ML23166A013
| ML23166A013 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 06/14/2023 |
| From: | Thomas Wengert NRC/NRR/DORL/LPL4 |
| To: | Reid M Entergy Operations |
| Wengert T | |
| References | |
| L-2022-LLA-0197 | |
| Download: ML23166A013 (11) | |
Text
From:
Thomas Wengert Sent:
Wednesday, June 14, 2023 11:12 AM To:
REID, MARK Cc:
Keele Jr, Riley D; Jennifer Dixon-Herrity; Thomas Byrd
Subject:
Arkansas Nuclear One, Unit 1 - TSTF-505 Audit Questions (EPID L-2022-LLA-0197)
Attachments:
ANO-1 TSTF-505 Audit Questions 06142023.pdf By letter dated December 22, 2022 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML22356A249), Entergy Operations, Inc. (the licensee), submitted a license amendment request (LAR) for Arkansas Nuclear One, Unit 1 (ANO-1). The proposed amendment would modify ANO-1 Technical Specification requirements to permit the use of risk-informed completion times in accordance with Technical Specifications Task Force (TSTF)
Traveler TSTF-505, Provide Risk-Informed Extended Completion Times - RITSTF [Risk-Informed TSTF] Initiative 4b, Revision 2.
On May 10, 2023 (ML23121A301), the NRC staff issued an audit plan that conveyed the staffs intent to conduct a regulatory audit to support its review of the subject licensing action. In the audit plan, the NRC staff requested an electronic portal setup and provided a list of documents to be added to the portal. The audit plan also indicated that the NRC may request information and audit meetings/interviews throughout the audit period.
The NRC staff has performed an initial review of these documents and developed a list of audit questions in the attachment to this email. Please post the responses to the questions to the online portal as the responses are completed (but no later than one week before the date of the audit meeting). The dates and times for the audit discussions have been established for July 11 through July 13, 2023, to conduct the meetings via MS Team teleconference call. The proposed agenda for the audit discussions will be provided at a later date.
Please contact me at any time prior if a clarification discussion is needed. We look forward to discussing these questions and Entergys responses during the virtual audit meeting.
Tom Wengert Project Manager - Arkansas Nuclear One NRR/DORL/LPL4 (301) 415-4037
Hearing Identifier:
NRR_DRMA Email Number:
2127 Mail Envelope Properties (PH0PR09MB7403F0D8E32EE0B1DB68F2448F5AA)
Subject:
Arkansas Nuclear One, Unit 1 - TSTF-505 Audit Questions (EPID L-2022-LLA-0197)
Sent Date:
6/14/2023 11:11:33 AM Received Date:
6/14/2023 11:11:00 AM From:
Thomas Wengert Created By:
Thomas.Wengert@nrc.gov Recipients:
"Keele Jr, Riley D" <rkeele@entergy.com>
Tracking Status: None "Jennifer Dixon-Herrity" <Jennifer.Dixon-Herrity@nrc.gov>
Tracking Status: None "Thomas Byrd" <Thomas.Byrd@nrc.gov>
Tracking Status: None "REID, MARK" <mreid1@entergy.com>
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PH0PR09MB7403.namprd09.prod.outlook.com Files Size Date & Time MESSAGE 1831 6/14/2023 11:11:00 AM ANO-1 TSTF-505 Audit Questions 06142023.pdf 253555 Options Priority:
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AUDIT QUESTIONS LICENSE AMENDMENT REQUEST TO REVISE TECHNICAL SPECIFICATIONS TO ADOPT TSTF-505, REVISION 2 ENTERGY OPERATIONS, INC.
ARKANSAS NUCLEAR ONE, UNIT 1 DOCKET NO. 50-313 By application dated December 22, 2022 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML22356A249), Entergy Operations, Inc. (the licensee) submitted a license amendment request (LAR) for Arkansas Nuclear One, Unit 1 (ANO-1). The amendment would revise technical specification (TS) requirements to permit the use of risk-informed completion times (RICTs) for actions to be taken when limiting conditions for operation (LCOs) are not met. The proposed changes are based on Technical Specifications Task Force (TSTF) Traveler TSTF-505, Revision 2, dated July 2, 2018 (ML18183A493). The U.S. Nuclear Regulatory Commission (NRC) issued a final revised model safety evaluation (SE)
(ML18269A041) approving TSTF-505, Revision 2, on November 21, 2018.
