Question: gas decay tank

The Gas Decay Tank is an integral part of the waste gas system, where radioactive gases are held until the decay occurs to reduce their activity levels before discharge.

The Waste Gas Compressor operates by removing the gas from the reactor coolant system to ensure a continuous and safe operational state.


Question: maximum time in vacuum drying

For the Yankee-MPC configuration, the time duration shall not exceed the time shown for the specified heat loads:


Question: What scenarios necessitate the activation of the Station Alarm as per the specified operational guidelines?

Which one of the following identifies which of these events require sounding the Station Alarm, in accordance with OP-63, Intra-Plant Communications System?


Question: nuclear reactor license application

The application for a nuclear reactor license must include detailed information regarding the reactor's design, safety systems, and operational procedures to ensure compliance with regulatory standards.

The cooling system of the reactor is crucial for maintaining optimal operational temperatures and preventing overheating during energy production.


Question: groundwater withdrawal impact

The NRC concluded that groundwater withdrawals by nuclear power plants would have a SMALL impact on groundwater quality.

In the 2013 license renewal GEIS, the NRC reviewed groundwater drawdown due to water withdrawals, which can draw water into the aquifer.


Question: LCO (Limiting Condition for Operation)

When a system is determined not to 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.

The reactor operates at full power while performing routine maintenance checks on various safety systems.


Question: 10CFR21 compliance in nuclear audits

During review of pictures of 10CFR21 postings at each location it was determined that the Houston and Trainer locations had Rev 7 of Procedure 100-QC-017.1 instead of Rev 8 and some locations had copies of 10CFR21 from 2021 or 2022 while there was a change made March 14, 2023. The 10CFR21 postings were updated during the audit.

Based on this audit, all elements of MISTRAS’s Quality Assurance Program are effectively implemented with exception of the identified Findings.


Question: The NRC's oversight included ensuring compliance with the National Environmental Policy Act during the uranium recovery licensing process.

the NRC Staff had fulfilled its responsibilities under the National Environmental Policy Act (NEPA) to characterize cultural resources at the

The Oglala Sioux Tribe (Tribe) and a group of individuals and organizations referred to as the “Consolidated Petitioners” (together, Petitioners) seek review of the Board’s decision as well as two interlocutory Board orders.


Question: What are the parameters used to calculate storm surge elevation?

The maximum storm surge elevation from wind generated surge was calculated from the cumulative effects of three different parameters: wave runup, wave setup, and wind setup.

Cross Section E-E' is determined to be a critical scenario fetch length and direction.


Question: What is the importance of the Seismic-Fire Interactions Assessment in determining fire risk?

The qualitative assessment of seismic induced fires should not be a source of model uncertainty as it is not expected to provide changes to the quantified FPRA model.

The selected truncation was confirmed to be consistent with the requirements of the PRA Standard.


Question: What is the purpose of the Fuse Holders program in the context of nuclear safety?

VC Summer SLRA section B2.1.43, “Fuse Holders,” states that the VC Summer Fuse Holders program is a new program that, when implemented, will be consistent with NUREG-2191, Section XI.E5, “Fuse Holders.”

The staff reviewed VC Summer AMP for fuse holders in the portal and notes discrepancies between parameters of the VC Summer AMP for fuse holders and those of the NUREG-2191 (GALL-SLR) AMP XI.E5, “Fuse Holders,” as follows:


Question: M5 Hydrogen Pickup Fraction

M5 Hydrogen Pickup Fraction is a key metric used to assess the accumulation of hydrogen in the cladding material of nuclear fuel rods, which can impact safety and performance in reactor conditions.

The distribution of upper bound P-M for Zr-4 hydrogen provides insights into the behavior of zirconium alloys under high-temperature reactor conditions, contributing to the understanding of material integrity.


Question: What is the process for nuclear reactor licensing?

The licensing process for a nuclear reactor involves multiple stages, including pre-application consultations, submitting a formal license application, and undergoing extensive reviews by the Nuclear Regulatory Commission (NRC).

Nuclear reactors operate under specific physical principles that involve complex thermodynamic processes and heat transfer mechanisms.


Question: What is the purpose of operational tests for nuclear plant valves?

TRS 3.7.H.9  Perform valve operational test.

TRS 3.7.H.14 Verify the battery-to-battery and terminal connections are clean, tight, free of corrosion, and coated with anti-corrosion material.


Question: The NRC reviews licensee submittals to ensure that systems relevant to safety are adequately evaluated.

Nuclear Instrumentation is in scope for LR as defined in 10 CFR 54.41.

Warehouse A is not included in the scoping results.


Question: compartment volume in nuclear safety

Compartment Volume: 2.36000E+06 (Cubic feet)

The behavior of nuclear materials is influenced by the type of reactors used, such as pressurized water reactors and boiling water reactors.


Question: What are the Emergency Plan Implementation Procedures (EPIPs) for nuclear plants?

The Shift Manager initially acts in the capacity of the Station Emergency Manager (SEM) and takes actions as outlined in the EPIPs.

The Site Vice President is ultimately responsible for the operation of the Station.


Question: nuclear licensing process

The nuclear licensing process involves several steps, including the submission of safety analyses and the evaluation of environmental impacts, which are critical for ensuring compliance with regulatory standards.

The archaeological findings from the survey included various artifacts, such as pottery, tools, and animal remains that provide insights into the historical use of the land.


Question: The audit report must be submitted within a stipulated time frame after the completion of the review process.

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

Audit Dates: Onsite – October 24-27, 2022


Question: The NRC requires recordkeeping and reporting from facilities that are no longer in operation.

In § 50.71, this final rule revises paragraph (e) to include NPUFs in the requirement and makes a verb tense correction to footnote 1.

In § 50.82, this final rule revises paragraph (b) to standardize terminology by replacing the term 'reactor' with the phrase 'production or utilization facility' and revises paragraph (b)(1) to include testing facilities licensed under § 50.21(c) and holders of a license issued under § 50.22.


Question: What is the process for submitting a construction permit for a nuclear power plant?

The notification of adverse effect findings provides an overview of the construction permit review process for the TerraPower Kemmerer Power Station Unit 1.

The distribution list includes various members from NRC and TerraPower, illustrating the collaborative nature of regulatory filings.


Question: What are the requirements for an independent certifying organization in industrial radiography?

An independent certifying organization shall have a certification program open to nonmembers, as well as members.

Membership shall not be restricted because of race, color, religion, sex, age, national origin or disability.


Question: What is the significance of activity measurements in nuclide identification?

Activity (pCi/grams) is a critical parameter in the Nuclide Identification Report, providing information on the quantity of radioactive material present in samples.

Errors quoted at 2.000 sigma are a standard way of expressing uncertainty in scientific measurements, indicating the confidence level of the data.


Question: The types of nuclides listed in this section provide information critical for assessing radiation safety in nuclear facilities.

The identified nuclides include K-40, PB-212, and BI-214, which are important for understanding their energy outputs and activities.

The reactor is designed to operate efficiently under various temperature and pressure conditions, ensuring optimal performance.


Question: What is the significance of the ALS and LBB provisions in nuclear safety management?

So, it seems like ALS, the difference is ALS does not use the beyond design basis for proof for the existing requirements, just adds to the LBB provision for FFRG.

We've worked on them for decades now to recover from that. Because of.


Question: site-specific target clean-up levels for uranium in soil

Site-Specific Target Clean-up Levels for Uranium in Soil at HF Spiking Station #2 at the Westinghouse Columbia Fuel Fabrication Facility (WCFFF)

Kevin M Harris, PE, Sr. Environmental Engineer


Question: The analysis of wind speed and direction is crucial for safety assessments in nuclear facilities.

| Year | Most Common Wind Angle (Degrees Clockwise from North) | Average Wind Speed (mph) | Resultant Wind Angle (Degrees Clockwise from North) | Resultant Wind Speed (mph) |

The temperature affects the different phases of matter, such as liquids turning into gases.


Question: The NAC-MPC SYSTEM is designed to ensure compliance with applicable regulations during transport, particularly under 10 CFR Part 50.

This program provides a means for evaluating various transport configurations and transport route conditions to ensure that the design basis drop limits are met.

An annual report shall be submitted pursuant to 10 CFR 72.44(d)(3) or 10 CFR 50.36(a).


Question: radiation monitors in nuclear reactors

RM-11 and 12 respond to venting RCS inside containment with normal RCS activity. 1% clad gap would raise RCS gross activity. A small amount of RCS leakage inside containment would cause RM-11 and RM-12 to respond and go into alarm.

Incorrect. Containment high range radiation monitors RM 30 and 31 would not come on scale without a gross release of activity to the Containment atmosphere (see explanation for answer choice A). Plausible because RM-71, 72, 73 and 74 are known to respond to high activity levels in containment (LBLOCA), but in this case, the activity is largely contained within the RCS piping and would not produce the shine necessary for them to respond.


Question: What are the functions of Phase A containment isolation valves?

This Attachment may be used to close individual Phase A valves or entire train of valves as applicable.

The valves listed are controlled using auxiliary switch locations which are designated in the control room.


Question: nuclear power plant safety systems

Safety systems in a nuclear power plant are designed to prevent accidents and mitigate the consequences of any incidents that may occur, comprising components such as reactor containment, emergency core cooling systems, and radiation monitoring equipment.

The financial analysis of the nuclear project indicated an increase in costs due to labor shortages and material price fluctuations, leading to a reassessment of the project's budget and timeline.


Question: What are reactivity manipulations in nuclear safety testing?

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

The NRC requires licensees to analyze abnormal operating conditions as part of their safety assessments.


Question: transfer of control notification

the NRC has completed the initial review of the requested transfer of control notification and no additional information is required at this time.

The public comment period will end on December 17, 2020 and the consent letter may be issued once the comment period ends.


Question: The NRC specifies minimum detection activity levels for different nuclides.

The Minimum Detectable Activity (MDA) for each nuclide serves as a critical threshold that indicates the lowest level of activity that can be reliably distinguished from background noise in regulatory assessments.

