ML21237A240

Safety Evaluation for Amendment No. LAR-21-001: Clarification of ITAAC Regarding Invessel Components

ML21237A240
Person / Time
Site:	Vogtle
Issue date:	10/15/2021
From:	Gleaves W, Christopher Welch NRC/NRR/VPOB
To:	City of Dalton, GA, Georgia Power Co, MEAG Power, Oglethorpe Power Corp, Southern Nuclear Operating Co
Gleaves B
Shared Package
ML21237A205	List: ML21237A233 ML21237A234 ML21237A235 ML21237A236 ML21237A237 ML21237A238 ML21237A239 ML21237A240
References
EPID L-2021-LLA-0153, LAR-21-001
Download: ML21237A240 (27)

Text

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION

RELATED TO AMENDMENT NOS. 188 AND 186

TO THE COMBINED LICENSE NOS. NPF-91 AND NPF-92, RESPECTIVELY

SOUTHERN NUCLEAR OPERATING COMPANY, INC.

GEORGIA POWER COMPANY

OGLETHORPE POWER CORPORATION

MEAG POWER SPVM, LLC

MEAG POWER SPVJ, LLC

MEAG POWER SPVP, LLC

CITY OF DALTON, GEORGIA

VOGTLE ELECTRIC GENERATING PLANT UNITS 3 AND 4

DOCKET NOS.52-025 AND 52-026

1.0 INTRODUCTION

By letter dated August 24, 2021 (Agencywide Documents Access an d Management System (ADAMS) Accession No. ML21236A305), Southern Nuclear Operating Company, Inc. (SNC or the licensee) requested that the U.S. Nuclear Regulatory Commis sion (NRC) amend the Vogtle Electric Generating Plant (VEGP) Units 3 and 4, Combined Licens e (COL) Nos. NPF-91 and NPF-92, respectively. The request, License Amendment Request ( LAR)21-001, Clarification of ITAAC Regarding Invessel Components, proposed changes that would revise Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) Index Nos. 68 (2.1.03.01), 75 (2.1.03.06.i),

515 (2.5.01.03e), 565 (2.5.05.02.i), and 570 (2.5.05.03b) in CO L Appendix C and plant-specific design control document (PS-DCD) Tier 1 information to (1) elim inate ITAAC requirements regarding verification of the location of certain equipment, (2 ) eliminate the requirement that certain inspections be performed on the as-built components, and (3) make other related changes. In part, the changes address cert ain components that cannot be installed in their final location until after fuel load.

Pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 52.103(g), the Commission must find prior to operation that the acceptance criteria in th e combined license are met (except for those acceptance criteria that the Commission found were me t under § 52.97(a)(2)), where operation includes the licensee loading the initial core into t he reactor. As noted in the VEGP Units 3 and 4 Updated Final Safety Analysis Report (UFSAR), Sub section 14.3.2.2, Inspections, Tests, Analyses, and Acceptance Criteria, one of the selection criteria for ITAAC is that the ITAAC do not include any inspections, tests, or an alyses that are dependent upon conditions that exist only after fuel load. The ITAAC that ar e the subject of this request relate to specific equipment, and include as-built components (for e xample, invessel components) that cannot be placed in their final operational location until after the 10 CFR 52.103(g) finding.

As explained below, ITAAC required to be performed on as-built structures, systems and components (SSCs) may be completed only after the SSCs in quest ion have been installed in their final operational location. Thus, those SSCs that cannot be installed in their final operational location until after the 10 CFR 52.103(g) finding s hould not have been subject to the as-built ITAAC requirement in accordance with 10 CFR 52.103(g ) and the selection criteria identified above.

The VEGP, Units 3 and 4, COL Appendix C, Section 1.1, Definiti ons, defines as-built to mean the physical properties of a structure, system, or compon ent following the completion of its installation or construction activities at its final locati on at the plant site. In cases where it is technically justifiable, determination of physical properties o f the as-built structure, system, or component may be based on measurements, inspections, or tests t hat occur prior to installation, provided that subsequent fabrication, handling, installation, a nd testing does not alter the properties. SNC stated that the ITAAC identified in LAR 21-00 1 cannot be completed as they currently exist in the COL, considering the COL Appendix C defi nition of as-built that establishes that as-built structures, systems and components (SSCs) must be in their final operational location upon ITAAC completion and prior to ITAAC C losure Notification (ICN) submittal, because certain components will not be placed in the ir final location until after the NRC finds that the acceptance criteria for all ITAAC have been met as required by 10 CFR 52.103(g). SNC, in the LAR, proposed changes to certain ITAAC acceptance criteria to address the fact that specific equipment cannot be verified as-built. However, the proposed changes, in part, also affect SSCs that can be installed prior to initia l fuel load.

Pursuant to Section 10 CFR 52.63(b)(1) and 52.98(f), SNC also re quested an exemption from the provisions of 10 CFR Part 52, Appendix D, Design Certifica tion Rule for the AP1000 Design,Section III.B, Scope and Contents. The requested ex emption would allow a departure from the corresponding portions of the certified info rmation in Tier 1 of the generic Westinghouse AP1000 DCD 1. In order to modify the PS-DCD Tier 1 information, the NRC mu st find SNCs exemption request included in its submittal for the LAR to be acceptable. The staffs review of the exemption request, as well as the LAR, is include d in this safety evaluation.

The staff notes that the proposed markups in Enclosure 3 of LAR 21-001, specifically for ITAAC Index No. 68 and ITAAC Index No. 565, item 2.i), for Units 3 an d 4, contain non-standard language for this type of change. The issued pages for Units 3 and 4 were edited by the NRC to reflect the typical language for similar changes. That is, for ITAAC Index No. 68, the NRC revised the text in the Design Commitment column to show the te xt, for Unit 3, Not used per Amendment No. 188 (and Not used per Amendment No. 186 for Un it 4), and deleted the Not used in both the Inspections, Tests, Analyses, and the Accepta nce Criteria columns. For ITAAC Index No. 565, Item 2.i), the staff replaced Not Used i n the ITA and AC columns with Not used per Amendment No. 188 for Unit 3 (and Not used per Amendment No. 186 for Unit

1 While the licensee describes the requested exemption as being from Section III.B of 10 CFR Part 52, Appendix D, the entirety of the exemption pertains to proposed departures from Tier 1 information in the generic DCD. In the remainder of this evaluation, the NRC will refer to the exemption as an exemption from Tier 1 information to match the language of Section VIII.A.4 of 10 CFR Part 52, Appendix D, which specifically governs the granting of exemptions from Tier 1 information.

4). The staff views this as an editorial change that is not in any way substantively different from what was proposed, and the change is being made only to conform with standard usage to eliminate the confusion that may be created if the text remaine d as shown in the LAR markup for ITAAC Index No. 68 and ITAAC Index No. 565, item 2.i).

The staff notes that the LARs proposed markups for ITAAC Index No. 75, item 6.iii), for both Units 3 and 4 do not contain the same language that was issued by the NRC in this license amendment. The issued text for ITAAC Index No. 75, item 6.iii) in this amendment applies the requested change to a more limited set of components than the m arkup in LAR Enclosure 3 to be consistent with the justification provided in the LAR and th e LARs discussion of SNCs planned actions. The basis for this limitation is discussed in the evaluation below. On October 13, 2021, SNC sent revised markup pages consistent with this li mitation in the staffs approval (ADAMS Accession No. ML21287A258).

2.0 REGULATORY EVALUATION

LAR 21-001 proposes changes to the COL, Appendix C, and Tier 1 PS-DCD for both VEGP Units 3 and 4, in a way that, if approved, would allow completi on of the identified ITAAC prior to fuel load consistent with the existing facility design. The st aff considered the following regulatory requirements in reviewing the LAR that included the proposed changes:

Appendix D,Section VIII.A.4 to 10 CFR Part 52 states that exem ptions from Tier 1 information are governed by the requirements in 10 CFR 52.63(b)(1) and 10 C FR 52.98(f). It also states that the Commission will deny a request for an exemption from T ier 1 if it finds that the design change will result in a significant decrease in the level of sa fety otherwise provided by the design.

10 CFR 52.63(b)(1) allows the licensee who references a design certification rule to request NRC approval for an exemption from one or more elements of the certification information. The Commission may only grant such a request if it determines that the exemption will comply with the requirements of 10 CFR 52.7, which, in turn, points to the requirements listed in 10 CFR 50.12 for specific exemptions. In addition to the facto rs listed in 10 CFR 52.7, the Commission shall consider whether the special circumstances, as required by 10 CFR 52.7 and 50.12, outweigh any decrease in safety that may result from the reduction in standardization caused by the exemption. Therefore, any exemption from the Tie r 1 information certified by Appendix D to 10 CFR Part 52 must meet the requirements of 10 C FR 50.12, 52.7, and 52.63(b)(1).

