance of archaeological resources in the context of the MNGP site?

The results of the literature and map review showed that there are 35 recorded archaeological resources, 9 architectural resources, and 8 cemeteries listed within 6 miles of MNGP.

Photographs taken during and after the construction of the MNGP facility are useful in showing the environmental context during that period.


Question: Licensees are obligated to provide thorough documentation to ensure regulatory compliance.

The NRC requires licensees to submit a series of regulatory filings that ensure compliance with safety standards and operational protocols.

The weather forecast predicts a sunny day with clear skies, making it ideal for outdoor activities.


Question: What is a probabilistic risk assessment (PRA) in the context of nuclear licensing?

In 10 CFR 52.47(a)(27) and 10 CFR 52.79(a)(46), the NRC requires an applicant for a DC, or a COL, respectively, to perform either a design-specific or a plant-specific PRA.

The applicant or licensee should describe existing public and occupational health issues, as appropriate.


Question: The NRC has classified groundwater use conflicts as a Category 2 issue due to the potential for significant impacts at high withdrawal rates.

Impacts to groundwater use during the subsequent period of extended operation are SMALL, which is consistent with the NRC’s conclusion in the 2013 GEIS (Table 4.5-1).

TVA follows BMPs intended to minimize the potential for release of fuels and hazardous materials, as specified in 'A Guide for Environmental Protection and Best Management Practices for Tennessee Valley Authority Construction and Maintenance Activities, Review A, 2022.'


Question: What is the methodology used for extended passive cooling in reactors?

This topical report presents the methodology for the extended passive cooling and reactivity control that is used in the NuScale design to enhance safety and reliability.

NuScale Power, LLC is committed to advancing innovative and scalable solutions for clean energy through the development of small modular reactors.


Question: What does pCi/g measure in nuclear quality control?

The units of measurement used in this report include pCi/g, which refers to picocuries per gram, a standard unit for measuring radioactivity.

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


Question: In the case of a steam generator break, the operator must act quickly to divert auxiliary feedwater to the operating generator within a specified timeframe.

According to the FSAR, with the single failure of the TDAFWP considered, it is assumed that 390 gpm total AFW flow will be delivered to two of the intact SGs at 1 minute after the trip and an additional 195 gpm of AFW flow will be delivered to the third intact SG at 10 minutes after the trip.

A TCOA associated with the accident to have pressurizer PORV available, this TCOA time is 8.6 minutes.


Question: PWR assembly loading pattern

Figure B2-2 shows the schematic of a 37-Assembly PWR Basket Preferential Loading Pattern, detailing the arrangement of assemblies within a pressurized water reactor (PWR).

Refer to Table B2-8 for Maximum Heat Loads.


Question: proprietary information in nuclear submissions

This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by Framatome and not made available to the public.

I am familiar with the criteria applied by Framatome to determine whether certain Framatome information is proprietary.


Question: What are the subsurface conditions at the North Anna site?

The results of the field exploration program indicate that the station area is underlain by three basically different, but related materials. These are, in order of increasing depth: Residual silty fine sand and sandy silt with some gravel (saporite);

Several springs are also found on the site but their discharge rates are dependent on climatic factors.


Question: Ensure that the backup power source is functioning correctly while transferring bus operations.

IF a degraded or undervoltage condition occurs while transferring bus, THEN immediately establish GREATER THAN 3.9 KV using T-4 LTC CONTROL.


Question: The NRC mandates that licensees implement a monitoring plan for fish populations affected by their operations.

The IPEC 2021 Spring Sturgeon Impingement Monitoring report outlines the procedures for evaluating the impacts of facility operations on sturgeon populations in the Hudson River.

The report includes a detailed schedule for sampling activities, ensuring that data is collected at appropriate intervals to accurately assess environmental impacts.


Question: What is a safety-related system in nuclear licensing?

A safety-related system is designed to mitigate potential hazards and ensure the safe operation of a nuclear facility under both normal and abnormal conditions.

The reactor pressure vessel is a critical component that houses the nuclear fuel and allows for controlled fission during operation.


Question: The conditions that require the operator to enter T-101 are specific to the reactor's operational parameters.

The operator will promptly enter T-101 (Reference 14) upon recognition of any of the following entry conditions: 1. RPV water level below +12.5 inches (Level 3) 2. RPV pressure above 1096 psig 3. A reactor scram condition with APRM above 4%.

For the postulated event of Turbine Trip Without Bypass, concurrent with a postulated CCF of PPS, sufficient automatic control functions, indications that are independent of the PPS, and operator actions, are/will be available to mitigate the event.