The nuclides are tabled with energy levels, yields, and activities detailing their respective properties and safety implications in various contexts.


Question: What is in-situ leach mining in the context of uranium recovery?

But in in-situ leach (“ISL”) mining cases, the use of a substantial quantity of water 'from a source that is reasonably contiguous to either the injection or processing sites' is sufficient to demonstrate an injury in fact.

In addition to standing, for an intervention petition to be granted, petitioners must demonstrate that they have submitted at least one admissible contention.


Question: harvesting materials for regulatory assessments

For example, legacy materials (e.g., fabrication methods, composition) that are no longer available, but may play an important role in a potential degradation mechanism, would have a higher priority than harvesting materials that can be obtained from other sources.

Degradation mechanisms that are harder to replicate with simulated aging conditions would be of higher priority for harvesting.


Question: What is the significance of heavy load drop analysis in nuclear safety?

The component drop analysis should be reviewed to verify, in general, that the potential offsite releases at the exclusion area boundary are within 10 CFR Part 100, 'Reactor site criteria,' limits and that equipment to maintain safe shutdown will be unaffected.

Training for the large PRP work force should be geared toward the specifics of pressurizer removal/replacement for both the radiation workers and radiation protection personnel.


Question: The report provides the confidence levels for the activities of various nuclides based on their identification and analysis.

### Notes
- * = Energy line found in the spectrum.
- ? = Manually added nuclide.
- @ = Energy line not used for Weighted Mean Activity.
- Energy Tolerance = 1.000 keV.
- Nuclide confidence index threshold = 0.30.
- Errors quoted at 2.000 sigma.


Question: nuclear regulatory process

Detailed summary of the 11/21/2024 public meeting on Section 206 is available in the NRC’s Agencywide Documents Access and Management System (ADAMS) at **ML24345A048**.

Includes links to all presentation slides and meeting transcript.


Question: Containment HVAC system impact

Subsequently, a ØB Containment Isolation occurs which affects the Containment HVAC and Pressure Control System.

A Unit 1 is at 100% power, indicating optimal operational conditions for the reactor.


Question: maximum flood levels

Table 3.14-1 summarizes the comparison of current and reevaluated maximum flood levels for all possible flood-causing mechanisms.

The simulated peak flow velocities varied between 2.87 feet per second (fps) and 5.02 fps, which are not expected to produce any erosion hazards, as described in Section 2.1.


Question: What are the operator actions required in response to conditions affecting Essential Raw Cooling Water?

Given a set of plant conditions, DESCRIBE operator actions required in response to the following per 0-AOI-13, Loss of Essential Raw Cooling Water: - g. Loss of flow on ALL ERCW supply headers.

The NRC emphasizes the importance of maintaining clear communication between departments to enhance operational efficiency.


Question: What is a recirculation spray system in nuclear safety?

The Simplified Recirculation Spray System includes components such as strainers and pumps that play a critical role in coolant recirculation during safety scenarios.

Line from Queen Start Pump and New Line from Cooling are indicative of general piping arrangements without direct relevance to safety functions.


Question: What is the significance of assembly average burnup in nuclear reactors?

For pressurized water reactors (PWR), the assembly average burnup must remain within the specified limits to ensure the effective performance and safety of the fuel assemblies during operation.

The minimum cooling time for a nuclear reactor is critical to ensure that the fuel is sufficiently cooled before it can be safely handled or processed.


Question: The NRC oversees the establishment of agreements with States to manage radiation control effectively.

The U.S. Nuclear Regulatory Commission (NRC) is authorized under Section 274 of the Atomic Energy Act of 1954, as amended, to enter into agreements with States—

The typical sequence for NRC in processing a State's request for an agreement is summarized as follows:


Question: reactor protection system

the Reactor Protection System (RPS) was initiated by the Main Turbine Loss of Load trip, resulting in an automatic reactor trip from 100 percent power.

An engineering analysis performed to determine the failure mechanism for the stationary failed field pole electrical connection identified cyclic fatigue as the cause for the failure.


Question: The minimum detectable activity for each nuclide is essential for accurately assessing radiological safety.

The minimum detectable activity (MDA) for various nuclides is crucial in determining the sensitivity of measurements associated with nuclear safety assessments.

The sample geometry is described as B1-010-001-FJFC-D47-GM and refers to the physical configuration used in detecting gamma radiation.


Question: The receiving system must identify unrecognized messages and handle them according to established guidelines.

Only predefined data sets should be used by the receiving system. Unrecognized messages and data should be identified and dispositioned by the receiving system in accordance with the pre-specified design requirements. Data from unrecognized messages must not be used within the safety logic executed by the safety function processor. Message format and protocol should be pre-determined. Every message should have the same message field structure and sequence, including message identification.

HSL: The topical report disposition directly applies to the PPS application without modification. <br> IRIG: The MTP is the communication processor in this arrangement whereas the AC160 is the function processor. The MTP performs the handshaking for this communication.


Question: What are the requirements for loading damaged fuel can locations?

Damaged fuel located in a damaged fuel can location in the damaged fuel basket must have a minimum burnup of 5 GWd/MTU, a maximum enrichment of 4.05 wt. % ²³⁵U, and a minimum cool time of 15 years.

HFRAs are only allowed for Westinghouse assemblies and may have a maximum exposure of 4.0 GWd/MTU and must have a minimum cool time of 16 years.


Question: The ambient noise levels near Unit 1 are typically low due to sparse population density and rangeland activities.

Most of the existing ambient noise in the vicinity is generated from passenger vehicle and truck traffic along Navajo 11, wind, and livestock-related activities. It is estimated that background noise levels generally are less than 50 dB [56 dBA].

Potential noise receptors (occupied dwellings) within and 3.2 km [2 mi] surrounding the Crownpoint project area are depicted on Figure 3.1-4.


Question: VCT level modulation signal

At 73%, 1LT112 sends a signal to 1CV112A to modulate open. Since that didn't occur, VCT level will continue to rise until 1LT185 reaches 95% at which time 1CV112A will fully open. Thus, VCT level will start to lower.

Using the flowchart, this question: - Can be answered solely by knowing “systems knowledge”, i.e., how the system works, flowpath, logic, component location. YES


Question: dissolved lead 210 regulatory limits

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

ACZ Laboratories, Inc. 2773 Downhill Drive  Steamboat Springs, CO  80487  (800) 334-5493


Question: nuclear reactor head fabrication requirements

Verify that material heat treatment which is used to enhance the mechanical properties of RV head material carbon, low alloy, and high alloy chromium (Series 4XX) steels is conducted in accordance with the ASME Code and approved vendor procedures or instructions and are consistent with the applicable ASME Code, Section III, "Rules for Construction of Nuclear Facility Components," requirements.

Security considerations associated with vital and protected area barriers that may be affected during replacement activities.


Question: HDI is expected to meet the NRC’s standards for site restoration and minimize potential environmental impacts.

The PSDAR states, however, that HDI will decommission Indian Point to meet the NRC's unrestricted release criteria, and New York does not point to any requirement that an applicant additionally discuss state law standards that might apply to certain site restoration activities.

In conclusion, we find that these arguments do not provide a basis for admitting Contention 2.


Question: What are the RCS cooldown rate limits?

The normal cooldown rate is 100°F/hour. The reason is vessel integrity.

Void formation is a concern while performing a natural circulation in E-0.2 and is avoided if possible.


Question: What is included in the energy released into containment during a pipe break?

The energy released into containment by a pipe break is energy that is (a) initially contained in the primary and secondary coolant systems fluids and the metal components of the system boundaries and the sensible heat stored in the core, plus (b) that additional energy that is produced and released subsequent to the break as a result of continued fission, fission product and actinide decay and metal-water reaction.

High MFW temperature is assumed in order to maximize the heat source effect of the SG.


Question: PB-214 nuclide characteristics


Question: Operators can start the containment spray system from the control room by using two switches simultaneously to prevent accidental activation.

The containment spray actuation signal starts the containment spray pumps and aligns the discharge of the pumps to the containment spray nozzle headers in the upper levels of containment.

Water is initially drawn from the RWST by the containment spray pumps.


Question: pressure relief valve testing requirements

The proposed alternative does not alter the requirement to test Class 1 pressure relief valves at least once every 5 years beginning from the date of the as-left set pressure test for each valve.

In addition, the proposed alternative does not alter the requirement of subparagraph I-1320(b)(2) to test valves removed from service within 12 months of removal from the system.


Question: The regulatory process requires careful consideration of enrichment levels of uranium in reactor fuel.

Minimum Initial Assembly Avg. Enrichment wt % 235U (E) refers to the percentage of the isotope uranium-235 contained within the fuel assemblies at the start of the reactor operation.

The amendments to the Certificate of Compliance No. 1031 are designed to ensure that the regulations keep pace with technological advancements.


Question: What are the guidelines for Risk-Informed Technical Specifications related to external hazards?

Topical Report NEI 06-09, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines, Revision 0-A (hereafter referred to as NEI 06-09-A), (Reference 1), as clarified by the United States Nuclear Regulatory Commission (USNRC) final safety evaluation (Reference 2), requires that the license amendment request (LAR) provide a justification for exclusion of risk sources from the Probabilistic Risk Assessment (PRA) model based on their insignificance to the calculation of configuration risk, and to discuss conservative analyses applied to the configuration risk calculation.

No new site-specific or plant-unique external hazards were identified through this review.


Question: What are the requirements for the chlorine detection system in nuclear facilities?

Two independent chlorine detection systems, with their alarm/trip setpoints adjusted to actuate at a chlorine concentration of less than or equal to 2 ppm, shall be OPERABLE.

Two independent broad range gas detection systems shall be OPERABLE with their alarm/trip setpoints adjusted to actuate at the lowest achievable Immediately Dangerous to Life or Health gas concentration level of detectable toxic gases providing reliable operation.


Question: Licensees are obligated to provide decommissioning financial assurance for significant quantities of radionuclides.