10 CFR 52.97(b) requires the NRC to identify within the combin ed license the inspections, tests, and analyses, including those applicable to emergency pl anning, that the licensee shall perform, and the acceptance criteria that, if met, are necessar y and sufficient to provide reasonable assurance that the facility has been constructed and will be operated in conformity with the license, the provisions of the [Atomic Energy Act of 1 954, as amended (AEA)], and the Commissions rules and regulations.

10 CFR 52.98(f) requires NRC approval for any modification to, addition to, or deletion from, the terms and conditions of a COL.

10 CFR 50.49 requires that each holder of a combined license es tablish a program for environmental qualification (EQ) of electric equipment important to safety and maintain documentation of that qualification.

10 CFR Part 50.55a, Codes and standards, paragraphs (a)(2)(ii i)-(iv) incorporate by reference the requirements of Institute of Electrical and Electronics Eng ineers (IEEE) 603-1991 and the correction sheet dated January 30, 1995, into the rule. The ru le also states, in part, in section 50.55a(h)(3) that Applications filed on or after May 13, 1999, for construction permits and operating licenses under this part, and for design approvals, d esign certifications, and combined licenses under 10 CFR Part 52 of this chapter, must meet the re quirements for safety systems in IEEE Std. 603-1991 and the correction sheet dated January 30, 1995.

10 CFR Part 50, Appendix A, General Design Criterion (GDC) 2, a s it relates to protection against natural phenomena states the following: Structures, s ystems, and components important to safety shall be designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiche s without loss of capability to perform their safety functions. The design bases for these str uctures, systems, and components shall reflect: (1) Appropriate consideration of the most severe of the natural phenomena that have been historically reported for the site and surrounding area, with sufficient margin for the limited accuracy, quantity, and period of time i n which the historical data have been accumulated; (2) appropriate combinations of the effects o f normal and accident conditions with the effects of the natural phenomena; and (3) t he importance of the safety functions to be performed.

10 CFR Part 50, Appendix A, GDC 4, as it relates to the environ mental and dynamic effects design bases states the following: Structures, systems, and c omponents important to safety shall be designed to accommodate the effects of and to be compa tible with the environmental conditions associated with normal operation, maintenance, testi ng, and postulated accidents, including loss-of-coolant accidents. These structures, systems, and components shall be appropriately protected against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids, that may result from equipment failures and from events and conditions outside the nuclear power unit. However, dynamic effects assoc iated with postulated pipe ruptures in nuclear power units may be excluded from the design basis when analyses reviewed and approved by the Commission demonstrate that the probability of fluid system piping rupture is extremely low under conditions consistent with the design ba sis for the piping.

10 CFR Part 50 Appendix A, GDC 24, as it relates to separation of protection and control systems states, in part, the following: The protection system shall be separated from control systems to the extent that failure of any single control system component or channel, or failure or removal from service of any single protection system compone nt or channel which is common to the control and protection systems leaves intact a sy stem satisfying all reliability, redundancy, and independence requirements of the protection sys tem.

Regulatory guidance referred to in this evaluation includes the following:

• Nuclear Energy Institute (NEI) 08-01, Revision 5 - Corrected, Industry Guideline for the ITAAC Closure Process Under 10 CFR Part 52, June 2014 (ADAMS A ccession No.

ML14182A158);

• Regulatory Guide (RG) 1.215, Revision 2, Guidance for ITAAC C losure Under 10 CFR Part 52, July 2015 (ADAMS Accession No. ML15105A447); and

• RG 1.89, Revision 1, Environmental Qualification of Certain E lectric Equipment Important to Safety for Nuclear Power Plants, June 1984 (ADAMS Accession No. ML003740271).

3.0 TECHNICAL EVALUATION

LAR 21-001 proposes changes that would revise ITAAC Index No. 6 8 (2.1.03.01), No. 75 (2.1.03.06.i), No. 515 (2.5.01.03e), No. 565 (2.5.05.02.i), and No. 570 (2.5.05.03b) to (1) eliminate ITAAC requirements regarding verification of the loca tion of certain equipment, (2) eliminate the requirement that certain inspections be performed on the as-built components, and (3) make other related changes. In part, the changes addres s certain components that cannot be installed in their final location until after fuel lo ad. As such, SNC states that these ITAAC conflict with 10 CFR 52.103(g), which requires that all I TAAC must be completed prior to loading the initial core, and statements of what constitutes an acceptable ITAAC in the VEGP UFSAR.

3.1 EVALUATION OF THE REQUESTED CHANGES

The NRC staff evaluated the information presented by the licens ee in Enclosures 1 - 3 of the August 24, 2021 submittal in determining the acceptability of L AR 21-001. The NRC staff did not evaluate the information provided in Enclosure 4, Draft Re vised Uncompleted Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) Notices (UINs) Pending NRC Approval of LAR and Exemption, because it was submitted for information o nly. The staff will evaluate UINs and ITAAC Closure Notifications (ICNs) when formally submi tted on the docket by the licensee.

3.1.1 ITAAC Index No. 68 (2.1.03.01)

As described in LAR 21-001, SNC proposed to remove the ITAAC In dex No. 68 requirement to verify the reactor system (RXS) functional arrangement. Append ix C of the VEGP Units 3 and 4 COLs defines functional arrangement (for a system) as the ph ysical arrangement of systems and components to provide the service for which the system is i ntended, and which is described in the system design description.

The purpose and scope of functional arrangement ITAAC are discu ssed in NEI 08-01, Industry Guideline for the ITAAC Closure Process Under 10 CFR Part 52, Revision 5 - Corrected (ADAMS Accession No. ML14182A158), which is approved for use by RG 1.215, Guidance for ITAAC Closure Under 10 CFR Part 52, Revision 2 (ADAMS Accessio n No. ML15105A447),

with certain exceptions and additional guidance not relevant to this LAR. The functional arrangement ITAAC for a system is limited to the components ide ntified in the Tier 1 Design Description for the system, including any referenced tables and figures.

In LAR 21-001, SNC requested that ITAAC Index No. 68 be deleted based on the following statements:

Because certain RXS components will not be in their final opera tional location until after fuel is loaded, ITAAC Index No. 68 cannot be closed in conforma nce with the interpretation and understanding that as-built SSCs must be i nstalled in their final operational location prior to ITAAC closure. Furthermore, for those certain RXS components, ITAAC Index No. 68 with this interpretation of as-built does not meet the UFSAR Subsection 14.3.2.2 selection criteria of the ITAAC do not include any inspections, tests, or analyses that are dependent upon conditi ons that exist only after fuel load.

Other ITAAC also demonstrate that the reactor system has been constructed in accordance with the design to the extent possible prior to fuel load. For example, ITAAC 2.1.03.02a (also referred to as ITAAC Index No. 69) demonstrate s that the as-built RXS accommodates the fuel assembly and control rod drive mechanism pattern shown in Figure 2.1.3-1 and the control assemblies (rod cluster and gray rod) and drive rod arrangement shown in Figure 2.1.3-2. ITAAC 2.1.03.02c (also re ferred to as ITAAC Index No. 71) demonstrates that the as-built RXS accommodates t he reactor vessel arrangement shown in Figure 2.1.3-3. ITAAC 2.1.03.03 (also ref erred to as ITAAC Index No. 72) demonstrates that inspections are performed and that th e specified RXS components and pressure boundary welds meet [American Society o f Mechanical Engineers] ASME Code Section III requirements; additionally, a hydrostatic test is performed as required by the ASME Code Section III. ITAAC 2.1. 03.06.i (also referred to as ITAAC Index No. 75) demonstrates that the specified equip ment has been adequately qualified under seismic conditions and harsh environ ments per design requirements. ITAAC 2.1.03.13 (also referred to as ITAAC Index No. 88) demonstrates that the fuel assemblies and rod cluster control assemblies int ended for the initial core load and listed in Table 2.1.3-1 have been designed and constru cted in accordance with the principal design requirements. Finally, ITAAC 2.1.03.14 (a lso referred to as ITAAC Index No. 89) demonstrates the acceptability of the reactor ves sel head top surface and penetration nozzles through a preservice visual examination.

The staff reviewed the information in Enclosure 1 of LAR 21-001 and finds that the fuel assemblies, the rod control cluster assemblies (RCCAs), the gra y rod cluster assemblies (GRCAs), and the incore instrument QuickLoc assemblies cannot b e installed in their final location to support plant operations until after the 10 CFR 52. 103(g) finding has been made and therefore, the as-built requirement for installed location cann ot be met. Additionally, the subsequent reactor vessel assembly and fuel load is adequately controlled and verified during the initial test program. Furthermore, the staff finds that th e ITAAC Index No. 68 requirement for these components does not meet the UFSAR Subsection 14.3.2.2 se lection criteria of the ITAAC do not include any inspections, tests, or analyses that a re dependent upon conditions that exist only after fuel load. The staff notes that functio nal arrangement, described in COL Appendix C, Section 2.1.3, includes Table 2.1.3-3, that also sp ecifies the location of the fuel assemblies, RCCAs, and GRCAs prior to fuel load is in the auxil iary building, indicating there was no expectation that these components would be installed in their final as-built location prior to fuel loading.