Question: The regulatory compliance for VCCs includes ensuring they meet safety standards for storage and containment.

The Affected Vertical Concrete Casks (VCCs) must comply with specific regulatory guidelines set forth by the NRC to ensure their safety and reliability.

The reactor operation is closely monitored to assess performance and prevent any potential incidents related to core damage.


Question: activity of nuclear isotopes

The activity of a nuclide is measured in picocuries per gram (pCi/grams) and indicates the rate at which the isotope decays and emits radiation, which is critical for safety assessments and regulatory compliance in nuclear applications.

The table also presents energy values measured in kiloelectronvolts (keV), which indicate the energy of photons emitted during radioactive decay, essential for understanding the radiation spectrum of the radioactive materials.


Question: What is the protocol for managing radiopharmaceutical extravasation?

Figure 1 shows the protocol that is in use in our hospital for management of radiopharmaceutical extravasation. It is based on the findings of this review.

Repetitive gamma camera measurements are performed in case of therapeutic extravasation in consultation with the physicist.


Question: initial conditions of nuclear reactors

Changed initial conditions and resultant classification.

This JPM previously used as RA-4 on 2013 NRC Exam.


Question: Minimum Channels OPERABLE requirement

With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, the contents of the tank(s) may be released to the environment provided that prior to initiating the release: a. At least two independent samples of the tank's contents obtained by two technically qualified members of the facility staff are analyzed, and b. At least two technically qualified members of the Facility Staff independently verify the release rate calculations, and c. At least two technically qualified members of the Facility Staff independently verify the discharge valve lineup; Otherwise, suspend release of radioactive effluents via this pathway.

Effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours during releases via this pathway.


Question: The report includes calculated activities for various nuclides such as Potassium-40 and Lead-212.


Question: The NRC monitors various groundwater levels that may affect safety and regulatory decisions.

Well Completed in the Lower Catahoula Formation is a significant measurement station for assessing water levels.

The table lists various historic dates alongside their corresponding water levels below land surface.


Question: What are the Standard Review Plans for environmental reviews in nuclear power plants?

Standard Review Plans for Environmental Reviews for Nuclear Power Plants provide a framework for evaluating the potential environmental impacts of nuclear power plant operations during the license renewal process.

The annual reports highlight trends in public engagement and stakeholder communication initiatives undertaken by the agency.


Question: The NRC mandates that license holders must comply with safety protocols during operational activities.

The licensees must comply with the licensing requirements set forth by the NRC to ensure compliance with safety regulations and operational guidelines.

A wide range of services is provided to the public by various organizations, which includes medical and industrial support to meet community health needs.


Question: Protective Strategy and Performance Evaluation Program

Provide documents/procedures that describe how the licensee will demonstrate and assess the effectiveness of the onsite physical protection program and protective strategy, to include: any implementing procedure or lower tier tool related to your PEP, any procedures or lower tier tool related to design and implementation of tactical drills and FOF exercise program.

A complete list of security force members that will participate in the exercise as players.


Question: weathered rock zones

The borings drilled as part of the ESP application subsurface investigation program (presented in Section 2.5.4) revealed severely weathered, fractured and jointed intervals in the Zone III-IV and Zone IV rock.

Significant water loss during drilling occurred in two of the fracture zones in boring B-803.


Question: The material's fracture toughness is crucial for ensuring structural integrity in low-temperature conditions.

K_IC FRACTURE TOUGHNESS (KSI - IN^1/2) is critical for assessing the material's performance under stress, especially at low temperatures, as indicated in the table of KID Lower Bound Fracture Toughness A533V.

The table provides various reference temperatures and their corresponding values, essential for the classification of materials used in reactor components.


Question: What are the waste management practices required for nuclear facility operations?

All waste would be properly managed onsite and disposed of at approved treatment or disposal facilities in accordance with applicable regulations. (Section 7.2.3.2)


Question: The procedure outlines the necessary guidelines to prevent unauthorized operations during infrequently performed testing.

This procedure describes the requirements, responsibilities, and administrative controls necessary for conducting infrequently performed tests or evolutions (IPTEs).

Personnel involved in the IPTE must be adequately briefed on the expectations prior to conducting the evolution.


Question: What is the method used for analyzing the solid samples for strontium-90?

The analytical method performed for strontium-90 in solid samples involves the use of the GFPC, Sr90, Solid 'Dry Weight Corrected' technique as outlined in the report.

The certificate of analysis report was dated November 14, 2019, and the client sample ID was L2-011-101-FSGS-005-SS.