Possessing an unsealed radionuclide in a quantity more than 1,000 times the Appendix B value requires decommissioning financial assurance;

DFPs require a comprehensive decommissioning cost estimate and updates every three years;


Question: Auxiliary Feedwater Pump temperature alarm

Identifies Input 593 for Aux FWP PPC Temperature Alarm and goes to Section 2.2.

Observes PK09-17, Motor Driven Aux FW PP alarming, and informs the Shift Foreman.


Question: inadequate core cooling instrumentation

Requires in part instrumentation to provide unambiguous, and easy-to-interpret indication of inadequate core cooling (ICC). The PPS modification does not impact the existing post accident monitoring instrumentation including reactor water level instrumentation.

The LGS is a Boiling Water Reactor and does not have a pressurizer.


Question: The NRC mandates thorough evaluations of thermal hydraulic characteristics following modifications in reactor components.

Reference 1 addressed the basic thermal hydraulic effects of the replacement of depleted fuel rods with stainless steel insert rods.

The addition of mixing vanes on six or seven mid-grids enhances the performance of the fuel assembly.


Question: INTERMEDIATE RANGE MONITOR operation

Per QCOP 0700-02, INTERMEDIATE RANGE MONITOR (IRM) OPERATION step E.1: IRM Rod Block setpoints are as follows: a. IRM downscale allowable value: (1) 5/215 of full scale. (2) Bypassed when on Range 1. b. IRM high allowable value: < 112/125 of full scale. c. IRM detector "NOT FULL IN" in MODE 2 (Startup). d. IRM inoperable.

The test ID is associated with the 2023 NRC ILT RO/SRO EXAM submissions, which involves various regulatory evaluations and assessments.


Question: The implementation of emergency procedures requires adherence to specific guidelines set forth by the organization.

The Shift Manager (SM) has implemented 02-OHP-4025.001.001, Emergency Remote Shutdown procedure.

The NRC 2018-INP02 outlines the process for performing local isolation operations on different reactor units.


Question: The NRC evaluates the reliability of safety systems to determine if they are operational under various conditions.

As expected, given the small sensitivity of the model to the change, the technical specifications are unaffected by this sensitivity.

Table E9-17 summarizes the results of these select RICT calculations.


Question: What is the regulatory limit for keff in nuclear fuel storage?

The criticality analysis for the storage of BWR assemblies in the DNPS SFP racks with Boral panels has been performed. The results for the normal condition show that keff is ≤ 0.95 with the storage racks fully loaded with fuel of the highest anticipated reactivity, at a temperature corresponding to the highest reactivity. The results for all accident conditions also show that keff is ≤ 0.95 with the storage racks fully loaded with fuel of the highest anticipated reactivity, at a temperature corresponding to the highest reactivity.

The NFV at DNPS is a reinforced concrete Class I structure that ensures the spacing is maintained even in a seismic event.


Question: The NRC evaluates the thermal performance of fuel rods through various licensing requirements.

The Thermal Performance of fuel rods is a critical aspect that the NRC takes into account when evaluating safety systems and compliance with regulatory standards.

The procurement process for safety equipment often involves thorough inspections and certification to ensure compliance with quality standards.


Question: What happens to the Instrument Air and Scram Air systems when a filter failure occurs?

Knowledge of the effect that a loss or malfunction of the following will have on the INSTRUMENT AIR SYSTEM: Filters

The plant is operating at 100% power with the following: - Instrument Air Dryer 39AD-4A is in service. - Instrument Air Dryer 39AD-4B is out of service for maintenance.


Question: cities within 50-mile radius of nuclear plants

Table 3.11-1 shows the cities, villages, and towns located totally or partially within a 50-mile radius of the DNPS.

The 2010 Census data indicates diverse population changes in various regions across the state.


Question: external hazards screening

A calculation 'Other External Hazards Screening' was developed to evaluate the applicability for other external hazards. All the external hazards that could potentially affect the WCGS site were screened out based on the criteria specified in SRs EXT-B1 or EXT-C1 of the ASME/ANS PRA Standard.

See Table 4-4 below. This table uses screening criteria that is described in the ASME/ANS PRA Standard 2009.


Question: concrete mechanical testing in nuclear applications

Mechanical testing (compressive, tensile, modulus of elasticity) is part of the scope for the concrete harvested from structures that are in close proximity to the reactor pressure vessel (RPV).

The timeline included initial discussions in 2014, harvesting in 2015, and testing from 2016 to 2018.


Question: What are the guidelines for core damage frequency (CDF) according to RG 1.174?

RG 1.174 does not establish firm limits for total CDF and LERF, but it recommends that risk-informed applications be implemented only when the total plant risk is no more than about 1E-4/year for CDF and 1E-5/year for LERF.

The purpose of this enclosure is to demonstrate that the Arkansas Nuclear One, Unit 1 (ANO-1) total CDF and total LERF are below the guidelines established in RG 1.174.


Question: Written examination outlines for nuclear operators

Updated forms 4.1-BWR, 4.1-PWR and 4.1-AP1000 for use with the current K/A catalogs and GFE re-integration

Form ES-401-3 Generic Knowledge and Abilities Outline (Tier 3) has been renamed Form 4.1-COMMON, Common Examination Outline and is for all vendor type Tier 3 and Tier 4 K/As.


Question: The aging management review assesses cumulative fatigue damage in components exposed to reactor coolant over extended periods.

Consistent with NUREG-1801. Cumulative fatigue damage is an aging effect assessed by a fatigue TLAA in Section 4.3. Further evaluation is documented in Section 3.1.2.2.1.


Question: The tornado missile hazard assessment can be performed using conservative models to ensure safety.

Although the average maintenance tornado missile CDF is <1E-6/year, the CDF due to tornado missiles for certain maintenance or LCO configurations was determined to be above 1E-6/year, requiring tornado missile risk to be accounted for during RICT configurations.

The CDF cutsets were further evaluated using the Advanced Cutset Upper Bound Estimator (ACUBE); CDF for Unit 1 reduced to approximately 8.3E-7/year, about a 5% reduction.


Question: The classification of emergencies at RBS includes various protocols for notification and activation during incidents.

Once an emergency classification has been declared at RBS, the Control Room will notify the Parish warning points.

The sequences of anticipated notifications and activation of emergency response personnel for each emergency classification are provided below.


Question: The NRC mandates that the reactor coolant system be monitored to ensure it operates under optimal safety conditions.

c. Reactor coolant system (RCS) pressurizer pressure.


Question: nuclear reactor regulatory processes

The regulatory process for nuclear reactors involves a series of approvals and assessments to ensure safety protocols are followed before licensing any new reactor technology.

Yamir Diaz-Castillo is a member of the Quality Assurance & Vendor Inspection Branch within the Office of Nuclear Reactor Regulation.


Question: Source Range Monitor System

Ability to (a) predict the impacts of the following on the SOURCE RANGE MONITOR (SRM) SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: upscale and downscale trips.

The examination outline includes various levels and tier numbers which are crucial for understanding the technical proficiency needed for the licensed operator.


Question: The NRC mandates assessments for seismic stability in structures within licensed nuclear facilities.

The licensing process for nuclear facilities requires a comprehensive evaluation of the site-specific seismic hazards to ensure the safety of operations.

The Mississippi embayment exhibits significant geological features that influence local engineering practices.


Question: temporary shielding in radiation work

Perform engineering load analysis in advance - Control installation and removal by written procedure - Inspect after installation - Conduct periodic radiation surveys - Minimize damage caused by heavy lead temporary shielding - Balance radiation exposure received in installation against exposure to be saved by installation - Shield components with abnormally high radiation level early in the maintenance period - Shield travel routes - Shield specific work locations - Use full-scale models (mock-up) to plan temporary shielding design and installation - Perform directional surveys to improve shielding design by locating radiation source

Plan access to/exit from work areas - Provide for service lines (air, welding, ventilation, etc.) - Provide appropriate communication (includes closed-circuit television) - Remove sources of radiation - Minimize discomfort of workers - Work in lowest radiation levels - State requirements for standard tools - Consider special tools.


Question: off-hours unannounced ERO report-in drill

Each NextEra site will conduct an off-hours unannounced ERO report-in drill at least once within an eight-year cycle.

The NextEra ERO notification is an all-call process.


Question: What is the impact of a complete loss of forced reactor coolant flow?

Complete loss of forced reactor coolant flow can lead to overheating of the reactor core due to a lack of adequate cooling.

The document also discusses various maintenance schedules required for reactor components to ensure operational efficiency.


Question: bedrock aquifers and their recharge

Bedrock aquifers within the 50-mi radius of the site are presented. Most of the glacial drift in the study area is underlain by Pennsylvanian bedrock that consists largely of shale and siltstone interbedded with limestone, sandstone, underclay, and coal.

Groundwater in the glacial drift is discharged to streams that intersect the aquifers (base flow), to the underlying glacial drift, to the Pennsylvania bedrock, and to pumping wells.


Question: ideal pedestal machining

Figure 2 illustrates the differences between correct and incorrect pedestal machining, highlighting the importance of the radius in the machining process.

Photo 2 shows the attempted bolt installation, which does not relate to the machining process discussed in Figure 2.


Question: tritium measurement in nuclear regulatory filings

ZBG002D  | 3/13/2024  | AIR TEMPERATURE                     |         |         | REG       | 9.000     | degC     | REG         |


Question: The procedures governing the lifting of the reactor vessel head include specific limits on load height and the material beneath the load.

In response to the September 14, 2007 NEI "Industry Initiative on Heavy Loads Lifts," McGuire Procedures used to control the lift and replacement of the reactor vessel head establish limits of load height and the medium present under the load.

The reactor building polar crane and fuel handling bridges were designed in accordance with CMAA-70 and Chapter 2-1 of ANSI B30.2-1976.


Question: instrumentation and control systems

With respect to the instrumentation and control systems review, the phased startup operations application supplement focuses on disabling inputs for various systems (such as HIPS equipment).

How will it be ensured that crane operators are trained to reliably implement any specific administrative controls associated with crane operations during continued construction due to the phased startup approach?