Additionally, the staff finds the functional arrangement for th e remainder of the equipment listed in Table 2.1.3-1, (i.e., the reactor vessel (RV), the reactor u pper and lower internals, the control rod drive mechanisms, and source, intermediate, and power range detectors), which can be installed prior to the 10 CFR 52.103(g) finding, is adequately verified via other ITAAC. In addition to those ITAAC identified by the licensee, the staff f inds ITAAC Index No. 78 performs both pre-and post-flow test visual inspections of the as-buil t reactor internals and verifies the lower internals are equipped with holders for material surveill ance capsules. ITAAC Index No. 80 verifies the flow area of the RV direct vessel injection (DVI) nozzles. Finally, the reactor coolant system (RCS) ITAAC Index No. 41 verifies RCS flow rate through the RV.

The staff finds that SNCs proposed changes do not modify the d esign of equipment, delete any technical requirements, or impact the ability of an SSC to perf orm its function and that the remaining ITAAC continue to provide reasonable assurance that t he facility has been constructed and will be operated in conformity with the license, the AEA, and NRC rules and regulations. Therefore, within the scope of this license amend ment request, the NRC concludes that 10 CFR 52.97(b) is satisfied and the proposed ch anges to Table 2.1.3-2 for ITAAC Index No. 68 (2.1.03.01), shown below, are acceptable.

COL Appendix C Table 2.1.3-2 (for ITAAC Index No. 68) is revised as follows for Unit 3 (the Unit 4 COL is revised identically, except that the Amendme nt No. is 186):

Table 2.1.3-2 Inspections, Tests, Analyses, and Acceptance Criteria

No. ITAAC Design Commitment Inspections, Tests, Acceptance Criteria No. Analyses 68 2.1.03.01 Not used per Amendment Inspection of the as-The as-built RXS conforms No. 1881. The functional built system will be with the functional arrangement of the RXS is as performed. arrangement as described in described in the Design the Design Description of this Description of this Section Section 2.1.3.

2.1.3.

3.1.2 ITAAC Index No. 75 (2.1.03.06.i)

Following license amendment 85 and 84 for VEGP, Units 3 and 4, respectively (ADAMS Accession No. ML17216A064), the am ended ITAAC Index No. 75 repr esented a consolidation of former ITAAC Index Nos. 75 (2.1.03.06.i), 76 (2.1.03.06.ii), 77 (2.1.03.06.iii), 81 (2.1.03.09a.i) and 82 (2.1.03.09a.ii). Thus, three ITAAC verification activit ies are required to demonstrate satisfaction of design commitment item 6 in Table 2.1.3-2, whic h states, The seismic Category I equipment identified in Table 2.1.3-1 can withstand seismic des ign basis loads without loss of safety function. These three ITAAC verification activities ar e: (1) verification that specific seismic Category I equipment are located on the Nuclear Island (ITAAC item 6.i)), (2) seismic qualification of the components (ITAAC item 6.ii)), and (3) ver ification that the as-built equipment including anchorage are bounded by the seismic qualif ication (ITAAC item 6.iii)).

SNCs proposed changes to ITAAC Index No. 75 affect ITAAC items 6.i) and 6.iii). No changes were proposed for design commitment item 6 or ITAAC item 6.ii).

ITAAC Index No. 75, Item 6.i) currently states, [i]nspection w ill be performed to verify that the seismic Category I equipment identified in Table 2.1.3-1 is loc ated on the Nuclear Island. The acceptance criterion states, [t]he seismic Category I equipmen t identified in Table 2.1.3-1 is located on the Nuclear Island.

SNC requested item 6.i) be revised to exclude components that m ay not be located on the nuclear island prior to the 10 CFR 52.103(g) finding based on t he following information:

ITAAC Index No. 75, item 6.i), requires that the seismic Catego ry I equipment identified in Table 2.1.3-1 is located on the Nuclear Island. The seismic Category I equipment identified in Table 2.1.3-1 includes components which are store d in protected environments off the nuclear island until it is time for loadin g the initial core, namely the fuel assemblies, the rod cluster control assemblies, the gray r od cluster assemblies, and the incore instrument QuickLoc assemblies.

Fuel loading and operation of the RXS will not and cannot occu r until the equipment is on the Nuclear Island. Thus, confirming this equipment is o n the Nuclear Island is an unnecessary requirement that does not serve the underlying purp ose of this ITAAC which is to confirm completion of the [seismic qualification] a ctivities that can be completed prior to the initial fuel load.

The staff reviewed the information in Enclosure 1 of LAR 21-001, for the change requested for ITAAC Index No. 75, Item 6.i). The staff finds the fuel assemb lies, RCCAs, GRCAs, and incore instrument QuickLoc assemblies cannot be installed in their fin al locations until after the 10 CFR 52.103(g) finding. Removing the components from their protecti ve storage environment, prior to installation, provides neither a technical or safety benefit an d would only serve to subject the components to potential damage. Furthermore, the future instal led location of the components is determined by the location of the reactor vessel, which is v erified by this ITAAC to be on the nuclear island. The staff finds that SNCs proposed changes do not delete any technical requirements, impact the ability of an SSC to perform its funct ion, or impact safety and the reasonable assurance finding.

The staff reviewed the information in Enclosure 1 of LAR 21-001, for the change requested for ITAAC Index No. 75, Item 6.iii). ITAAC Index No. 75, Item 6.ii i) states, Inspection will be performed for the existence of a report verifying that the as-built equipment including anchorage is seismically bounded by the tested or analyzed conditions. The acceptance criterion states,

[a] report exists and concludes that the as-built equipment in cluding anchorage is seismically bounded by the tested or analyzed conditions.

SNC, in LAR 21-001, requested ITAAC Index No. 75, Item 6.iii) b e revised to remove the as-built attribute from the inspection requirement for the equipm ent listed in VEGP, Units 3 and 4, COL Table 2.1.3-1 and acceptance criteria based on the followin g:

The seismic Category I equipment identified in Table 2.1.3-1 in cludes the reactor vessel, the upper and lower internals assemblies, the fuel assemblies, rod assemblies [such as RCCAs and GRCAs], control rod drive mechanisms, incore instrume nt QuickLoc assemblies, and the source, intermediate and power range detect ors. Per the COL license conditions governing testing and fuel loading and UFSAR Section 4.2 describing the fuel system design, the fuel assemblies, the rod cluster co ntrol assemblies, the gray rod cluster assemblies, and the incore instrument QuickLoc asse mblies will not be installed in their final operational locations until after the initial core is loaded with fuel assemblies in the reactor vessel. Accordingly, these portions of ITAAC Index No. 75 cannot be closed in conformance with the interpretation and und erstanding that as-built SSCs must be installed in their final operational location prio r to ITAAC closure.

Therefore, the inspection of these invessel components listed in Table 2.1.3-1 with this interpretation of as-built do not meet the UFSAR Subsection 1 4.3.2.2 selection criteria for ITAAC that the ITAAC do not include any inspections, tests, or analyses that are dependent upon conditions that exist only after fuel load and should be excluded from ITAAC Index No. 75.

To the extent that the installation prior to fuel load is possi ble for other seismically-qualified components, an inspection is conducted to confirm the satisfactory installation of the seismically qualified components identified in the table. The inspection verifies the equipment make/model/serial number, as-designed equipment mount ing orientation, anchorage and clearances, and electrical and other interfaces. For components not installed prior to fuel load, the pre-fuel-load inspection is a ccomplished by verifying a quality assurance data package exists that concludes that the e quipment was constructed as per design. Additional verifications are perfor med following 52.103(g) as addressed in UFSAR Subsection 14.2.10.

The inspection conducted for each component in the table [Table 2.1.3-1] considers the critical seismic attributes identified in the associated Equipm ent Qualification Report for that component. The inspection confirms that the equipment, in cluding anchorage, is seismically bounded by the tested or analyzed conditions.

The staff reviewed the information in Enclosure 1 of LAR 21-001, for the changes requested for ITAAC Index No. 75, Item 6.iii). SNCs proposed revision to it em 6.iii) is to remove the phrase as-built from both the ITA and AC of the ITAAC. This change effectively removes the as-built requirement for all of the equipment listed in Table 2.1.3-1. However, the staff understands based on the LAR that eight types of components listed in this table can be verified in their as-built locations. The staff has divided the equipment listed i n Table 2.1.3-1 into two evaluation sections below, based on these groups.

Fuel Assemblies, RCCAs, GRCAs, and Incore Instrument Quickloc A ssemblies

The staff reviewed UFSAR Section 4.2 and the fuel system design and finds that the fuel assemblies, the RCCAs, GRCAs, and the incore instrument QuickLo c assemblies cannot be installed in their final operational locations until after the 10 CFR 52.103(g) finding and the initial core is loaded in the reactor vessel, and that applying the as -built requirement in this ITAAC to these components does not meet the UFSAR Subsection 14.3.2.2 se lection criteria for ITAAC.