The NRC staff determined that the following information is needed to complete its review.
Probabilistic Risk Analysis (PRA) Licensing Branch A (APLA) PRA Acceptability and Risk-informed Approach APLA Question 01 - Digital Instrumentation and Control (I&C) Modeling Concerning the quality of the PRA model, Nuclear Energy Institute (NEI) 06-09-A, Risk-Informed Technical Specifications Initiative 4b Risk-Managed Technical Specifications (RMTS)
Guidelines, Revision 0-A (ML12286A322), states that Regulatory Guide (RG) 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis and RG 1.200, Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities define the quality of the PRA in terms of its scope, level of detail, and technical adequacy. The quality must be compatible with the safety implications of the proposed TS change and the role the PRA plays in justifying the change.
Regarding digital I&C, the NRC staff notes the lack of consensus industry guidance for modeling these systems in plant PRAs to be used to support risk-informed regulatory applications. In addition, known modeling challenges exist, such as the lack of industry data for digital I&C components, the difference between digital and analog system failure modes, and the complexities associated with modeling software failures, including common-cause software failures. Also, although reliability data from vendor tests may be available, this source of data is not a substitute for in-the-field operational data. Given these challenges, the uncertainty associated with modeling a digital I&C system could impact the RICT program. Therefore, address the following:
a) Clarify whether digital I&C systems are credited in the PRA models that will be used in the RICT program.
b) If digital I&C systems are credited in the PRA models that will be used in the RICT program, provide justification (e.g., describe and provide the results of a sensitivity study) that demonstrates the modeling uncertainty associated with crediting digital I&C systems has an inconsequential impact on the RICT calculations.
Alternatively, if a justification is not provided, identify which LCOs are determined to be impacted by digital I&C systems modeling for which risk management actions (RMAs) will be applied during a RICT. Explain and justify the criteria used to determine what level of impact to the RICT calculation requires additional RMAs.
APLA Question 02 - Consideration of Shared Systems in RICT Calculations RG 1.200, Revision 2, states, in part: The base PRA serves as the foundational representation of the as-built and as-operated plant necessary to support an application.
The LAR does not appear to address the existence of crossties between units. However, the NRC staff has reviewed system documents in the portal that have shared systems. The NRC staff notes that for some of these systems, it appears the sharing of a system is not consistent between units. It appears that some operational aspects, such as alternate alignments, were excluded from the PRA models. For multi-unit events (e.g., loss of offsite power and seismic events), credit for a shared system may be limited to one unit.
Clarify what systems are shared, how they are shared, and whether they can support the other unit in an accident. Explain how the shared systems are credited for each unit in the PRA models. This discussion should also address the following:
a) Explain whether shared systems are credited in the internal events, including flood and fire PRA models for both units and, if so, identify those systems.
b) If shared systems are credited in the Real Time Risk (RTR) model that supports the RICT calculations, then explain how the shared system is modelled for each unit in a dual unit event demonstrating that shared systems are not over-credited in the PRA models.
c) If a shared system is credited in the RTR model that supports the RICT calculations and the impact of events that can create a concurrent demand for the system shared by both units is not addressed in the PRA models, then justify that this exclusion has an inconsequential impact the RICT calculations APLA Question 03 - Impact of Seasonal Variations The Tier 3 assessment in RG 1.177, An Approach for Plant-specific, Risk-informed Decision-making: Technical Specifications, Revision 2 (ML20164A034), stipulates that a licensee should develop a program that ensures the risk impact of out-of-service equipment is appropriately evaluated prior to performing any maintenance activity. NEI 06-09-A and its associated NRC SE (ML071200238) state that, for the impact of seasonal changes, either conservative assumptions
should be made, or the PRA should be adjusted appropriately to reflect the current (e.g.,
seasonal or time of cycle) configuration.
The LAR does not appear to address modeling adjustments needed to account for seasonal and time of cycle dependencies and what kind of adjustments will be made. Therefore, address the following to clarify the treatment of seasonal and time of cycle variations:
a) Explain how the RICT calculations address changes in PRA data points, basic events, and structures, systems, and component (SSC) operability constraints as a result of extreme weather conditions, seasonal variations, other environmental factors, or time of cycle. Also, explain how these adjustments are made in the configuration risk management program (CRMP) model and how this approach is consistent with the guidance in NEI 06-09-A and its associated NRC final SE.
b) Describe the criteria used to determine when PRA adjustments due to extreme weather conditions, seasonal variations, other environmental factors, or time of cycle variations need to be made in the CRMP model and what mechanism initiates these changes.