Question: spent nuclear fuel storage

Onsite storage of spent nuclear fuel                                   | 1          | Applicable     | 4.11.2

Microbiological hazards to the public (plants with cooling ponds or canals or cooling towers that discharge to a river) | 2          | Applicable     | 4.9.4


Question: What does PMF refer to in nuclear plant evaluations?

PMF Peak Flow Rate (1) is evaluated at 600,000 cfs, associated with an event centered above Brownville, NE.

The Gavins Point Dam is located near Yankton, South Dakota, at River Mile 811.


Question: What are the components and functions of the High Pressure Coolant Injection System?

The following aging effects associated with the High Pressure Coolant Injection System components require management: - Cracking - Cumulative fatigue damage - Hardening or loss of strength - Long-term loss of material - Loss of material - Loss of preload - Loss of material due to selective leaching - Reduction of heat transfer - Wall thinning due to flow-accelerated corrosion

The materials of construction for the Residual Heat Removal System components are: - Aluminum - Cast austenitic stainless steel - Copper alloy - Copper alloy (>15% Zn or >8% Al) - Gray cast iron - High-strength Steel - Metallic.


Question: What are the Pasquill-Turner stability classes?

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

The table displays the number of inversion occurrences recorded across various years, detailing how frequency varied over time.


Question: The NRC mandates licensees to perform an assessment of hazards that could affect the safety of nuclear power plants.

NRC Regulatory Guide (RG) 1.78, "Evaluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release," Revision 1, (ADAMS Accession No. ML013100014), December 2001.

Entergy letter to NRC, "Flooding Hazard Re-evaluation Report - Required Response for Near-Term Task Force (NTTF) Recommendation 2.1," (OCAN091602), (ADAMS Accession No. ML1620A006), dated September 14, 2016.


Question: What is the significance of Radium-226 in water quality monitoring?

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


Question: What are potential stressors tricolored bats face from nuclear power plants?

The potential stressors that tricolored bats could experience from operation of a nuclear power plant (generally) are as follows: - mortality or injury from collisions with nuclear power plant structures and vehicles - habitat loss, degradation, disturbance, or fragmentation, and associated effects - behavioral changes resulting from refurbishment or other site activities

The contribution of Oconee Station operations to climate change-related effects on monarch butterflies would be too small to be meaningfully measured, detected, or evaluated.


Question: What is the procedure for qualifying candidate Safety Instrumented Systems (SIS) in nuclear applications?

Candidate SIS not identical in construction to the parent SIS may be qualified by extension through appropriate analysis and/or testing. The procedure for candidate SIS qualification requires a high degree of similarity to ensure that the mechanical strength, stiffness, and critical design tolerances of the candidate SIS favorably compare with the qualified parent SIS.

The cooling system for the reactor is designed to maintain optimal temperatures and ensure the safety of the plant under normal and abnormal conditions.


Question: The assessment of whether a party has offered an admissible contention is crucial for establishing participation in regulatory processes.

Discretionary intervention is allowed only where at least one petitioner has established standing and offered an admissible contention, such that a hearing will be held.

Courts have also rejected such claims as a basis for standing because the injury is not “concrete and particularized.”


Question: technical specifications for loss of ventilation chill water

Post-examination analysis revealed three generic weaknesses associated with applicant performance on the written examination. The applicants displayed weaknesses associated with actions on a loss of ventilation chill water, technical specifications associated with an inoperable station service water valve, and procedure selection during a natural circulation cooldown.


Question: What are the actions required when facing both offsite circuit inoperability and inoperable diesel generators?

For TS 3.8.1.D, one required offsite circuit inoperable and one required DG inoperable, RMAs may include: Actions to increase risk awareness and control...

Implement rigorous communications and coordination between site operations and transmission grid operators to ensure the activities rendering the offsite circuit inoperable are progressing as expected and if not, additional resources are applied to reduce the duration of the activity.


Question: The locations for planned samples are documented based on their room assignments and any necessary relocations.

Four (4) planned locations (P#) were relocated due to being colocated with discretionary (D#) locations.

- There is no discretionary location #3 in Room 27.


Question: What is a check valve flapper in nuclear systems?

(1) CHECK VALVE FLAPPER HAS BEEN REMOVED

The Main Steam System is divided into three separate trains, A, B, and C, each with specific components and configurations.


Question: What are the pathways for containment sump water in nuclear safety?

1. **Pathway 1:**  Containment Sump Water to Auxiliary Building

The plant has four compartments with varying configurations and isolation mechanisms.


Question: Nuclide Minimum Detectable Activity

The Minimum Detectable Activity (MDA) for a nuclide is defined as the lowest level of activity that can be reliably detected, calculated using statistical methods and taking into account the background counts, efficiency of the measurement system, and other relevant factors.

The results of the analysis indicate a variety of peak sizes and energy levels, which are essential for identifying specific isotopes in a sample.


Question: radiation protection procedures

policymakers could also draw some technical and safety standards from NRC’s legacy radiological standards, including general radiation protection procedures and low-level waste disposal at 10 C.F.R. Part 20 and portions of 10 C.F.R. Part 30 for site decommissioning plans.

Given the results of some analyses of coal power plants’ emissions, perhaps coal plants should be subject to radioactive emissions standards as well.


Question: What is the significance of pCi/L in nuclear monitoring?

Bars indicate precision for Thorium 230 and Lead 210 analytical results.


Question: What are the requirements for submitting a nuclear license application?

The Nuclear Regulatory Commission (NRC) requires applicants to submit detailed technical and safety analyses that demonstrate compliance with the regulations.

Tables were compiled to provide a summary of the data collected during the safety assessment process.


Question: NuFuels will ensure that monitor wells are placed outside the areas that have been mined previously to maintain safety standards.

NuFuels will verify that the production area monitor wells will be outside the locations of workings by superimposing their surveyed locations on existing surveyed maps which illustrate the working locations.

The geology of an individual production area will be evaluated in conjunction with well field development to assure proper placement.


Question: What are the requirements for austenitic stainless steel materials in nuclear systems?

It is required that all austenitic stainless steel materials used in the fabrication, installation, and testing of NSSS components and systems be handled, protected, stored, and cleaned according to recognized and accepted methods and techniques.

The calculated stresses at operating speed are based on stresses due to centrifugal forces.


Question: What does groundwater stability monitoring involve in nuclear licensing?

This report describes the results associated with the groundwater stability monitoring plan established by Amendment #56.

One hardcopy of the report is being provided to you and a digital format version (PDF) is being uploaded to the NRC Document Control application.


Question: lead-210 concentration regulatory limits

Precision bar values may exceed the scale.


Question: SRO-only question determination

Does the question require one or more of the following: assessment of plant conditions (normal, abnormal, or emergency) and then selection of a procedure or section of a procedure to mitigate or recover, or with which to proceed?

Can the question be answered solely by knowing immediate operator actions?


Question: The licensee must ensure robust defenses against cyber threats to protect digital safety systems.

10 CFR 73.54, Protection of Digital Computer and Communication Systems and Networks, requires licensees to develop cyber security plans and programs to protect critical digital assets, including digital safety systems, from malicious cyber-attacks.

The safety system cabinets are provided with door alarms.


Question: The first response in an emergency at the facility begins with the on-shift personnel who are trained to address the situation promptly.

The first line of control in an emergency at Surry Power Station lies with the onshift personnel. The shift complement is staffed with personnel qualified to take the initial actions necessary to respond to an emergency.

This notification and information are available to the Shift Manager in the Control Room to enable a timely classification of the emergency and subsequent actions.


Question: What is the role of the Control Room Ventilation System in nuclear safety?

The Control Room Ventilation System is designed to isolate and maintain the design conditions within the control room during fires in either the control room or outside the emergency zone. 10 CFR 54.4(a)(3)

The UFSAR references include sections 6.4.2, 6.4.3, and 9.4.1, which provide detailed descriptions of various safety systems and their functionalities.


Question: ITAAC notification process

An ITAAC (Inspection, Tests, Analyses, and Acceptance Criteria) notification is a key regulatory tool that documents the completion of necessary inspections and analyses as established in a licensing process.

The report contains many technical details regarding the construction sequence and the safety measures during the reactor design phase.


Question: What precautions should be taken when using the Residual Heat Removal System?

ENSURE 0-DOOR-410-ABD2, U2 RB Equipment Access Door is CLOSED prior to placing RHR in service for shutdown cooling.

Temporary Fabric Door, 0-DOOR-410-R003, is not a qualified ABSCE boundary with U1 in Mode 4 and RHR in service for shutdown cooling.


Question: What type of reactor is planned for the Kemmerer Power Station Unit 1?

The facility is planned to be a pool-type reactor that uses liquid sodium as the coolant instead of water.

The U.S. Department of Energy (DOE) Office of Clean Energy Demonstrations (OCED) serves as a cooperating agency for the environmental review (ML23213A147), and the NRC is the lead Federal agency for the environmental review.


Question: What are the requirements for medical examinations in relation to radiation protection?

Frequency of Medical Examinations for Use of Respiratory Protection Equipment

Low-Level Waste Shipment Manifest Information and Reporting is crucial for tracking radioactive waste effectively.


Question: The Environmental Impact Statement (EIS) can be accessed through the NRC's public database known as ADAMS.

The FEIS is available through the ADAMS Public Documents collection at [https://www.nrc.gov/reading-rm/adams.html](https://www.nrc.gov/reading-rm/adams.html) under ML250313A307.

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](mailto:Jean.Trefethen@nrc.gov).


Question: The NRC assesses potential deterioration in structural support materials, especially under varying environmental conditions.

Cracking, Loss of Bond, and Loss of Material (Spalling, Scaling) are aging effects requiring management for reinforced concrete used as a flood barrier in the reactor building.

Increase in Porosity and Permeability, Loss of Strength are conditions noted for missile barriers under water flowing environments.


Question: What are microreactors in nuclear regulation?

Another activity that's going on right now is microreactors, which are receiving a lot of attention.

Siting is usually one of those issues that's discussed a lot in terms of microreactors.


Question: nuclear device registration

NRC Device Key: 719644 (Internal Control Number)


Question: strontium-90 analysis

Emergency planning is crucial for ensuring that all personnel are prepared for potential accidents.