Also, the seismic qualification of these components can still b e demonstrated without the components being installed. UFSAR Section 3.10.1.2, Performan ce Requirements for Seismic Qualification, states that an equipment qualification data pac kage is developed for the instrumentation and electrical equipment classified as seismic Category I. Each equipment qualification data package establishes the safety-related funct ional requirements of the equipment to be demonstrated during and after a seismic event. UFSAR Section 3.10.2 identifies that the methods and procedures for qualification of seismic Category I electrical equipment, instrumentation, and mechanical equipment is by anal ysis and testing to demonstrate structural integrity and operability. UFSAR Append ix 3D, Attachment E, Section E.6, Qualification by Analysis, states analysis is used to demo nstrate the structural adequacy of the passive mechanical equipment being qualified by showing tha t the calculated stresses do not exceed the design allowable stresses specified in ASME Code,Section III. UFSAR Section 3.10.2.2, Seismic and Operability Qualification of Active Mech anical equipment, states that active mechanical equipment is qualified for both structural in tegrity and operability for its intended service conditions by a combination of test and analys is. The test and analysis methods utilized in qualification of these components provide a dequate confidence of operability under required plant conditions because the qualification metho ds explicitly consider the installation and components location. With the removal of the as-built requirement for these components, ITAAC item 6.iii) still requires inspection and ver ification that equipment in Table 2.1.3-1 is seismically bounded by the tested or analyzed condit ions including anchorage.

Additional verifications are performed following the 10 CFR 52. 103(g) finding, as addressed in UFSAR Subsection 14.2.10, Startup Test Procedures. Therefore, the subsequent reactor vessel assembly and fuel load is adequately controlled and veri fied during the initial test program. And in accordance with 10 CFR Part 50, Appendix B, Q uality Assurance Criteria for Nuclear Power Plants and Fuel Repr ocessing Plants, the license e must follow quality assurance requirements for installation to ensure that the fina l installed configuration is bounded by the analysis and test used for qualification. The staff is approving the proposed change to remove as-built from the ITAAC item 6.iii) for only these fou r types of components.

Reactor Vessel, Reactor Upper Internals Assembly, Reactor Lower Internals Assembly, CRDMs, Source Range Detectors, Intermediate Range Detectors, Up per and Lower Power Range Detectors

SNC did not provide, for these eight types of components, a bas is for how removing as-built from item 6.iii) meets the requirement in 10 CFR 52.97(b) that ITAAC verify that the facility has been constructed as required. Instead, the LAR indicates that SNC intends to complete the ITAAC for these components after their installation. Since a j ustification for removing the as-built requirement for these eight types of components was not provided, and the LAR indicates that SNC, in fact, plans to satisfy the as-built requirement for these components, the staff is not approving the removal of the as-built requirement for the se eight types of components.

Therefore, the staffs approval of the change to ITAAC Index no. 75, item 6.iii) is limited to the four types of components previously discussed. For the remaini ng eight types of components, removal of the as-built requirement is not approved. For tha t reason, the NRC is adding a sentence to SNCs proposed revision of the ITA in ITAAC Index N o. 75 item 6.iii) as follows:

Inspection will be performed for the existence of a report veri fying that the equipment including anchorage is seismically bounded by the tested or ana lyzed conditions. This inspection must be performed on the as-built equipment except f or the fuel assemblies, rod cluster control assemblies, gray rod cluster assemblies, an d incore instrument QuickLoc assemblies.

The staff provided its analysis of SNCs proposed change to exc lude certain equipment from coverage by ITAAC Index No. 75, item 6.i) and remove as-built from ITAAC Index No. 75, Item 6.iii), discussed above. The staff finds that SNCs proposed c hanges, as limited by the NRC, do not delete any technical requirements, do not impact the abilit y of an SSC to perform its function, or impact safety and the reasonable assurance finding. Based on these findings the staff concludes that with the proposed changes, as modified by NRC, the ITAAC will be sufficient to verify that the facility has been constructed and will operate in accordance with the license, the AEA, and the Commissions rules and regulations. Therefore, within the scope of this license amendment, the NRC concludes that 10 CFR 52.97(b) is satisfied. Thus, SNCs proposed changes, as modified, are acceptable.

COL Appendix C Table 2.1.3-2 (for ITAAC Index No. 75) is revised as follows:

Table 2.1.3-2 Inspections, Tests, Analyses, and Acceptance Criteria

No. ITAAC Design Commitment Inspections, Tests, Acceptance Criteria No. Analyses

75 2.1.03.06.i 6. The seismic i) Inspection will be The seismic Category I Category I equipment performed to verify that the equipment identified in Table identified in Table seismic Category I 2.1.3-1 (except fuel 2.1.3-1 can withstand equipment identified in assemblies, rod cluster seismic design basis Table 2.1.3-1 (except fuel control assemblies, gray rod loads without loss of assemblies, rod cluster cluster assemblies, and safety function. control assemblies, gray rod incore instrument QuickLoc cluster assemblies, and assemblies) is located on incore instrument QuickLoc the Nuclear Island.

assemblies) is located on the Nuclear Island.

ii) (no changes) ii) (no changes)

iii) Inspection will be iii) A report exists and performed for the existence concludes that the as-built of a report verifying that the equipment including as-built equipment including anchorage is seismically anchorage is seismically bounded by the tested or bounded by the tested or analyzed conditions.

analyzed conditions. This inspection must be performed on the as-built equipment except for the fuel assemblies, rod cluster control assemblies, gray rod cluster assemblies, and incore instrument QuickLoc assemblies.

3.1.3 ITAAC Index No. 515 (2.5.01.03e)

For ITAAC Index No. 515, the Design Commitment is [t]he sensor s identified on Table 2.5.1-3 are used for DAS input and are separate from those being used b y the PMS and plant control system, with an Acceptance Criteria of, [t]he sensors identif ied on Table 2.5.1-3 are used by DAS and are separate from those being used by PMS and plant con trol system. The inspection to be performed to support confirmation of the diver se actuation system (DAS) input sensors states that [i]nspection of the as-built system will b e performed.

The DAS is a non-safety-related system that provides a diverse backup to the safety-related Protection and Safety Monitoring System (PMS) in the AP1000 des ign. The DAS is designed to meet the applicable requirements and protective functions estab lished by 10 CFR 50.62, Requirements for reduction of risk from anticipated transients without scram (ATWS) events for light-water-cooled nuclear power plants. 10 CFR 50.62 provide s, in part, that ATWS equipment (e.g., DAS) must be independent and diverse (from sensor output to the final actuation device) from the existing reactor trip system (i.e., PMS). The non-saf ety-related DAS is not required to prevent or mitigate the effects of design basis accidents or to provide or perform safety-related functions or safety-related features to mitigate or prevent aga inst the effects of design basis accidents. As such, the DAS must meet applicable criteria in 1 0 CFR Part 50 Appendix A, GDC 24, Separation of protection and control systems, which r equires, in part, The protection system shall be separated from c ontrol systems to the extent that failure of any single control system component or channel, or fail ure or removal from service of any single protection system component or channel which is common to the control and protect ion systems leaves intact a system satisfying all reliability, redundancy, and independence requirements of the protection system.

In LAR 21-001 SNC requested ITAAC Index No. 515 be revised to r emove the as-built location attribute for these inspections only for the core exit temperat ure (CET) sensors on the following basis:

[T]he inspection requirement is of the as-built system. The sensors identified in Table 2.5.1-3 include the core exit temperature sensors. However, pu rsuant to UFSAR Subsection 4.4.6.1, the sensors are to be installed within the core and thus, cannot be installed in their final operational location prior to constitu tion of a core by the initial fuel load. Accordingly, as currently written, ITAAC Index No. 515 c annot be closed in conformance with the interpretation and understanding that as-built SSCs must be installed in their final operational location prior to ITAAC cl osure. Furthermore, the inspection of these sensors with this interpretation of as-bui lt does not meet the UFSAR Subsection 14.3.2.2 selection criteria for ITAAC that, t he ITAAC do not include any inspections, tests, or analyses that are dependent upon con ditions that exist only after fuel load, and should be excluded from ITAAC Index No. 5 15.

Construction drawings illustrate the DAS sensor flow and indic ation architecture. An inspection of Quality Release and Certificate of Conformance do cument, construction drawings, and completed construction records is performed to co nfirm that the sensors identified in the table were installed per the DAS sensor input requirements and are separate from those being used by the PMS and plant control sys tem with the exception of the core exit temperature sensor installation.

The licensee also described measures, other than ITAAC 515, use d to confirm that the CET sensors used by the DAS are separate than those being used by t he PMS and the plant control system. For example, UFSAR Section 14.2.9.1.13, Incore Instru mentation System Testing, describes provisions to ensure DAS CET sensors display properly in the main control room.