Question: delta temperature monitoring

Temperature is sensed by an aspirated dual temperature sensor at the 60-m level, and an aspirated dual temperature sensor at the 10-m level. One-half of the dual sensor at each elevation is used for ambient temperature. The other half of each sensor is used to provide a differential temperature (delta temperature) between the 10-m and 60-m elevation.

The wind direction and wind speed sensors are checked for normal operation, according to vendor specifications.


Question: Does the proposed alteration impact confidence in the ability of safety barriers to contain radiation?

Margin of safety is associated with confidence in the ability of the fission product barriers (i.e., fuel cladding, reactor coolant system pressure boundary, and containment structure) to limit the level of radiation dose to the public.

The revised Emergency Plan will continue to provide the necessary support.


Question: What is wind direction persistence in nuclear safety assessments?

North Anna wind direction persistence data from 9/16/71 to 9/15/72 shows the probability distribution of various wind durations, essential for evaluating the safety implications of airborne releases.

The reactor's cooling system operates under stringent regulatory standards to ensure efficient heat removal during normal and emergency conditions.


Question: dedicated processor modules in nuclear safety systems

Figure 3.5.1.1-2 shows the dedicated processor modules performing the coincidence voting for ECCS and NSSSS only.

Figure 3.5.1.1-1 LCL Segmentation for Reactor Trip (One Division)


Question: low-level waste disposal costs

In the 2019 decommissioning fund status reporting cycle, in which licensees provided decommissioning trust fund data to the NRC by March 31, 2019, as required by 10 CFR 50.75(f), 74 of the 98 operating reactors in the U.S. applied LLW burial cost escalation factors based on the Table 2-1 scenario in which non-compact affiliated generators used the Utah and Texas disposal sites.

Licensees may use the escalation factors from this NUREG in their cost analyses, or they may generate and report site-specific cost estimates that result in a total cost estimate of no less than the amount estimated by using the 10 CFR 50.75(c) formula and cost escalation factors presented in this NUREG.


Question: shutdown cooling system activation

IF ANY of the following exists: - S/G press less than 675 psig, - OR RCS press less than 1870 psig, - OR Cntnt press greater than 1.5 psig THEN ACTUATE SI manually.

4. CHECK SI actuated: a. Any SI annunciator LIT.


Question: The configuration of the non-mixing grid is critical for determining the local heat flux during operation.

The axial geometry involves specific distance measurements from the edge of the heated length to the thermocouples.


Question: alternate AC sources

To establish the AACS, the DG-3 cross-connection to power selected safe shutdown loads is available and an additional AC source, a 480-volt diesel generator (DG-4), is staged and available.

The configuration risk management program required by 10 CFR 50.65(a)(4) are to be put in place to assure that significant risk configurations are avoided during the extended DG inoperability.


Question: lower support plate inspection methods

40 – 60 Methods and Frequency: ASME XI VT-3 for loss of material due to wear, as supplemented by TB 16-4 methods.

Only to confirm that TB 16-4 applies. No issue with the merging of Items W12a and W12b into new Item W12.


Question: loss of power instrumentation

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

**A.** One or more required channels inoperable. | **A.1** Enter the Condition referenced in Table 3.3.8.1-1 for the channel. | Immediately


Question: Cs-137 activity in nuclear safety analysis

Cs-137 was detected in concentrations above MDC in two (2) of the eleven (11) pucks analyzed.

The top half-inch pucks from concrete core samples B1-010-001-CJFC-C04-CV, B1-010-001-CJFC-C09-CV, and B1-010-001-CJFC-C11-CV were sent off-site to GEL Laboratories for gamma spectroscopy and HTD analysis of the full suite of ROC.


Question: steam generator replacement procedures

The inspection should focus on SGRP activities that verify proper restoration of pressure boundaries of the reactor coolant system (RCS), secondary systems, exclusion of foreign materials, and plant modifications that could affect plant risk during subsequent plant operation.

The licensee may install temporary or permanent modifications to support the SGRP well in advance of the SGRP outage.


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

All peaks were identified. M = First peak in a multiplet region  m = Other peak in a multiplet region  F = Fitted singlet  Errors quoted at 2.000 sigma


Question: What is the significance of the energy level of nuclides in nuclear safety?

Nuclear licensees must maintain compliance with the reporting requirements set forth by the NRC.


Question: The U.S. NRC issues Provisional Operating Licenses for the various units of the Indian Point Energy Center, demonstrating the regulatory framework governing their operation.

The U.S. Nuclear Regulatory Commission (NRC, the Commission) licenses for the IPEC are Provisional Operating License No. DPR-5 for IP1, Renewed Facility Operating License Nos. DPR-26 and DPR-64 for IP2 and IP3, respectively, and the general license for the ISFSI.