Question: Yankee Class Fuel Assembly specifications

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

- As specified in Table B2-2 for the applicable fuel assembly type.


Question: internal flooding risks

The GGNS internal events PRA model includes evaluation of risk from internal flooding events.

The terrain surrounding the immediate vicinity of the plant is such that landslide is highly unlikely.


Question: What are the safety considerations for Radium 226 in regulatory filings?

Radium 226, dissolved is analyzed using the EPA 903.1 method for assessing the safety and regulation of this radioactive material in various samples.

The ACZ Laboratories, Inc. is located at 2773 Downhill Drive Steamboat Springs, CO 80487, providing analytical services across a range of environmental samples.


Question: nuclear power plant safety systems

Safety systems play a crucial role in ensuring the reliable operation of nuclear power plants and the protection of public health and safety.

The density of water varies depending on temperature and pressure, which influences its behavior in different experimental conditions.


Question: What are common cause failures in nuclear safety systems?

Does the methodology described in draft NEI 20-07 require an assessment of potential common cause failure (CCF) vulnerabilities in a proposed system, prior to implementation of this methodology?

The present regulatory infrastructure for HSSSR systems acknowledges that it is possible to identify a potential CCF vulnerability due to a latent defect has such a low likelihood of occurrence that it may be treated as “beyond design basis”, and therefore its consequences may be evaluated using best-estimate methods.


Question: It is essential to provide protective clothing and prevent unauthorized access to the site before cleanup operations.

Issue protective clothing, and secure site from unauthorized entry.

Calibrated beta, gamma, alpha survey meter is essential for detecting radiation levels present in the area.


Question: What are the alarm indicators for a tornado warning in a nuclear plant?

Annunciators 912-5 C-1, RX BLD 1 LOW DP and 912-5 C-4, RX BLD 2 LOW DP are in alarm.

Unit 1 and Unit 2 are operating at rated power when Security reports a tornado sighted in the Protected Area.


Question: source term calculations for reactor accidents

The concentration of nuclides in the primary and secondary system prior to and following the LRA are determined as follows: A. The iodine activity in the reactor coolant prior to the accident is based upon an iodine spike which has raised the reactor coolant concentration to 60 µCi/g of dose equivalent (DE) I-131.

The atmospheric dispersion factors used in the analysis were calculated based on the onsite meteorological measurement programs described in subsection 3.3.


Question: What are the design requirements under 10 CFR 53.440 for advanced reactors?

The design requirements outlined in 10CFR53.440 ensure that advanced reactors meet safety and operational standards established by the NRC.

The comment on FR Doc # 2024-23434 relates to general procedural aspects of submittals but does not delve into specific technical requirements.


Question: Access Authorization program requirements

10 CFR 73.55(b)(7) requires licensees maintain an Access Authorization program in accordance with 10 CFR 73.56. Paragraphs 10 CFR 73.56(m) and (o) require licensees ensure the confidentiality and integrity of Access Authorization (AA) system data.

This would be an adverse impact to access control functions.


Question: What are the immediate steps the RO must take following a reactor trip?

RO trips the reactor and performs immediate actions of EOP-TRIP-1.


Question: ultimate heat sink operational requirements

**LCO 3.7.9** The UHS shall be OPERABLE.

**APPLICABILITY:** MODES 1, 2, 3, and 4.


Question: reactor coolant system operability assessment

Evaluation of the operability of the reactor coolant system (RCS) following questions regarding leakage from valve 2RC003A, as documented in IR 4810722 during the week ending October 26, 2024.

Review and evaluation of the risk associated with 2CV8385B and winter weather conditions during the week ending December 21, 2024.


Question: What are the requirements for obtaining a nuclear reactor license?

To obtain a license for a nuclear reactor, applicants must submit a detailed safety analysis report that demonstrates compliance with regulatory requirements as established by the NRC.

The cooling system of a nuclear reactor plays a crucial role in maintaining safe operational temperatures and preventing overheating.


Question: Intermediate Range Monitor (IRM) operational requirements

Per QCOP 0700-02, INTERMEDIATE RANGE MONITOR (IRM) OPERATION step E.1: IRM Rod Block setpoints are as follows: a. IRM downscale allowable value: (1) 5/125 of full scale. (2) Bypassed when on Range 1. b. IRM high allowable value: < 112/125 of full scale. c. IRM detector "NOT FULL IN" while in MODE 2 (Startup). d. IRM Inoperable.

The IRM INOP condition is generated by placing the Function switch out of the OPERATE position, unplugging any of the modules in the IRM Drawer, or the IRM detector voltage is low. This is bypassed only when the Mode Switch is in RUN.


Question: The diagrams for flow and valve operation pertaining to the Safety Injection System in Unit 1 illustrate key operational parameters.

11715-FM-096A Flow/Valve Operating Numbers Diagram: Safety Injection System, Unit 1

The referenced drawings are not part of the UFSAR. This is not intended to be a complete listing of all Station Drawings referenced from this section of the UFSAR.


Question: A508 Cl. 2 material specification

The closure head dome is made of A533B, Cl. 1 material specification, while the closure head ring and flange are specified as A508 Cl. 2 according to the project requirements.

The upper shell of the reactor is constructed with A508 Cl. 2, featuring a material specification that ensures structural integrity under operating conditions.


Question: What are low-level waste compacts?

The figure and table below identify the composition of all LLW compacts as of May 2014.

Alaska and Hawaii belong to Northwest Compact.


Question: What is the purpose of the Turbine Bypass System in a nuclear reactor?

The Turbine Bypass System is designed to provide an artificial Reactor Coolant System load by providing the following steam dump capabilities: 1. 100 percent full load steam dump to atmosphere via main steam safety valves. 2. 40 percent full load steam dump to condenser via pressure reducing control valves.

Proper operation of the Turbine Gland Sealing System is verified during unit startup. Normal operating system performance monitoring will detect any deterioration in the performance of system components and will be corrected by appropriate means as necessary.


Question: soil sampling in nuclear inspections

**Soil samples** - Develop a sample plan that includes exact location coordinates, and the pipe segments in the area.

The inspection requirements for these planned periodic inspections should include certain elements to ensure compliance and safety.


Question: EPA 200.7 testing standards


Question: equilibrium temperature in nuclear safety

The highest equilibrium temperature was found to be 84.2°F for July 1993, therefore the original equilibrium temperature of 88°F for June 1952 remains bounding.

The total heat rejected (see Figure 9-41 and Table 9-16 for the 30 day period) comes from three basic sources: Supporting information can be found in Ref 15.


Question: shutdown cooling system

At Time = 40 minutes, Shutdown Cooling (SDC) is restored.

A plant shutdown and cooldown is in progress with the following: - RHR pump B is running in a Shutdown Cooling lineup.


Question: site characterization in groundwater modeling

Approximately 3 months after Kickoff Meeting - First hybrid technical meeting (in-person and virtual participation) will be held for the purpose of discussing site characterization, model input parameters, and model boundary conditions.

HMC to provide the groundwater model and model files to the agencies.


Question: The Environmental Impact Statement must consider all public feedback before finalization.

Supplemental Environmental Impact Statement is a detailed analysis that evaluates the potential environmental effects of a proposed action and must be considered carefully.

The Public Scoping Meeting addressed various community concerns regarding infrastructure development in the region.


Question: What is the significance of Plutonium-238 in radiochemistry?

Plutonium-238 is a radioactive isotope that is commonly used in radioisotope thermoelectric generators (RTGs) to provide power for spacecraft and various other applications.

The QC summary provided lists the batch information, sample qualities, and unit measurements relevant to radiation detection.


Question: What is a condensate cleanup system in nuclear plants?

The condensate cleanup system processes the full flow of condensate and has five condensate filter demineralizer vessels with 25-percent capacity each (one on standby), five holding pumps, piping, instrumentation, and controls.

Codes and standards applicable to the condensate cleanup system are listed in table 3.2.2-1.


Question: climatological changes and nuclear plant impacts

Overall – Are expected climatological changes likely to alter the impacts of plant activities on local facilities and residency patterns, including housing, public schools, recreational resources, and first-responder agencies? Justify your answer in 4 sentences or less. Expected climatological changes are not likely to alter these impacts.

By the middle of the next century, southwestern Michigan is expected to face increased rainfall and flood risks, which may place pressure on regional transportation infrastructure.


Question: The NRC has emphasized that safety culture traits should be integrated into a facility's procedures but are not subject to enforcement by the NRC.

Individuals and organizations performing regulated activities bear the primary responsibility for safety and security. The performance of individuals and organizations can be monitored and trended and, therefore, may be used to determine compliance with requirements and commitments and may serve as an indicator of possible problem areas in an organization’s safety culture.

In 2003, Commissioner Jeffrey Merrifield proposed a rule for oversight of a safety conscious work environment.


Question: What is the role of the Public Information Officer in emergency response procedures?

The Public Information Officer (PIO) is responsible for disseminating information to the public during an emergency situation, ensuring clear communication between the agency and the community.

The Director of the Office of Homeland Security and Emergency Preparedness oversees the revisions to procedures and ensures compliance with federal regulations.


Question: What is the relationship between the initial assembly enrichment levels and fuel burnup in PWR reactors?

| Initial Assembly Avg. Enrichment wt % 235U (E) | WE 14x14 Assembly Average Burnup (B) GWd/MTU |

The parameters of the safety analysis must account for various potential operational scenarios.


Question: The analysis confirmed the presence of specific nuclides through statistical methods that gauge their activity levels.

Activity measurements are taken in units of pCi/M² to quantify radioactive emissions from various nuclides in a sample.


Question: What are the ASME OM Code requirements referenced in 10 CFR 50.55a?

All other ASME OM Code requirements as incorporated by reference in 10 CFR 50.55a for which relief or an alternative was not specifically requested, and granted or authorized (as appropriate), in the subject request remain applicable.

If you have any questions, please contact the Project Manager, Robert Kuntz, at 301-415-3733 or by e-mail to Robert.Kuntz@nrc.gov.