That testing involves the DAS CET sensors and sensor flow data along with other testing to verify the DAS sensors are operating properly.

The staff reviewed the information in Enclosure 1 of LAR 21-001 and finds that the incore instrument thimble assemblies (IITAs), which contain the CET se nsors identified in COL Appendix C, Table 2.5.13 (tag number IIS-009, IIS-013, IIS-030, and IIS-034), cannot be installed in their final as-built location to support plant ope ration until after the 10 CFR 52.103(g) finding has been made and initial fuel load has been completed. Therefore, applying the as-built requirement in this ITAAC to these components does not m eet the UFSAR Subsection 14.3.2.2 selection criteria for ITAAC. The proposed change doe s not remove the DAS CET sensors from verification by ITAAC; the DAS CET sensors are sti ll part of the system to be inspected and are within the scope of, and must meet, the accep tance criteria. The proposed change only means that the inspection may be completed without the DAS CET sensors being installed. Verification that the DAS CET sensors are separate from those provided to PMS and PLS based on review of design, construction, and installation d ocuments is reasonable.

Additionally the subsequent reactor vessel assembly and fuel lo ad is adequately controlled and verified during the initial test program; this addresses instal lation of the DAS CET sensors. The staff finds that SNCs proposed change is acceptable. The prop osed change does not delete any technical requirements or impact the ability of an SSC to p erform its function. Based on these findings, the staff concludes that the LARs proposed rev isions to ITAAC Index No. 515 will be sufficient to verify that the facility has been constru cted and will operate in accordance with the license, the provisions of the AEA, and the Commission s rules and regulations.

Therefore, the NRC concludes that 10 CFR 52.97(b) is satisfied and SNCs proposed changes to be acceptable.

COL Appendix C Table 2.5.1-4 is revised (for ITAAC Index No. 515) as follows:

Table 2.5.1-4 Inspections, Tests, Analyses, and Acceptance Criteria

No. ITAAC No. Design Commitment Inspections, Tests, Acceptance Criteria Analyses 515 2.5.01.03e 3.e) The sensors Inspection of the as-The sensors identified on identified on Table 2.5.1-3 built system will be Table 2.5.1-3 are used by are used for DAS input performed except for DAS and are separate and are separate from the core exit from those being used by those being used by the temperature sensor the PMS and plant control PMS and plant control installation. system.

system.

3.1.4 ITAAC Index No. 565 (2.5.05.02.i)

Following license amendments 85 and 84 for VEGP Units 3 and 4, respectively, ITAAC Index No. 565 represents a consolidation of former ITAAC Index Nos. 5 65 (2.5.05.02.i), 566 (2.5.05.02.ii), 567 (2.5.05.02.iii), 568 (2.5.05.03a.i), and 56 9 (2.5.05.03a.ii). Current ITAAC Index No. 565 verifies the seismic and environmental qualificat ion of components identified in COL, Appendix C, Table 2.5.5-1. The only components listed in Table 2.5.5-1 are the IITAs, which are labeled Incore Thimble Assemblies in Table 2.5.5-1.

As described in UFSAR Subsection 4.4.6.1, Incore Instrumentati on, the in-core instrumentation system (IIS) consists, in part, of 42 in-core i nstrument thimble assemblies (IITAs) which are installed within various fuel assemblies with in the core. Each IITA is composed of multiple self-powered detectors (SPDs) and one core exit thermocouple assembly, contained within individual inner sheaths and collectively an o uter sheath. The IITAs provide Class 1E core exit thermocouple inputs to the Protection and Sa fety Monitoring System (PMS),

non-Class 1E core exit thermocouple inputs to the DAS, and non-Class 1E SPD signals to the on-line power distribution monitoring system (OPDMS).

SNCs current ITAAC Index No. 565 is described below.

There are two design commitments in ITAAC Index No. 565 and a s et of ITAAC activities related to each design commitment. The first design commitment, item 2, states: The seismic Category I equipment identified in Table 2.5.5-1 can withstand seismic design basis dynamic loads without loss of safety function. Three ITAAC verificati on activities are required to demonstrate that this design commitment is met: (1) verificati on the installed components are located on the Nuclear Island (ITAAC item 2.i)), (2) seismic qu alification of the components (ITAAC item 2.ii)), and (3) verification that the as-built comp onents are bounded by the seismic qualification (ITAAC item 2.iii)).

The second design commitment, item 3.a), states: The Class 1E equipment identified in Table 2.5.5-1 as being qualified for a harsh environment can withstan d the environmental conditions that would exist before, during, and following a design basis a ccident without loss of safety function, for the time required to perform the safety function. Two ITAAC verification activities are required to demonstrate that this design commitment is met: (1) qualification of the Class 1E equipment for a harsh environment (item 3.a.i)) and (2) veri fication that the as-built Class 1E components are bounded by the qualification (item 3.a.ii)).

SNC in LAR 21-001 requested ITAAC Index No. 565, item 2.i) be d eleted and that items 2.iii) and 3.a.ii) be revised to remove the as-built attribute from the inspection, test, and analysis requirements and the acceptance criteria based on the following discussion:

ITAAC Index No. 565, Acceptance Criteria item 2.i), requires th at the seismic Category I equipment identified in Table 2.5.5-1 is located on the Nuclear Island. The seismic Category I equipment identified in Table 2.5.5-1 is the Incore Thimble Assemblies (at least three assemblies in each core quadrant). The incore thim ble assemblies are stored in protected environments off the nuclear island until i t is time for loading the initial core. These include the core exit temperature sensors of ITAAC Index No. 515.

Storing such sensitive equipment (on the Nuclear Island) would unnecessarily subject it to potential damage. Further, fuel loading and loading of the incore thimble assemblies will not and cannot occur until the equipment is on the Nuclea r Island. Thus, this is an unnecessary requirement that does not serve the underlying purp ose of ITAAC which are required to be completed prior to the initial fuel load.

The seismic Category I equipment identified in Table 2.5.5-1 i s the Incore Thimble Assemblies (at least three assemblies in each core quadrant). As discussed above, this equipment will not be installed in their final operational loca tions until after the initial core is loaded with fuel assemblies in the reactor vessel. Accordin gly, these portions of ITAAC Index No. 565 cannot be closed in conformance with the in terpretation and understanding that as-built SSCs must be installed in their f inal operational location prior to ITAAC closure. Furthermore, the inspection of the Inc ore Thimble Assemblies with this interpretation of as-built do not meet the UFSAR Su bsection 14.3.2.2 selection criteria for ITAAC that the ITAAC do not include any inspectio ns, tests, or analyses that are dependent upon conditions that exist only after fuel load. Prior to fuel load, these attributes are verified by vendor inspections and testing; foll owing the 52.103(g) finding, associated fuel loading and precritical testing will be perform ed in accordance with COL License Condition 2.D.(3).

An inspection is conducted to confirm that the seismic categor y I equipment identified in Table 2.5.5-1, the Class 1E Incore Thimble Assemblies, were manufactured per the qualified design. The inspection verifies the equipment make/m odel/serial number, as well as the as-designed anchorage point to the integrated grid assembly. An EQ Reconciliation Report (EQRR) is completed to verify the seismic Category I equipment listed in the Table, including anchorage, is seismically bounde d by the tested or analyzed conditions, IEEE Standard 344-1987, and NRC Regulatory Guide (RG) 1.100, Revision 2, Seismic Qualification of Electric and Mechanical E quipment for Nuclear Power Plants.

UFSAR Subsection 3.11.5 identifies that the environmental quali fication (EQ) files developed by the reactor vendor are maintained as applicable fo r equipment and certain post-accident monitoring devices that are subject to a harsh en vironment. The contents of the qualification files are discussed in Section 3D.7. Appe ndix 3D, Subsection 3D.7.2.1 indicates that equipment is identified by manufacture r, model or model series, and reference to other documents describing or depicting its co nstruction, configuration, and modifications that are uniquely necessary after manufacture to its application in the AP1000 plant design. Subsection 3D.7.2.2 goes on to address i nstallation requirements, noting So that the qualification represents the in-plant condition, the method of installation, as specified in Section 1.2 of Attachme nt A, is in accordance with the supplier's installation instructions. Differences unique t o safety-related applications in the AP1000 design are included, with appropriate reference t o drawings, technical manual supplements, or mandatory modification packages.

In addition, Quality Control reviews the work package completio n to confirm that the equipment is installed in a manner that is consistent with the as-tested/as-analyzed configuration.

SNCs proposed change deletes ITAAC Index No. 565, item 2.i) be cause the Incore Thimble Assemblies, listed in COL, Appendix C, Table 2.5.5-1, are not located on the nuclear island until after fuel loading is authorized. This change proposes t o remove the ITAAC requirement to inspect and verify that the location of the Incore Thimble Asse mblies is on the nuclear island prior to the NRCs 10 CFR 52.103(g) finding, which aligns with t he UFSAR 14.3.2.2 criteria, as previously mentioned, and its removal does not change any seism ic qualifications for, or verification of, the Incore Thimble Assemblies. The Incore Thi mble Assemblies will be located on the nuclear island following authorization to load fuel and will constitute part of the reactor system prior to operation. Thus, the deletion of ITAAC Index N o. 565, item 2.i), is acceptable.