Operation of IP1 permanently ceased on October 31, 1974 and all fuel was permanently removed from the IP1 reactor vessel by January 1976; operation of IP2 permanently ceased on April 30, 2020 and all fuel was permanently removed from the IP2 reactor vessel on May 12, 2020; and operation of IP3 is scheduled to permanently cease by April 30, 2021.


Question: The NRC establishes requirements for the licensing process of nuclear facilities, ensuring compliance with safety standards.

The regulation of nuclear reactors involves rigorous licensing protocols defined by the NRC to ensure safety and environmental protection.

The Government Publishing Office provides access to various federal documents, including the ability to call toll-free for customer service in Washington, DC.


Question: What is the relative shift (Δσ) in nuclear survey design?

The relative shift (Δσ) for the survey unit data set is defined as shift (Δ), which is the Upper Boundary of the Gray Region (UBGR), or the DCGL (SOF of 1), minus the Lower Bound of the Gray Region (LBGR) (SOF of 0.5), divided by sigma (σ), which is the standard deviation of the data set used for survey design.

A Prospective Power Curve was generated using COMPASS, a software package developed under the sponsorship of the United States Nuclear Regulatory Commission (USNRC) for implementation of the MARSSIM in support of the decommissioning license termination rule (10CFR20, Subpart E).


Question: non-qualified visual inspection techniques

Requirements for developing technical justifications of volumetric examinations, such as ultrasonic testing (UT), of bolts are provided in the inspection standard, MRP-228…

The need for performance of physical measurements applied to a very few number of RVI components for confirming the initial component-specific design dimensions were with a small allowable design range; the physical measurement needs to be formed either just prior to the ILR period or within a short time of entering into the ILR period, with subsequent aging management to be performed using periodic VT-3 inspections of the components that were subject to the baseline physical measurement activities (normally on a 10-year frequency).


Question: What are intra-system common cause factors in nuclear risk assessments?

CCFs have been shown to be important contributors in PRAs. As limited plant-specific data is available, generic common cause factors are commonly used.

The methods used to address HFE dependencies are integrated into the HRA Calculator, which is designed to comply with the American Society of Mechanical Engineers / American Nuclear Society (ASME/ANS) requirements.


Question: nuclear power station avian protection plans

Duke Energy maintains an avian protection plan that addresses every avian incident. The plan includes employee and contractor training, guidance for reducing avian interactions with nuclear power plant infrastructure, procedures for responding to and required reporting of avian incidents to the FWS, and associated corrective actions.

The survey identified five bat species occurring onsite, but they did not include the northern long-eared bat or any of the two State-listed bat species.


Question: climate change impacts on agriculture

Negative impacts on crop & livestock productivity due to increased incidence of weather extremes

Increased vulnerability to ecosystem changes & tree mortality


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

**Peak Locate From Channel:** 100
**Peak Locate To Channel:** 4096
| Peak No. | Energy (keV) | Peak Size (CPS) | Peak CPS (%) | Uncertainty | Peak Type | Tolerance Nuclide |


Question: What is the definition of isotopes in nuclear licensing?

| Isotope (e.g. AM241) | Activity (e.g. 100) | Unit (e.g. mCi) |

Provide information about other devices you have that are subject to registration.


Question: The NRC mandates rigorous assessment of potential aging effects on key components such as valve bodies.

Loss of Material is a significant aging effect requiring management for components like valve bodies primarily made of stainless steel.

One-Time Inspection procedures must be adequately documented to ensure all safety protocols are followed.


Question: What is a Liquid Radioactive Waste Management System?

The Liquid Radioactive Waste Management System is designed to manage and process liquid waste generated by nuclear reactors, ensuring proper containment and treatment before disposal or release.

Effluents containing chemicals or biocides are managed separately from radioactive waste and must adhere to specific environmental regulations for disposal.


Question: The design of the steam generator is critical for the overall performance of the reactor cooling process.

THE ABBREVIATIONS S.G. AND R.C.P. REFER TO THE STEAM GENERATOR AND THE REACTOR COOLANT PUMP RESPECTIVELY.

THE DRAWING ELEVATION -258 FT LINE REPRESENTS THE WALL AT ELEVATION 254 FT.


Question: The confidence level for counting uncertainty is typically expressed as 1.96-sigma.

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

The address for LaCrosseSolutions is S4601 State Hwy 35, Genoa, Wisconsin 54632.


Question: Were the samples designated as hazardous by the client during the shipment process?

Did client designate the samples as to be received as hazardous? - COC notation or relevant slides on containers equal client designation.

All requirements are recorded in Claris.