APLA Question 04 - Open Phase Condition Section C.1.4 of RG 1.200 states that the base (e.g., Model of Record) PRA is to represent the as-built, as-operated plant to the extent needed to support the application. The licensee is to have a process that identifies updated plant information that necessitate changes to the base PRA model.
In response to the January 30, 2012, Open Phase Condition (OPC) event at the Byron Generating Station, the NRC issued Bulletin 2012-011. As part of the initial Voluntary Industry Initiative for mitigation of the potential for the occurrence of an OPC in electrical switchyards2, licensees have made the addition of an Open Phase Isolation System (OPIS). As per SRM-SECY-16-0068 3, the NRC staff was directed to ensure that licensees have appropriately implemented OPIS and that licensing bases have been updated accordingly. From the revised voluntary initiative4 and resulting industry guidance in NEI 19-025 on estimating OPC and OPIS risk, it is understood that the risk impact of an OPC can vary widely dependent on electrical switchyard configuration and design. Considering these observations, provide the following information:
a) For ANO-1, discuss the evaluation of the risk impact associated with OPC events including the likelihood of OPC initiating plant trips and the impact of those trips on PRA modeled SSCs. Also, explain whether an OPIS has been installed and if it has been installed, then discuss its functionality and any operator actions needed to operate the system or needed in response to the system.
1 U.S. NRC Bulletin 2012-01, Design Vulnerability in Electric Power System ((ML12074A115).
2 Anthony R. Pietrangelo to Mark A. Satorius, Letter re: Industry Initiative on Open Phase Condition - Functioning of Important-to-Safety Structures, Systems and Components (SSCs), dated October 9, 2013 ((ML13333A147).
3 U.S. NRC SRM-SECY-16-0068, Interim Enforcement Policy for Open Phase Conditions in Electric Power Systems for Operating Reactors, dated March 9, 2017 (ML17068A297).
4 Doug True to Ho Nieh, Letter re: Industry Initiative on Open Phase Condition, Revision 3, dated June 6, 2019 (ML19163A176).
5 Nuclear Energy Institute (NEI) 19-02, Guidance for Assessing Open Phase Condition Implementation Using Risk Insights, Revision 0, April 2019 (ML19122A321).
b) Clarify whether any installed OPIS equipment and associated operator actions are credited in the PRAs that support this application. If OPIS equipment and associated operator actions are credited, then provide the following information:
ii. Describe the impact that this treatment, if any, has on key assumptions and sources of uncertainty for the RICT program.
iii. Discuss Human Reliability Analysis (HRA) methods and assumptions used for crediting OPIS alarm manual response.
iv. Discuss how OPC related scenarios are modelled for non-internal event scenarios such as internal floods, fire, and seismic.
- v. Regarding inadvertent OPIS actuation:
- a. Explain whether scenarios regarding inadvertent actuation of the OPIS, if applicable, are included in the PRA models that support the RICT calculations.
- b. If inadvertent OPIS actuation scenarios are not included in the PRA models, then provide justification that the exclusion of this inadvertent actuation does not impact the RICT calculations.
c) If OPC and OPIS are not included in the application PRA models (whether OPIS equipment is installed or not), then provide justification that the exclusion of this failure mode and mitigating system does not impact the RICT calculations.
As an alternative to Part (c), propose a mechanism to ensure that OPC-related scenarios are incorporated into the application PRA models prior to implementing the RICT program.
APLA Question 05 - Performance Monitoring The NRC SE for NEI06-09-A, states in part: The impact of the proposed change should be monitored using performance measurement strategies. NEI 06-09-A considers the use of NUMARC93-01, Revision 4F, Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants (ML18120A069), as endorsed by RG 1.160, Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, Revision 4 (ML18220B281), for the implementation of the Maintenance Rule. NUMARC 93-01, Section 9.0, contains guidance for the establishment of performance criteria.