The staff reviewed the information in Enclosure 1 of LAR 21-001 and finds that the Incore Thimble Assemblies cannot be installed in their final location to support plant operation until after the 10 CFR 52.103(g) finding has been made and the initia l fuel load of the reactor vessel.

Thus, ITAAC Index No. 565, item 2.iii), cannot be completed in the IITAs as-built condition, and the as-built requirement should not have been included fo r this ITAAC. With the proposed revisions, the licensee would still be required to ver ify that the IITAs, including anchorage, are seismically bounded by the seismic qualification performed under ITAAC Index No. 565, item 2.ii). This seismic qualification explicitly con siders the IITA installation method and location. Consequently, the ITAAC as revised provide suffi cient assurance that the IITAs can withstand seismic design basis loads without loss of safety function. Additionally, the subsequent reactor vessel assembly and fuel load is adequately controlled and verified during the initial test program; this addresses installation of the II TAs. For these reasons, the licensees proposal to remove as-built from item 2.iii) is ac ceptable.

SNCs proposed changes to ITAAC Index No. 565, item 3.a.ii) mai ntain the design functions of these systems and do not substantively change what the existing ITAAC is intended to verify, except that the ITAAC would be completed without the IITAs havi ng been installed. ITAAC Index No. 565, item 3.a.ii), cannot be completed in the IITAs as-built condition, and the as-built requirement should not have been included for this ITAAC. With the proposed revisions, the licensee would still be required to verify that the IITAs, and the associated wiring, cables, and terminations, are bounded by the harsh environment qualific ation performed under ITAAC Index No. 565, item 3.a.i). As stated in the LAR and UFSAR, th e qualification represents the in-plant condition and the method of installation. Consequently, the ITAAC as revised provide sufficient assurance that the IITAs can withstand the environme ntal conditions that would exist before, during, and following a design basis accident without l oss of safety function, for the time required to perform the safety function. Additionally, the sub sequent reactor vessel assembly and fuel load is adequately controlled and verified during the initial test program; this addresses installation of the IITAs. The staff also finds that SNCs pro posed changes do not delete any technical requirements, impact the ability of an SSC to perform its function, or impact safety and the reasonable assurance finding with respect to qualification of the components. Therefore, the EQ requirements defined in 10 CFR 50.49, GDC 2, and GDC 4 c ontinue to be met. For these reasons, the licensees proposal to remove as-built fro m item 3.a.ii) is acceptable.

Based on these findings the staff concludes that the LARs prop osed revisions to ITAAC Index No. 565 will continue to be sufficient to verify that the facil ity has been constructed and will operate in accordance with the license, the provisions of the A EA, and the Commissions rules and regulations. Therefore, within the scope of this license a mendment, the NRC concludes that 10 CFR 52.97(b) is satisfied. Thus, SNCs proposed change s shown below are acceptable.

COL Appendix C Table 2.5.5-2 is revised (for ITAAC Index No. 565) as follows for Unit 3 (the Unit 4 COL is revised identically, except that the Amendme nt No. is 186):

Table 2.5.5-2 Inspections, Tests, Analyses, and Acceptance Criteria

No. ITAAC No. Design Inspections, Tests, Acceptance Criteria Commitment Analyses 565 2.5.05.02.i 2. The seismic i) Not used per i) Not used per Category I Amendment No. 188 Amendment No. 188 The equipment Inspection will be seismic Category I identified in Table performed to verify that equipment identified in 2.5.5-1 can the seismic Category I Table 2.5.5-1 is located withstand seismic equipment identified in on the Nuclear Island.

design basis Table 2.5.5-1 is located dynamic loads on the Nuclear Island.

without loss of ii) (no changes) ii) (no changes) safety function. iii) Inspection will be iii) A report exists and performed for the concludes that the as-existence of a report built equipment verifying that the as-built including anchorage equipment including is seismically anchorage is seismically bounded by the tested bounded by the tested or analyzed or analyzed conditions. conditions.

3.a) The Class 1E i) (no changes) i) (no changes) equipment ii) Inspection will be ii) A report exists and identified in Table performed of the as-concludes that the as-2.5.5-1 as being built Class 1E built Class 1E equipment qualified for a harsh equipment and the and the associated environment can associated wiring, wiring, cables, and withstand the cables, and terminations identified in environmental terminations located in Table 2.5.5-1 as being conditions that a harsh environment. qualified for a harsh would exist before, environment are during, and bounded by type tests, following a design analyses, or a basis accident combination of type tests without loss of and analyses.

safety function, for the time required to perform the safety function.

3.1.5 ITAAC Index No. 570 (2.5.05.03b)

COL Appendix C, Section 2.5.5, In-Core Instrumentation System, Design Description 3.b) states: [t]he Class 1E cables between the Incore Thermocouple elements and the connector boxes located on the integrated head package have sheaths. Th e design requirement is verified in ITAAC Index No. 570. For ITAAC 570, the design com mitment is [t]he Class 1E cables between the Incore Thermocouple elements and the connect or boxes located on the integrated head package have sheaths, with an acceptance crite rion of [t]he as-built Class 1E cables between the Incore Thermocouple elements and the connect or boxes located on the integrated head package have sheaths. To verify that the acce ptance criterion is met, ITAAC Index No. 570 requires that [i]nspection of the as-built syste m will be performed.

SNC in LAR 21-001 requested that Design Description 3.b) and IT AAC Index No. 570 (2.5.05.03b) be revised for clarity and to remove as-built fr om the ITAAC based on the following information:

(page 6 of 19)The incore instrumentation system consists of in core instrument thimble assemblies, which house fixed incore detectors, core exit therm ocouple assemblies contained within an inner and outer sheath assembly, and associ ated signal processing and data processing equipment. There are 42 incore instrument thimble assemblies:

each is composed of multiple fixed incore detectors and one the rmocouple.

(page 13 of 19) The core exit temperature sensor is located in the Incore Instrument Thimble Assembly (IITA) which is inserted into the core. The o ther end of the IITA is outside the reactor vessel head at the QuickLoc. The IITA conn ects to the head area cable assembly which is then routed through and around the Inte grated Head Package (IHP), across the cable bridge, through the IHP cable rack asse mbly and connects with its matching cable mounted on the connector plate, which is par t of the Operating Deck Connector Panel.

To give effect to the scope of this ITAAC expressed by NRC sta ff, the clarity changes include replacing the phrase Incore Thermocouple elements wit h Core Exit Temperature sensors for consistency with other ITAAC nomenclat ure, removing the location information and clarifying the scope by replacing c onnector boxes with connector plates consistent with terms used in the FSAR and r eplacing system with Class 1E cables so that the inspection language is consistent with the Design Commitment and the Acceptance Criteria.

The ITAAC inspections for equipment which cannot be inspected i n its final location include review of documentation such as the system design speci fications, records of inspection of the components performed by the manufacturer prio r to shipment to the plant site, Quality Release and Certificate of Conformance docu mentation, construction drawings, and completed construction records, including those p erformed on-site, to the extent possible given that the facility design does not allow c ertain components to be installed in their final operational configuration until after fuel is installed in the vessel.

Design specifications require internal metallic sheaths that su rround and separate the individual Class 1E thermocouple wires from non-Class 1E detect or wires, which are contained within an external spiral wound sheath. The design s pecifications also include performance tests for overvoltage, insulation resistance, and c ontinuity. Successful test results indicate that the sheaths protect against credible sing le faults between the Class 1E and non-Class 1E signals.

The Quality Release and Certificate of Conformance document ver ifies the head area cable assembly acceptance test results. The Field Service Repo rt document verifies the head area cable assemblies were installed on the integrated hea d package in accordance with design drawings and installation specifications issued for construction, and work package requirements. An additional Quality Release a nd Certificate of Conformance document verifies that the invessel Class 1E cables were installed within the incore thimble assemblies in accordance with design drawing s and installation specifications and contains the incore thimble assemblies cable acceptance test results.

The Class 1E cables between the Core Exit Temperature sensors ( these sensors are located within the incore thimble assemblies) and the connector plates (as described above, these connectors are beyond the integrated head package on the operating deck connector panel) are inspected to verify that the design specif ication and installation specifications are satisfied, to enable each cable to convey the safety-related core exit thermocouple signals to the PMS.

The inspections are performed and documented in accordance with manufacturer and vendor quality verification programs. The results of the inspe ctions are documented in support of the ITAAC 2.5.05.03b Completion Packages. The inspe ctions confirm that the Class 1E cables between the Core Exit Temperature sensors a nd the connector plates have sheaths.