Question: What is the process for appealing a decision made by the Atomic Safety and Licensing Board?

In accordance with the provisions of 10 C.F.R. § 2.311, as this memorandum and order rules upon an intervention petition, any appeal to the Commission from this memorandum and order must be taken within 25 days after this issuance is served.

The September 27, 2021 hearing request of petitioners Beyond Nuclear, Inc., and the Sierra Club, Inc., and the accompanying 10 C.F.R. § 2.335 waiver request are denied and this proceeding is terminated.


Question: What is the purpose of radiological surveys in nuclear facilities?

Radiological surveys are conducted to assess the radiological conditions of a facility and ensure that the levels of radiation exposure are within acceptable limits.

The instrument data collected on various models provides insightful information regarding the calibration and performance of the measurement probes used in monitoring equipment.


Question: The NRC specifies criteria for the effective performance of emergency cooling systems in light-water reactors.

Code of Federal Regulations, Title 10, Part 50, Section 46, “Acceptance Criteria For Emergency Core Cooling Systems For Light-Water Nuclear Power Reactors,” August 2007.

EMF-2328(P)(A), Revision 0, *PWR Small Break LOCA Evaluation Model, S-RELAP5 Based*, March 2001.


Question: What actions has Xcel Energy taken regarding the Monticello facility's historic and cultural significance?

While preparing its application for subsequent license renewal, Xcel Energy contacted the Minnesota State Historic Preservation Office (MN SHPO) by letter dated March 17, 2022, seeking input on the potential effects from continued operation of Monticello on historic and cultural resources.

The FWS published a proposed rule to list the tricolored bat (*Perimyotis subflavus*) as endangered under the Endangered Species Act on September 14, 2022 (87 FR 56381). This species range includes both Wright and Sherburne Counties.


Question: The North Anna Power Station relies on several nearby hydrological features for its operations.

The North Anna Power Station is a nuclear facility located near the North Anna River and Pamunkey Creek, utilizing these water sources for cooling and other operational needs.

Louisa is a small unincorporated community that is part of Louisa County, Virginia, and serves as a rural locality.


Question: automatic reopening protection in relief valves

ERV B and C are the only valves with auto reopening protection.

QCOP 0203-01, step F.2 directs alternating use of B and C if repeated cycling of relief valves is required.


Question: management of aging effects in reactor coolant systems

The aging management program for the Expansion Joint includes monitoring for Loss of Material and Cracking due to exposure to Treated Borated Water and Reactor Coolant environments, following guidelines from Water Chemistry and ASME Section XI.

The reactor coolant system is critical for maintaining safe operating conditions, ensuring the removal of heat from the reactor core.


Question: What is the function of the Reactor Core Isolation Cooling (RCIC) system?

The Reactor Core Isolation Cooling (RCIC) system is designed to provide core cooling and maintain the integrity of the reactor core during an emergency by using steam from the reactor vessel to drive a turbine, which in turn supplies coolant to the reactor.

The table discusses various isolation valves and their corresponding attributes for different systems, detailing their locations and failure probabilities in containment isolation scenarios.


Question: TCE plumes in groundwater monitoring

TCE was detected in groundwater at concentrations at or above its MCL of 5.0 µg/L from 10 of the 86 surficial aquifer monitoring wells. There are four TCE plumes at CFFF in the surficial aquifer. These CVOC plumes are generally referenced as the Western Groundwater AOC, the main plume, the eastern plume, and a plume near the southwest corner of the plant building.

The main PCE plume in the surficial aquifer – lower zone appears to emanate from an area between WL2 and the plant building and migrates to the southwest. It is currently unknown whether the Western Groundwater AOC PCE plume is the result of groundwater migration from the developed area of the facility or another source.


Question: The NRC recently updated the terms of the Renewed Facility License for the Ohio State University Research Reactor, indicating it now has no expiration date.

By order dated February 6, 2025, the U.S. Nuclear Regulatory Commission (NRC) modified Renewed Facility License No. R-75 for the Ohio State University Research Reactor such that it has no fixed term.

If you have any questions, please contact me at (301) 415-4002 or via email at Jessica.Lovett@nrc.gov.


Question: What are the inspection requirements for PWR pressurizer head welds?

The inspection requirements for PWR pressurizer head welds include visual examinations, and in certain cases, non-destructive testing such as ultrasonic examinations to detect subsurface defects.

The report outlines various parameters that influence the thermal performance of reactors, including coolant flow rates and reactor pressure maintenance.


Question: NELAC certification requirements

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

All samples and sub-samples associated with this project will be disposed of after November 01, 2024.


Question: What is the average burnup for fuel assemblies in nuclear reactors?

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

Certification of Compliance No. 1031 B2-69 Amendment No. 6, Revision No. 2


Question: lubricating oil analysis

Consistent with NUREG-1801. Loss of material of stainless steel piping, piping components, and tanks exposed to lubricating oil in the TGAS is managed by the DCPP Lubricating Oil Analysis AMP (B.2.3.25). The DCPP One-Time Inspection AMP (B.2.3.19) will verify the effectiveness of the DCPP Lubricating Oil Analysis AMP (B.2.3.25) to manage loss of material.

Reduction of heat transfer of aluminum heat exchanger tubes exposed to lubricating oil is managed by the DCPP Lubricating Oil Analysis AMP (B.2.3.25). The DCPP One-Time Inspection AMP (B.2.3.19) will verify the effectiveness of the DCPP Lubricating Oil Analysis AMP (B.2.3.25) to manage reduction of heat transfer.


Question: benthic invertebrates near nuclear sites

Table 3.7-2 provides a list of benthic invertebrates in the vicinity of the DNPS Site, including various species such as Dragonfly (Gomphus sp.) and several types of beetles and caddisflies.

The safety evaluation in this report identifies several potential hazards associated with the operation of the nuclear facility, including seismic events and mechanical failures.


Question: What is the flowrate measurement in nuclear safety procedures?

Applicant performs calculation and verifies flowrate to be 77.48 gpm.


Question: lightning flash frequency calculation

Hence, the expected frequency of lightning flashes in the exclusion area per year is 31, as calculated below: 9.4 flashes (km²) (year) x 3.3 km² = 31 flashes year

Approximately 65 percent of Illinois tornadoes have occurred during the months of March through June, with the highest statewide probability of a tornado occurrence being in April.


Question: Nickel-63 analysis in nuclear materials

Nickel-63 is identified in the analysis results, specifically measured in pCi/g for the Rad Liquid Scintillation Analysis parameters.

The following Analytical Methods were performed: DOE EML HASL-300, Am-05-RC Modified, ASTM C 1475-00 Modified, DOE EML HASL-300, Pu-11-RC Modified, and others.


Question: What is the 50.59 process in nuclear regulation?

the 50.59 process involves evaluating whether a proposed change to a nuclear facility requires prior NRC approval or can be implemented without a license amendment.

The comments were provided by NEI members representing the 50.59, 72.48, and PRA communities.


Question: What is the purpose of the Residual Heat Removal system in nuclear reactors?

Residual Heat Removal is crucial for maintaining safe reactor conditions by transferring excess heat from the reactor core to prevent overheating after shutdown.

Grounds are established to ensure the electrical distribution system functions correctly under both normal and fault conditions.


Question: The NRC mandates specific designs for the safety baskets in boiling water reactors to ensure optimal performance.

Figure B2-5 presents the schematic of the BWR 81-Assembly Basket, which is essential for the containment and organization of nuclear fuel assemblies.

The arrangement of different materials in a reactor core affects the thermal efficiency greatly, especially under high-pressure conditions.


Question: What is the Minimum Detectable Concentration in nuclear analysis?

Strontium Carrier: 92.8, Acceptable Limits: (25%-125%)


Question: What is the dry weight basis in analytical reports?

The samples in this SDG were analyzed on a "dry weight" basis.

Sample 1204336298 (L1-010-107-AJGS-018-SSDUP) was recounted due to a peak shift.


Question: nuclear regulatory licensing process

The licensing process for nuclear reactors involves a thorough review of safety, environmental impacts, and operational protocols to ensure compliance with established regulations.

The efficiency of the turbine generator is a crucial factor in the overall performance of the power plant, influencing both energy output and operational costs.


Question: What is the activity of CS137 in nuclear measurements?

The activity of CS137 is measured at 100 mCi.

The distributor license number for Thermo Process Instruments is L03524.


Question: What is average burnup in nuclear fuel?

The average burnup for nuclear fuel is measured in gigawatt-days per metric ton of uranium (GWd/MTU).

Safety assessments are a critical component of the licensing process, involving detailed evaluations of all potential hazards and risks associated with a facility.


Question: What statistical method was used to evaluate survey designs without a background reference area?

The Sign Test was selected as the non-parametric statistical test. The use of the Sign Test did not require the selection or use of a background reference area, which simplified survey design and implementation.

The survey design specified fourteen (14) soil samples for non-parametric statistical testing.


Question: What is the dam breach flood flow and stage hydrograph?

Dam Breach Flood Flow and Stage Hydrograph at the VEGP Site illustrates the relationship between the stage of the flow and the flow rate measured in cubic feet per second (cfs) during the time period observed.

River flow is measured in cubic feet per second and can vary greatly depending on upstream conditions and rainfall.


Question: What are the strategies for maintaining injection during long-term station blackout scenarios?

The Dresden FLEX strategies provide a means of maintaining injection in long term SBO scenarios.

A: AC/DC, B: Core Cooling, C: FW/Condensate, D: HVAC, E: Instrument Air/Nitrogen, F: Containment Phenomena, G: Containment Bypass, H: ATWS, I: Internal Flooding, J: Seismic, K: Fire, L: Other, M: Cooling Water, N: Seal Cooling.


Question: NRC reasonable accommodation policy

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

The policy statement may be found on the NRC website: http://www.nrc.gov/reading-rm/doc-collections/commission/policy/67fr36920.html


Question: SEMS definition in nuclear licensing

For purposes of this paragraph as applied to this license, SEMS means any SEMS that has been referenced in a staff SER, TER, EA, or EIS and supplements and amendments thereof.