In addition, the NEI 06-09-A methodology satisfies the five key safety principles specified in RG1.177, Revision 2 relative to the risk impact due to the application of a RICT. Moreover, NRC staff position C.3.2 provided in RG 1.177, Revision 2, for meeting the fifth key safety principle acknowledges the use of performance criteria to assess degradation of operational safety over a period. It is unclear how the licensees RICT program captures performance monitoring for the SSCs within the scope of the RMTS program. Therefore:
a) Confirm that the ANO-1 Maintenance Rule program incorporates the use of performance criteria to evaluate SSC performance as described in NUMARC 93-01, as endorsed by RG 1.160.
b) Alternatively, describe the approach or method used by ANO-1 for SSC performance monitoring, as described in NRC staff position C.3.2 of RG 1.177, Revision 2, for meeting the fifth key safety principle. In the description, include criteria (e.g., qualitative, or quantitative) along with the appropriate risk metrics and explain how the approach and criteria demonstrate the intent to monitor the potential degradation of SSCs in accordance with the NRC SE for NEI 06-09-A.
APLA Question 06 - In-Scope LCOs and Corresponding PRA Modeling The NRCs SE for NEI 06-09-A specifies that the LAR should provide a comparison of the TS functions to the PRA modeled functions to show that the PRA modeling is consistent with the licensing basis assumptions or to provide a basis for when there is a difference. Table E1-1, In Scope TS/LCO Conditions to Corresponding PRA Functionsof LAR Enclosure 1 identifies each LCO in the TSs proposed for inclusion in the RICT program. The table also describes whether the systems and components covered by the LCO are modeled in the PRA and, if so, presents both the design success criteria and PRA success criteria. For certain LCOs, the table explains that the associated SSCs are not modeled in the PRAs but will be represented using a surrogate event that fails the function performed by the SSC. For some LCOs, the LAR did not provide an adequate description for the NRC staff to conclude that the PRA modeling will be sufficient.
a) Regarding Improved Technical Specification (ITS) LCO 3.3.1.B, LAR Table E1-1 states that, for any reactor protection system (RPS) instrumentation not modeled, the RPS channel logic that is modeled will be used as a surrogate. It is unclear to the NRC staff what surrogates will be used for each affected function and if they are bounding and conservative.
i) Clarify which component of RPS channel logic will be used as a surrogate for each of the RPS instruments affected.
ii) Provide justification that the surrogate conservatively bounds each of the RPS instruments.
b) Regarding ITS LCO 3.3.6.A, Table E1-1 states that, for engineered safeguards actuation system (ESAS) manual initiation, that either ESAS master relays, operator actions, or automatic actuations for the related function that is modeled will be used as a surrogate.
It is unclear to the NRC staff what surrogates will be used for this affected function and if they are bounding and conservative.
i) Clarify which of the three listed items will be used as a surrogate for the ESAS function affected.
ii) Provide justification that the surrogate conservatively bounds the ESAS function.
c) Regarding ITS LCO 3.3.12.A and 3.3.12.B, Table E1-1 states that, for emergency feedwater initiation and control system (EFIC) manual initiation, that either EFIC/Main Steam Line Isolation master relays, operator actions, or automatic actuations for the
related function that is modeled will be used as a surrogate. It is unclear to the NRC staff what surrogates will be used for each affected function and if they are bounding and conservative.
i) Clarify which of the three listed items will be used as a surrogate for the ESAS function affected.
ii) Provide justification that the surrogate conservatively bounds the EFIC function.
d) Regarding ITS LCO 3.6.2.C, Table E1-1 states that, for any reactor building air lock not modeled, a pre-existing containment failure event that is modeled will be used as a surrogate. It is unclear to the NRC staff what surrogates will be used for each affected function and if they are bounding and conservative.
i) Clarify which containment failure event will be used as a surrogate for each of the system isolation functions affected.
ii) Provide justification that the surrogate conservatively bounds each of the air lock functions.
e) Regarding ITS LCO 3.6.3.A and 3.6.3.C, Table E1-1 states that, for any reactor building isolation valves not modeled, a representative leak event that is modeled will be used as a surrogate. In the LAR, the licensee also states that multiple conditions regarding these LCOs will use a representative surrogate. It is unclear to the NRC staff which pathways will be used for each affected function and if they are bounding and conservative.
i) Clarify which representative leak event will be used as a surrogate for each of the system isolation functions affected.
ii) Provide justification that the surrogate conservatively bounds each of the isolation functions.
f) Regarding ITS LCO 3.6.6.A, (i.e., ANO-1 TS 3.6.5.A), Table E1-1 states that the reactor building coolers can be modeled as a surrogate for the reactor building spray system affected function. It is unclear to the NRC staff what surrogates will be used for this affected function and if they are bounding and conservative.
i) Clarify which component of reactor building coolers will be used as a surrogate for the reactor building spray function affected.
ii) Provide justification that the surrogate conservatively bounds the reactor building spray functions.