The staff reviewed the information in Enclosure 1 of LAR 21-001 for the changes proposed to clarify Design Description 3.b) and ITAAC Index No. 570 and to remove as-built from the ITAACs required inspection and corresponding acceptance criter ia.

As described in the LAR and in UFSAR Subsections 4.4.6.1, Inco re Instrumentation, 9.1.4.2.2.2, Phase II - Reactor Disassembly, and 3.9.7.2, De sign Description, the CET sensors are located within the IITAs located in the in-core the rmocouple elements, and the head area cables are connected at the IHP connector plates and not boxes. Therefore, the staff finds the revised nomenclature is consistent with the terms use d in the plant UFSAR and the proposed nomenclature changes are acceptable.

The LAR states that installation of the IITAs, which include th e CET sensors and cables, requires fuel be installed in the reactor vessel. This means t hat the as-built requirement in the ITAAC, as applied to the CET sensors and associated Class 1E ca bles connecting to the IHP connector plates, is not consistent with the requirement in 10 CFR 52.103(g) that the NRC find that the acceptance criteria are met prior to operation (which includes fuel load), nor with the UFSAR Subsection 14.3.2.2 selection criteria for ITAAC that th e ITAAC do not include any inspections, tests, or analyses that are dependent upon conditi ons that exist only after fuel load. However, as described in the LAR, the IHP cable from th e operating deck connector panel to the incore instrument QuikLoc assemblies, has already been inspected and verified in its as-built location (on the IHP), so the staff finds the pr oposed change does not materially affect the verification required of this cable. The staff also noted that the ITAACs method of completing acceptance criteria is a visual inspection of the Cl ass 1E cables sheaths in the IITAs, which cannot be accomplished once the components have be en installed. Therefore, applying the as-built provision to the ITAAC does not provide any substantial benefit or added assurance that is not already obtained by the inspections and t esting performed at the vendor facilities for the IITAs.

Furthermore, the ITAAC process intended to verify the design re quirements, including the separation provision between Class 1E and non-Class 1E componen ts, is addressed in NUREG-1793, Revision 0, Final Safety Evaluation Report Related to Certification of the AP1000 Standard Design, dated September 2004. WCAP-17226, As sessment of Potential Interaction between the Core Exit Thermocouple Signals and the Self-Powered Detector Signals in the AP1000TM In-core Instrumentation System, Revision 2, addressed by anal ysis separation concerns of Class 1E and non-Class 1E cables within both the II TAs and head area cable assemblies, and was accepted by staff in NUREG-1793. This LAR does not affect the staffs evaluation in NUREG-1793 since the physical routing of those ca bles are not impacted. In addition, the LAR indicates that design specifications will be reviewed which contain performance tests for overvoltage, insulation resistance, and c ontinuity to ensure presence of the required sheaths and that no credible faults can occur betw een the Class 1E and non-Class 1E cables in both the IITAs and the head area cable assembly.

Based on the above discussion, the staff finds that SNCs propo sed changes do not delete or affect any applicable technical requirements, the method of ver ifying the sheaths on the Class 1E cables, or their design function. The staff also finds that the proposed changes do not impact the ability of the Class 1E cables to perform their safe ty functions, and that ITAAC Index No. 570 still verifies that the cables have sheaths. Specifica lly, the revised ITAAC will still be completed by inspection to verify that specified sheaths are pr esent on the cables by confirming design specifications requiring performance tests for overvolta ge, insulation resistance, and continuity to ensure that specified sheaths protect against cre dible single faults between the Class 1E and non-Class 1E cables. Additionally, the subsequent reactor vessel assembly and fuel load is adequately controlled and verified during the init ial test program; this addresses installation of the IITAs. Based on these above findings, the staff concludes that SNCs proposed changes to ITAAC Index No. 570 will continue to be suf ficient to verify that the facility has been constructed and will operate in accordance with the li cense, the provisions of the AEA, and the Commissions rules and regulations. Therefore, within the scope of this license amendment, the NRC concludes that 10 CFR 52.97(b) is satisfied. Thus, SNCs proposed changes to Design Description 3.b) and ITAAC Index No. 570 are acceptable.

COL Appendix C Table 2.5.5-2 is revised (for ITAAC Index No. 570) as follows:

Table 2.5.5-2 Inspections, Tests, Analyses, and Acceptance Criteria

No. ITAAC No. Design Commitment Inspections, Tests, Acceptance Criteria Analyses 570 2.5.05.03b 3.b) The Class 1E cables Inspection of the The as-built Class 1E cables between the Core Exit as-built system between the Core Exit Temperature sensors Incore Class 1E cables will Temperature sensors Incore Thermocouple elements and be performed. Thermocouple elements and the connector plates boxes the connector plates boxes located on the integrated located on the integrated head package have head package have sheaths.

sheaths.

3.2 EVALUATION OF EXEMPTION

INTRODUCTION

The regulations in Section III.B of Appendix D to 10 CFR Part 5 2 require a holder of a COL referencing Appendix D to 10 CFR Part 52 to incorporate by refe rence and comply with the requirements of Appendix D, including certified information in Tier 1 of the generic AP1000 DCD. Exemptions from Tier 1 information are governed by the ch ange process in Section VIII.A.4 of Appendix D of 10 CFR Part 52. Because SNC has iden tified changes to plant-specific Tier 1 information, with corresponding changes to the associated COL Appendix C information resulting in the need for a departure, an exemption from the certified design information within plant-specific Tier 1 material is required t o implement the LAR.

The Tier 1 information for which a plant-specific departure and exemption was requested is described above. The result of this exemption would be that SN C could implement the requested modifications to Tier 1 information, with correspondi ng changes to COL Appendix C.

Pursuant to the provisions of 10 CFR 52.63(b)(1), an exemption from elements of the design as certified in the 10 CFR Part 52, Appendix D, design certificati on rule is requested for the specified Tier 1 change described and justified in LAR 21-001. This is a permanent exemption limited in scope to the particular Tier 1 information specified.

As stated in Section VIII.A.4 of Appendix D to 10 CFR Part 52, an exemption from Tier 1 information is governed by the requirements of 10 CFR 52.63(b)( 1) and 52.98(f). Additionally,Section VIII.A.4 of Appendix D to 10 CFR Part 52 provides that the Commission will deny a request for an exemption from Tier 1 if it finds that the desig n change will result in a significant decrease in the level of safety otherwise provided by the desig n. Pursuant to 10 CFR 52.63(b)(1), the Commission may, upon application by an applica nt or licensee referencing a certified design, grant exemptions from one or more elements of the certification information, so long as the criteria given in 10 CFR 52.7, which in turn, refer ences 10 CFR 50.12, are met.

Also, the Commission must consider whether the special circumst ances which are defined by 10 CFR 50.12(a)(2) outweigh any potential decrease in safety du e to reduced standardization caused by the exemption.

Pursuant to 10 CFR 52.7, the Commission may, upon application b y any interested person or upon its own initiative, grant exemptions from the requirements of 10 CFR Part 52.

10 CFR 52.7 further states that the Commissions consideration will be governed by 10 CFR 50.12. In accordance with 10 CFR 50.12, an exemption ma y be granted when: (1) the exemption is authorized by law, will not present an undue risk to public health and safety, and is consistent with the common defense and security; and (2) specia l circumstances are present.

10 CFR 50.12(a)(2) lists six special circumstances for which an exemption may be granted. It is necessary for one of these special circumstances to be present in order for NRC to consider granting an exemption request. The licensee stated that the re quested exemption meets the special circumstances of 10 CFR 50.12(a)(2)(ii). That subsecti on defines special circumstance when [a]pplication of the regulation in the particular circums tances would not serve the underlying purpose of the rule or is not necessary to achieve t he underlying purpose of the rule.

The staffs analysis of each of the required exemption criteria is presented below.

3.2.1 AUTHORIZED BY LAW

This exemption would allow the licensee to implement the amendm ent described above. This is a permanent exemption limited in scope to particular Tier 1 inf ormation. Subsequent changes to this plant-specific Tier 1 information, and corresponding ch anges to Appendix C, or any other Tier 1 information would be subject to the exemption process sp ecified in Section VIII.A.4 of Appendix D to 10 CFR Part 52 and the requirements of 10 CFR 52. 63(b)(1). As stated above, 10 CFR Part 52, Appendix D, Section VIII.A.4 allows the NRC to grant exemptions from one or more elements of the Tier 1 information. The NRC staff has det ermined that granting the licensees proposed exemption will not result in a violation of the AEA or the Commissions regulations. Therefore, as required by 10 CFR 50.12(a)(1), the exemption is authorized by law.

3.2.2 NO UNDUE RISK TO PUBLIC HEALTH AND SAFETY

As discussed above in the technical evaluation, the proposed ch anges comply with the NRCs substantive safety regulations. Therefore, there is no undue r isk to the public health and safety.