The licensee shall maintain records of any changes made pursuant to this condition until license termination.


Question: The NRC emphasizes the importance of managing age-related degradation in reactor components to ensure safety and performance.

Loss of material due to pitting, crevice corrosion is an aging effect requiring management for Flow Device (Class 1) components made of stainless steel that operate in treated water above 60°C.

Closed-cycle cooling water (External) is critical for maintaining optimal performance in a reactor's heat exchanger systems, but its environmental conditions must be monitored.


Question: environmental justice in regulatory filings

The overall socioeconomic environment of the area has not changed appreciably, and the factors that affect environmental justice in addition to those listed in Section 3.10 can be summarized as follows (IHS 2019).

Diseases of the heart, malignant neoplasm, unintentional injuries, and diabetes are the leading causes of American Indian and Alaska Native deaths (2009-2011).


Question: nuclear licensing regulatory process

The nuclear licensing process involves a thorough review of safety analyses and regulatory compliance, thereby ensuring that the design, construction, and operation of the nuclear facility meet the required safety standards and regulations set forth by the NRC.

Mr. Michael Franovich, NRR, is responsible for overseeing the quality assurance program within the regulatory body, ensuring that all protocols are followed and maintained in accordance with established standards.


Question: What is the significance of uranium mined lands in nuclear licensing?


Question: The flow rates reported in our tests account for the conditions unique to the CT Test, with particular emphasis on the head loss measurements.

The clean screen head loss for the CT Test was 0.03 psi.

Figure 3.f.4-12 and Figure 3.f.4-13 show plots of raw head loss test data for the CT Test with time to demonstrate the key testing activities.


Question: What are the statistics regarding housing availability in the Region of Influence?

Table 3-20 lists the total number of occupied and vacant housing units, vacancy rates, and median values in the two-county ROI.

The Pickens County School District comprises 23 public schools, with approximately 16,400 students for the 2020–2021 school year.


Question: What is online monitoring technology for nuclear plant pressure transmitters?

Online Monitoring Technology to Extend Calibration Intervals of Nuclear Plant Pressure Transmitters refers to the methods implemented to reduce the frequency of calibration needs while ensuring the accuracy and reliability of pressure measurements in nuclear plants.

The NRC documents detail various safety assessments and technical specifications that outline the necessary operational procedures to ensure public safety during routine inspections.


Question: NUREG 1556 series guidance

Please prepare your licensing requests in accordance with the most recent Final version of NUREG 1556 Series Guidance at: http://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr1556/

Please only submit one complete, physically and legibly signed copy of your correspondence to us.


Question: status of a reactor operator's license

The applicant determines RO B’s license is INACTIVE

Either 12-hour or 8-hour shifts is the normal.


Question: radioisotope activity measurement

The radionuclide activity is measured in microcuries per gram of dry sample (µCi/g Dry) to ensure accurate quantification of radioactive materials during regulatory filings.

Measurements of uncertainty are crucial in any scientific analysis, but their implications vary depending on the context and type of data collected.


Question: The assembly average enrichment must meet specific requirements for fuel assemblies used in pressurized water reactors.

Minimum Initial Assembly Avg. Enrichment wt % 235U (E) 42 < Assembly Average Burnup ≤ 43 GWd/MTU

The safety features of the reactor are designed to prevent overheating and ensure containment integrity during an incident.


Question: Thermo-Lag 330-1 fire barrier material

The Thermo-Lag 330-1 fire barrier material was assumed to be comprised of 90% (by volume) particulate and 10% fiber.

See the Response to 3.c.1 for the material and bulk densities of the various types of debris.


Question: The NRC has established guidelines for the comprehensive safety analysis of reactor operations.

The Division of Operating Reactor Licensing oversees the licensing process of commercial nuclear power plants, ensuring compliance with federal regulations and safety standards.

The report outlines the various approaches taken by engineering teams to enhance project management efficiency and streamline communication.


Question: Combined License implementation schedule

Two license conditions require SNC to submit, no later than twelve months after issuance of the COL, the following: - A schedule for implementation of the operational programs listed in the Final Safety Analysis Report Table 13.4-201, including the associated estimated date for initial loading of fuel (COL section 2.D.(11))

The U.S. Nuclear Regulatory Commission (NRC) issued Combined Licenses (COLs) NPF-91 and NPF-92 to Southern Nuclear Operating Company (SNC) pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Section 52.97, "Issuance of combined licenses," on February 10, 2012.


Question: What is the significance of Rn=Ru in nuclear reactor safety?

Rn=Ru represents the ratio of the design load to the allowable load for structural components in nuclear facilities, guiding safety evaluations and regulatory compliance.

The NRC's regulatory framework includes various processes for ensuring that nuclear plants adhere to safety standards and operational guidelines.


Question: What is the maximum allowable wellhead injection pressure?

The casing test pressure will vary with the maximum allowed injection pressure as described below. In all cases the well will be sealed, filled with water, and pressured up with air to at least 125% of the maximum allowable wellhead injection pressure (MAWHIP).

In all cases the well will be sealed, filled with water, and pressured up with air to at least 125% of the maximum allowable wellhead injection pressure (MAWHIP). The MAWHIP will be determined as in Section 6.5.3, and posted and monitored as described in Section 6.6 to ensure that the formation fracture pressure is not exceeded.


Question: maximum fuel element cladding temperature

PRM-50-93 requests that the Nuclear Regulatory Commission (“NRC”) make new regulations: to require that the calculated maximum fuel element cladding temperature not exceed a limit based on data from multi-rod (assembly) severe fuel damage experiments;

In the event of a LB LOCA, there would be variable reflowed rates throughout the core; however, at times, local reflowed rates could be approximately one inch per second or lower.


Question: HYSPLIT meteorological data

HYSPLIT is a model used to simulate the dispersion of aerosol concentrations and is crucial for analyzing release locations in nuclear safety assessments.

Output Control Parameters are essential for adjusting the simulations to meet specific operational needs.


Question: peak discharge for PMF analysis

The governing PMF peak discharge of 1,173,128 cfs resulted in a peak stage on the Chattahoochee River at Plant Farley of ___ ft NAVD88 (___ ft NGVD29).

Hydrologic Failure refers to the failure modes associated with significant changes in water flow, often from extreme weather events.


Question: Plutonium 239 analysis


Question: How do the calculations of radial power profiles differ between GALILEO and FRAPCON models?

Figure 5: Comparison of GALILEO and FRAPCON calculated radial power profiles at an enrichment of 3%, Gd2O3 concentration of 8wt% and 60 GWd/MTU burnup

The construction of the containment structure is designed to withstand seismic events and ensure the safety of the reactor.


Question: What is the significance of irradiated gemstone testing requirements?

Lightbox is presently unable to perform its requirement to test no less than 2% of irradiated gemstones prior to distribution.

As a matter of urgency, we have contacted a number of suppliers in the US within various NRC Agreement States that have the capability to test our diamonds.


Question: What is the process for submitting a License Amendment Request in nuclear regulation?

**Wolf Creek Letter ET 21-0005** (Reference 56) submitted a License Amendment Request for a risk informed resolution to GSI-191.

The following list summarizes the key correspondences issued by the NRC or submitted to the NRC by Wolf Creek regarding the resolution of GL 2004-02 up to 2021.


Question: What is the role of Probabilistic Risk Assessment (PRA) in nuclear regulatory activities?

The U.S. Nuclear Regulatory Commission (NRC) discusses the use of Probabilistic Risk Assessment methods in nuclear regulatory activities, as outlined in their final policy statement published in the Federal Register, which emphasizes the importance of these methods in evaluating risks associated with nuclear plants.

The U.S. Code of Federal Regulations specifies the licensing process for production and utilization facilities, providing guidelines for domestic licensing without reference to specific risk assessment methodologies.


Question: reactor coolant system

The reactor coolant system (RCS) is responsible for transferring heat from the reactor core to the steam generators and regulating the temperature and pressure of the reactor.

The emergency response plan must be updated annually to ensure that all personnel are familiar with the procedures in case of an incident.


Question: What are the analytical procedures for nuclear sample analysis?

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.

The samples in this SDG were analyzed on an "dry weight" basis.


Question: loss of forced reactor coolant flow

Figure 15-57 illustrates the reactor power response during a complete loss of forced reactor coolant flow, showing the fraction of initial reactor power over time, from 1.0 at t=0 seconds to 0.01 at t=10 seconds.

The reactor design must comply with the specified regulatory requirements that ensure safety and environmental protection throughout its operational lifetime.


Question: The construction of the Mihama 1 plant, which is a 2-loop design, was completed in 44 months, culminating in first power production in August 1970.

Mihama 1, a 2-loop, 320-MWe unit, is owned and operated by the Kansai Electric Power Company. Mihama 1 marks the beginning of a line of Westinghouse pressurized water reactors supplying the generation needs of the Far East.

Power was supplied to the system on December 2, 1969.


Question: acceptance criteria for RVI components

Acceptance Criteria: Section 5 of MRP-227, Revision 1-A (as supplemented), which includes Table 5-1 for B&W-designed RVIs, Table 5-2 for CE-designed RVIs, and Table 5-3 for Westinghouse-designed RVIs, or MRP-227 (as supplemented) provides the specific examination and flaw evaluation acceptance criteria for the “Primary” and “Expansion” RVI component examination methods.

The program applies applicable fracture toughness properties, including reductions for thermal aging or neutron embrittlement, in the flaw evaluations of the components in cases where cracking is detected in an RVI component and is extensive enough to warrant a supplemental flaw growth or flaw tolerance evaluation.


Question: What are the variable parameters used in the Excel Model Builder?

The Excel Model Builder generates cases 1, 2, 3, etc., for various combinations of variables. These variables are chosen by the user for each case (shaded cells) and the program calculates the model inputs.

The variable parameters include run time, source length, source type, source top elevation, source half length in cm, source area, and the file name.


Question: The RICT Program requires prior management approval for entry during planned activities and requires prompt action during emergent situations.