APLA Question 07 - Systems Not Credited in the Fire PRA RG 1.200 states in part: NRC reviewers, [will] focus their review on key assumptions and areas identified by peer reviewers as being of concern and relevant to the application. The NRC staff evaluates the acceptability of the PRA for each new risk-informed application and as discussed in RG 1.174, recognizes that the acceptable technical adequacy of risk analyses necessary to support regulatory decision-making may vary with the relative weight given to the risk assessment element of the decision-making process. The NRC staff notes that the calculated
results of the PRA are used directly to calculate a RICT, which subsequently determines how long SSCs (both individual SSCs and multiple, unrelated SSCs) controlled by technical specifications can remain inoperable. Therefore, the PRA results are given a very high weight in a TSTF-505 application and the NRC staff requests additional information on the following issues that have been previously identified as potentially key fire PRA assumptions.
In reviewing the ANO-1 Uncertainty Analysis [PSA-ANO1-06-4B-SOU] provided on the portal, the NRC staff noted that a sensitivity was performed on fire PRA assumption of failing components with unknown cable locations. The sensitivity study results demonstrated a significant impact on a few of the RICT program LCOs.
The NRC staff notes that some conservative PRA modeling could have a nonconservative impact on the RICT calculations. If an SSC is part of a system not credited in the fire PRA or it is supported by a system that is assumed to always fail, then the risk increases due to taking that SSC out of service are masked. Therefore, address the following:
a) Identify the SSCs that are assumed to be always failed in the fire PRA, or are not included in the fire PRA, due to lack of cable tracing or other reasons.
b) Justify that this assumption has an inconsequential impact on the RICT calculations.
c) If, in response to part (b) above, it cannot be determined that the cited assumption has an inconsequential impact on the estimated RICTs, then identify what programmatic changes will be considered to compensate for this uncertainty and the basis for their consideration (e.g., identification of additional RMAs).
Probabilistic Risk Analysis (PRA) Licensing Branch C (APLC) PRA Acceptability and Risk-informed Approach APLC Question 01 - Determination of the High Winds CDF and LERF Penalty Section 2.3.1, Item 7, of NEI 06-09, Revision 0-A, states, in part, that the impact of other external events risk shall be addressed in the RMTS program, and explains that one method to do this is by performing a reasonable bounding analysis and applying it along with the internal events risk contribution in calculating the configuration risk and the associated RICT.The NRC staffs safety evaluation for NEI 06-09 states, in part, that [w]here PRA models are not available, conservative or bounding analyses may be performed to quantify the risk impact and support the calculation of the RICT.
Section 4.2 of LAR Enclosure 4 provides the results of the development of Core Damage Frequency (CDF) and Large Early Release Frequency (LERF) penalty factors to include in RICT calculations to bound the impact of tornado-generated missiles for certain maintenance or LCO configurations.It is stated that these penalty factors are conservative. However, only the results of this assessment are provided; no description is provided of the methodology, input, and assumptions used to develop the risk model and to justify that the results are conservative.
Provide the following:
a) Identify the SSCs that are the tornado missile risk targets for the development of the tornado-generated missile CDF and LERF penalty factors for the RICT calculations and provide justification for why these were selected for evaluation.
b) A description of the approach used for the development of the tornado-generated missile CDF and LERF penalty factors for the RICT calculations with justification for the results of the approach being conservative. The description and justification should (i) include information about the tornado missile failure frequencies, conditional failure probabilities for impacted SSCs, and the plant response model, and (ii) identify any deviations from the Tornado Missile Risk Evaluator methodology approved for use for ANO-1 (ML20135H141).
Electrical Engineering Branch (EEEB) Audit Questions EEEB Question 01-TS LCO 3.8.1, Conditions A, B, and C General Design Criterion (GDC) 17 requires, in part, that both offsite and onsite electrical power systems be provided. LCO 3.8.1, Conditions A, B, and C are exclusively for the inoperability of one offsite circuit, one diesel generator (DG), and two offsite circuits, respectively.
ANO-1 Safety Analysis Report (SAR) Amendment 30 (ML21288A074), Section 8.3.1.1.2, states that there are two offsite sources for ANO-1, Startup Transformers 1 and 2, that provide safe shutdown for ANO-1 and maintain it in a safe shutdown condition. For loss of one offsite power source with the main generator unavailable, the available offsite power source is sufficient for unit safe shutdown. TS Bases 3.8.1, Background, states that for the loss of offsite power to the startup transformer supplying ANO-1, an undervoltage condition trips its associated bus feeder breakers, the feeder breakers for the alternate Startup Transformer automatically close allowing it to supply ANO-1 at reduced load (if Startup Transformer 2 is supplying ANO-1). Upon loss of the normal (main generator) and the standby power sources (both offsite sources), each of the two 4160 V engineered safeguards buses are energized from its respective diesel generator.
The design success criteria (DSC) in LAR Table E1-1 for TS LCO 3.8.1, Conditions A, B, and C appears inconsistent with the LCO by not listing minimum power source(s) of the type identified in the respective LCO condition. Clarify or explain this inconsistency.
a) Condition A - Minimum offsite power circuit(s) Startup Transformer 1 or 2 to address LCO.
b) Condition B - Minimum DG(s) to address LCO for design basis accident (DBA).
c) Condition C - Minimum DG(s) to address LCO for DBA since both offsite circuits inoperable.
EEEB Question 02 - TS LCO 3.8.7, Condition A GDC 17 requires, in part, that both offsite and onsite electrical power systems be provided. This includes onsite electrical systems for 120 V alternating current (AC) vital loads.
SAR Section 8.3.1.1.6 states that each of the four redundant 120 V vital AC distribution panels supply power to one of the four channels of Nuclear Instrumentation and Reactor Protection Systems. Each of three channels of ESAS is supplied by one 120 V vital AC distribution panel.
The DSC in LAR Table E1-1 for TS LCO 3.8.7, Condition A appears inconsistent with the LCO by not identifying the minimum number of inverters per train for a DBA. Clarify or explain this inconsistency.
EEEB Question 3 - TS LCO 3.8.9, Conditions A, B, and C GDC 17 requires, in part, that both offsite and onsite electrical power systems be provided. This includes onsite dc electrical systems.
SAR Section 8.3.1.1.2 describes that there are two offsite sources for ANO-1, Startup Transformers 1 and 2, capable of providing a source of power for ANO-1 safe shutdown and post-shutdown conditions. For the loss of one offsite power source with the main generator unavailable, the available offsite power source is sufficient for unit safe shutdown. Upon loss of normal (main generator) and standby power sources (both offsite sources), each of the two 4160 V engineered safeguard buses are energized from its respective DG. ANO-1 SAR Section 8.3.1.1.6 states that each of the four redundant 120 V vital AC distribution panels supplies power to one of four channels of Nuclear Instrumentation and Reactor Protection Systems.
Each of three channels of ESAS is supplied by one 120 V vital AC distribution panel. SAR Section 8.3.2.1 states that the 1E 125 V direct current (VDC) system consists of two 125 V batteries that provide DC power to the two 125 VDC control centers (one per train) and distribution panels. Four battery chargers are supplied, with two serving as normal supplies to the DC control centers. The second battery charger for each control center serves as a standby battery charger.
The DSC in Table E1-1 for TS LCO 3.8.9, Conditions A, B, and C appear inconsistent with the LCO by not listing SSCs (AC safety buses or DC buses or 120 VAC distribution panels) as shown in Table B 3.8.9-1 of the TS Bases for each respective LCO condition. Clarify or explain this inconsistency.
a) Condition A - Minimum AC safety buses for DBA.
b) Condition B - Minimum 120 VAC distribution panels for DBA.
c) Condition D - Minimum DC buses for DBA.
Technical Specifications Branch (STSB) Audit Questions STSB QUESTION 01 - TS 5.5.18 Clarity The proposed administrative controls for the RICT Program in TS 5.5.18 paragraph e of to the LAR was based on the TS markups of TSTF-505, Revision 2.
The NRC staff recognizes that the model SE for TSTF-505, Revision 2 contains improved phrasing for the administrative controls for the RICT Program in TS 5.5.7 paragraph e.
Specifically, the phrasing approved for use with this program instead of used to support this license amendment.
Discuss whether the phrases used to support Amendment # xxx or, as discussed in the TSTF-505 model SE, approved for use with this program would provide more clarity for this paragraph, in lieu of the original phrasing in TS 5.5.18 paragraph e.