The procedures for the RICT program will address the following attributes consistent with NEI 06-09: - Plant management positions with authority to approve entry into the RICT Program.

This training will be conducted for the following DTE staff positions.


Question: licensed operator responsibilities

Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, “no-solo” operation, and maintenance of active license status, 10 CFR part 55

Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.


Question: What are the guidelines for operator licensing examinations in nuclear plants?

Nuclear Regulatory Commission (NRC) examiners conducted an initial examination in accordance with the guidelines in Revision 11 of NUREG-1021, "Operator Licensing Examination Standards for Power Reactors."

Eight applicants were issued licenses commensurate with the level of examination administered on September 30, 2020.


Question: What is the Prandtl number and how is it relevant in heat transfer in nuclear systems?

Prandtl number, as seen in Eq. (130) (repeated below as Eq. (163)), only depends on fluid properties that are solely temperature-dependent.

For characterization of forced convective heat transfer in a packed bed, following appropriate heat transfer guidance from Section 4.2, Reynolds and Prandtl numbers must be matched between the prototype and scaled experiment.


Question: high moisture content in instrument air

Excessive moisture can result in desiccant carryover causing high differential pressures across the after filter (1/2, 1A and U-2 IAC only).


Question: What is required for emergency communications in nuclear regulatory frameworks?

Systems are established for prompt communication among principal emergency response organizations.

Adequate equipment is maintained to support emergency response.


Question: What are the regulatory requirements for subsequent license renewal of a nuclear plant?

Xcel Energy has prepared Table 1.1-1 to verify conformance with regulatory requirements. Table 1.1-1 indicates the sections in the MNGP subsequent license renewal (SLR) ER that respond to each requirement of 10 CFR 51.53(c) and 10 CFR 51.45.

The subsequent renewed license for MNGP would expire at midnight on September 8, 2050.


Question: The organizational structure for emergency response includes key roles like the Station Emergency Manager and the Radiological Assessment Director.

### Radiological Assessment Director
- RP Supervisor

#### Dose Assessment Team Leader*
- Offsite Monitoring Team 1
- Offsite Monitoring Team 2

### Corporate Response Manager

### Station Emergency Manager

- Emergency Procedure Coordinator

#### Emergency Maintenance Director
- OSC Director


Question: Zero Trust architecture

Zero Trust provides a collection of concepts and ideas designed to minimize uncertainty in enforcing accurate, least privilege per-request access decisions in information systems and services in the face of a network viewed as compromised.

All data sources and computing services are considered resources.


Question: The High Pressure Coolant Injection (HPCI) system must be declared inoperable when its operational capability is compromised.

Declare High Pressure Coolant Injection (HPCI) System inoperable.

In accordance with the Risk Informed Completion Time Program.


Question: The NRC mandates that applicants thoroughly evaluate design interfaces related to safety systems in their licensing submissions.

The design interfaces between the Natrium Nuclear Island and Energy Island are critical for ensuring regulatory compliance and operational safety.

The report includes various administrative details and a list of recipients involved in the correspondence.


Question: What is required for nuclear training programs?

Possession Requires Specific Permission from Nuclear Training.

Nuclear power plants must maintain a high level of safety through continuous monitoring and assessment of their operational performance.


Question: What is the significance of grid spacing in nuclear reactor safety analysis?

Definition of grid spacing and distance-to-grid

Framatome Inc. is a leading company in the nuclear energy sector.


Question: single failure criterion in nuclear safety systems

The single-failure criterion applies to the safety systems whether control is by automatic or manual means. IEEE Std 379-1988 provides guidance on the application of the single-failure criterion.

The PPS actuating division arrangement is the same arrangement as in the LGS design basis.


Question: What is the GALILEO fission gas release assessment database?

GALILEO steady state fission gas release assessment database and model predictions (from response to RAI 24)

The NRC requires licensees to maintain a comprehensive safety program that ensures compliance with facility operations.


Question: The evaluation of aging effects on components such as reactor vessel nozzles is crucial to ensure structural integrity.

Loss of material due to general, pitting, crevice corrosion


Question: nuclide identification in regulatory filings


Question: nuclear reactor licensing process

The licensing process for a nuclear reactor involves multiple stages, including the review of construction permits, operating licenses, and amendments to the existing licenses depending on modifications to the facility or changes in regulation.

The cooling system of a nuclear reactor is designed to remove heat from the reactor core and transfer it to the secondary system, ensuring the reactor operates within safe thermal limits.


Question: The Component Cooling Water System is crucial for ensuring that heat exchangers operate within safe temperature ranges during reactor operations.

The Component Cooling Water (CCS) system provides essential cooling for the reactor heat exchangers, supporting safe operational standards.

Emergency protocols must be established and regularly updated to reflect any new safety findings and technological advancements.


Question: The NRC issues reports based on their inspections that can include findings related to nonconformance with safety regulations.

A licensee-identified violation which was determined to be of very low safety significance is documented in this report.

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


Question: What is the purpose of temperature transmitters in cooling water systems?

The recirculated cooling water heat exchangers have temperature test points on the cooling water lines after the equipment. Downstream of each heat exchanger there is a temperature transmitter that sends a signal to modulate the amount of condenser circulating water going through the heat exchangers.

An alarm is sent to the operator in the event a low level is reached in the recirculated cooling water storage tank.


Question: What are the operational requirements for MSIVs during reactor modes?

The MSIVs must be OPERABLE in MODE 1, and in MODES 2 and 3 except when closed and de-activated (vented or prevented from opening), when there is significant mass and energy in the RCS and steam generators.

The 8 hour Completion Time is greater than that normally allowed for containment isolation valves because the MSIVs are valves that isolate a closed system penetrating containment.


Question: The regulatory oversight by the NRC includes assessing the safety and licensing processes for nuclear facilities.

The NRC reviews the applications for licenses based on the safety analysis and technical assessments provided by the facility operators.

Periodic maintenance checks are essential for the operational integrity of the plant's cooling systems.


Question: What is the Risk-Informed Completion Time (RICT) program in nuclear regulations?

The RICT Program procedures and processes already implemented at other CEG sites will be used at Dresden Nuclear Power Station (DNPS), Units 2 and 3, where applicable.

The Operations Department (licensed operators) is responsible for compliance with the Technical Specifications (TS) and will be responsible for implementation of RICTs and RMAs.


Question: What are containment isolation valves in nuclear safety?

Specifically, the LAR requested a revision to the action statements associated with Technical Specification (TS) Limiting Condition for Operation (LCO) 3.6.4, "Containment Isolation Valves," to replace the term "valve" with the term "barrier" to encompass all components providing the containment isolation function and specify that actions to address an inoperable containment isolation valve apply to the affected penetration flow path only, rather than all flow paths associated with the penetration.

By letter dated April 30, 2020 (Agencywide Documents Access and Management System Accession No. ML20121A185), Dominion Energy South Carolina, Inc., (DESC) submitted a license amendment request (LAR) to modify Technical Specifications related to the containment isolation valves for Virgil C. Summer Nuclear Station (VCSNS), Unit No. 1.


Question: What is the function of fission product barriers in concrete structures?

Concrete (accessible areas): all Fission Product Barrier, Flood Barrier, Shielding, Shelter and Protection, Structural Support

Cracking due to expansion from reaction with aggregates


Question: identified nuclides in nuclear safety reports

The Nuclide Library Used: C:\Canberra\Apex\Root\Dairyland_NPPL\Library\HOTLAB.NLB


Question: What is the role of inspectors in nuclear facility licensing?

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

On November 21, 2024, the inspectors presented the Operational Safety inspection results to Roy Stutts and other members of the licensee staff.


Question: Condensate Pump operation verification

Verifies proper operation of the Feed Water Reg Valves.

As the EO dispatched to Bus 13, wait 1 minute, then report: 'There is an overcurrent target up on Bus 13 cubicle 8, for the 1B Cond/Cond Booster Pump. No other targets are up on any of the compartments.'


Question: TSTF-505 regulatory amendment

In order to adopt TSTF-505, Revision 2, PG&E is proposing to restructure its plant-specific TS 3.5.2 RA for Condition A to be consistent with TSTF-505, Revision 2 by deleting the Note regarding different RAs for planned and unplanned activities, deleting RAs A.2.1, A.2.2, and A.2.3, and applying the RICT Program to the RA A.1 72-hour Completion Time.

RA A.2.1 is no longer required since the extended CT of 14 days in RA A.2.3 is deleted.


Question: proposed amendment application

WCNOC requests approval of the proposed amendment within 12 months following acceptance. Once approved, the amendment will be implemented within 180 days from the date of issuance.

This letter contains no commitments. If you should have any questions regarding this submittal, please contact me at (620) 364-8831 x8687, or Dustin Hamman at (620) 364-4204.


Question: What role does Antek Instruments play in regulatory compliance for nuclear technology?

Antek Instruments, Inc., is known for producing a range of measurement and analysis devices that contribute to safety and compliance in nuclear applications.

Bard Brachytherapy, Inc. offers various products utilized in cancer treatment, including brachytherapy sources and delivery systems.


Question: What effect does a malfunction in the Reactor Manual Control System (RMCS) have on fuel handling operations during refueling?

Knowledge of the effect that a loss or malfunction of the following will have on the FUEL HANDLING EQUIPMENT: Reactor manual control system: Plant-Specific

The typical maintenance schedule for nuclear reactors includes regular inspections and repairs to ensure all systems function properly.


Question: The radioactive isotopes listed in the document include Tc-99 and Ru-106, which are used in medical applications.

The table contains numerous values that are necessary for the analysis of reactor performance.


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

The assembly average enrichment weight percent of uranium-235 (E) is crucial for determining the fuel’s performance and safety characteristics during operation.

The loading table indicates minimum cooling times required based on different assembly average burnup levels, which impacts how the fuel must be handled after discharge.


Question: average burnup for PWR fuel

| WE 14x14 Assembly Average Burnup (B) GWd/MTU | 57 < B | 58 < B | 59 < B |

The table lists various ranges of average enrichment for different assembly specifications.