3.2.3 CONSISTENT WITH COMMON DEFENSE AND SECURITY

The proposed exemption would allow changes as described above i n the technical evaluation, thereby departing from the AP1000 certified (Tier 1) design inf ormation. The changes do not alter or impede the design, function, or operation of any plant structure, system, or component (SSC) associated with the facilitys physical or cyber security, and therefore does not affect any plant equipment that is necessary to maintain a safe and secure plant status. In addition, the changes have no impact on plant security or safeguards. Theref ore, as required by 10 CFR 50.12(a)(1), the staff finds that the common defense and security is not impacted by this exemption.

3.2.4 SPECIAL CIRCUMSTANCES

Special circumstances, in accordance with 10 CFR 50.12(a)(2)(ii ), are present, in part, whenever application of the regulation in the particular circum stances would not serve the underlying purpose of the rule or is not necessary to achieve t he underlying purpose of the rule.

Special circumstances are present in the particular circumstanc es discussed in LAR 21-001 because the application of the specified Tier 1 information doe s not serve the underlying purpose of the rule or is not necessary to achieve the underlyi ng purpose of the rule. As discussed above, the changes to ITAAC Index Nos. 75, 515, 565, and 575 regarding the location of components on the nuclear island and the performanc e of certain ITAAC verifications in the as-built condition only have a practical effect for co mponents that cannot be installed before fuel load. Because these requirements are intended to r eflect the installed condition of the components and because ITAAC must be satisfied before fuel load, the application of the requirements to those components would not serve the underlying purpose of these requirements. In addition, the remaining ITAAC verifications c ombined with post-fuel load verifications serve the underlying purpose of the ITAAC to veri fy that the installed components meet the design requirement. The proposed changes also elimina te an unnecessary requirement by deleting a functional arrangement ITAAC (Index N o. 68) that is either duplicated by other ITAAC (for those components that can be installed befo re fuel load) or cannot be completed as stated and will be verified by other means (for th ose components that cannot be installed before fuel load). The proposed changes also revise the nomenclature for certain components to align with the current licensing basis. The prop osed changes do not adversely affect any function or feature used for the prevention and miti gation of accidents or their safety analyses. The changes described above do not impact the abilit y of any SSC to perform its function or negatively impact safety. The revisions to the Tie r 1 information and corresponding changes to Appendix C will continue to meet applicable regulato ry requirements. Therefore, for the above reasons, the staff finds that the special circumstanc es required by 10 CFR 50.12(a)(2)(ii) for the granting of an exemption from the Tier 1 information exist.

3.2.5 SPECIAL CIRCUMSTANCES OUTWEIGH REDUCED STANDARDIZATION

This exemption would allow the implementation of changes to COL Appendix C and corresponding Tier 1 information based on the LAR. The effect of the changes in LAR 21-001 is to (1) remove requirements to complete certain ITAAC in the as -built location for components that cannot be installed prior to the 10 CFR 52.103(g) finding, (2) eliminate a functional arrangement ITAAC that is duplicated by other ITAAC, with the e xception of those components that cannot be installed before the 10 CFR 52.103(g) finding, ( 3) eliminate a requirement to verify equipment is located on the nuclear island, when that equipment cannot be installed before the 10 CFR 52.103(g) finding, and (4) revise nomenclatur e used in the ITAAC to improve clarity. Also, for those components that cannot be installed b efore the 10 CFR 52.103(g) finding, there are post-fuel load requirements that adequately address verifications regarding the locations and installed condition of these components. The se changes, as modified by NRC, do not alter any technical requirements or impact the abil ity of an SSC to perform its function. The design functions of the system associated with t his request will continue to be maintained because the associated revisions to the Tier 1 infor mation support the design function of the associated SSCs. The only practical effect of the changes is to briefly delay un til after fuel load the verification of the location and installed condition of components that cannot be installed before fuel load. Therefore, there is no meaningf ul reduction in standardization, and the safety impact that may result from any reduction in standardization is minimized, because the proposed change does not result in a reduction in the level of safety. In addition, the special circumstances discussed above outweigh any potential decrease i n safety from reduced standardization because the only practical effect of the change s is to eliminate a contradiction between the ITAAC as written and the requirement to complete IT AAC before scheduled fuel load. Thus, the staff finds that 10 CFR Part 52.63(b)(1) is sa tisfied.

3.2.6 NO SIGNIFICANT REDUCTION IN SAFETY

The exemption request proposes to depart from the certified des ign by allowing changes discussed above in the technical evaluation. These changes, as modified by NRC, do not alter any technical requirements or impact the ability of an SSC to p erform its function. The changes will not modify the design or modi fy the operation of any syste ms or equipment, there are no new failure modes introduced by these changes, and the level of safety provided by the current structures, systems, and components will be unchanged. As disc ussed above, the only practical effect of the changes is to briefly delay until after fuel load the verification of the location and installed condition of components that cannot be i nstalled before fuel load.

Therefore, based on the foregoing reasons and as required by 10 CFR Part 52, Appendix D, Section VIII.A.4, the staff finds that granting the exemption w ould not result in a significant decrease in the level of safety otherwise provided by the desig n.

For the reasons given above, the standards for an exemption fro m the specified Tier 1 information have been satisfied.

3.3

SUMMARY

In LAR 21-001, SNC proposed to make changes that would affect t he COL Appendix C and corresponding PS-DCD Tier 1 information. None of the above pro posed changes represent changes to the design or operation of the plant. Therefore, th e requirements defined in 10 CFR 50.49, 10 CFR 50.55a, GDC 2, GDC 4, and GDC 24 continue to be m et. The staff finds that with the changes, as modified by the NRC, the ITAAC continue to be sufficient to verify that the facility has been constructed and will be operated in accordanc e with the license, the AEA, and NRC rules and regulations. Therefore, within the scope of this license amendment, the NRC finds that 10 CFR 52.97(b) is satisfied. The NRC documented it s review of the above changes in Section 3.1 of this safety evaluation and finds the changes acceptable.

4.0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION

The NRC staff published its proposed no significant hazards con sideration determination in the Federal Register on September 3, 2021 (86 FR 49572). Under its regulations, th e Commission may issue an amendment and make it immediately effective, notwi thstanding the pendency before it of a request for a hearing from any person, where it has made a final determination that no significant hazards consideration is involved.

The NRCs regulation in 10 CFR 50.92(c) states that the NRC may make a final determination, under the procedures in 10 CFR 50.91, that a license amendment involves no significant hazards consideration if operation of the facility, in accordan ce with the amendment, would not:

(1) involve a significant increase in the probability or conseq uences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

An evaluation of the issue of no significant hazards considerat ion is presented below:

1. Does the proposed amendment involve a significant increase i n the probability or consequences of an accident previously evaluated?

Response: No.

The proposed revisions have been found to continue to provide t he required functional capability of the safety systems for previously eval uated accidents and anticipated operational occurrences. The affected system is no t an initiator of any accident analyzed in the Updated Final Safety Analysis Report ( UFSAR), nor do the changes involve an interface with any SSC accident initiator or initiating sequence of events, and thus, the probabilities of the accidents evaluated in the UFSAR are not affected. The proposed changes do not involve a change to any mitigation sequence or the predicted radiological releases due to postulat ed accident conditions; thus, the consequences of the accidents evaluated i n the UFSAR are not affected.

The UFSAR describes the analyses of various design basis transi ents and accidents to demonstrate compliance of the design with the acceptance cri teria for these events. The acceptance criteria for the various events are bas ed on meeting the relevant regulations and general design criteria and are a func tion of the anticipated frequency of occurrence of the event and potential radiological consequences to the public. The revised ITAAC maintains the plant conditions, and thus, maintains the frequency designation and consequence level as previously evalu ated.

Therefore, the proposed amendment does not involve a significan t increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed revisions have been found to continue to confirm t he required functional capability of the safety systems for previously eval uated accidents and anticipated operational occurrences. The proposed revisions do not change the function of the related systems, and thus, the changes do not i ntroduce a new failure mode, malfunction or sequence of e vents that could adversely affect safety or safety-related equipment.

Therefore, the proposed amendment does not create the possibili ty of a new or different kind of accident from any accident previously evaluat ed.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed revisions have been found to continue to provide t he required functional capability of the safety systems for previously eval uated accidents and anticipated operational occurrences. The proposed revisions do not change the function of the related systems nor significantly affect the ma rgins provided by the systems. No safety analysis o r design basis acceptance limit/c riterion is challenged or exceeded by the requested changes.

Therefore, the proposed amendment does not involve a significan t reduction in a margin of safety.

Based on the above evaluation, the staff concludes that the thr ee standards of 10 CFR 50.92(c) are satisfied. Therefore, the staff has made a final determina tion that no significant hazards consideration is involved for the proposed amendment and that t he amendment should be issued as allowed by the criteria contained in 10 CFR 50.91.

5.0 STATE CONSULTATION

In accordance with the Commission's regulations in 10 CFR 50.91 (b)(2), on September 10, 2021, the Georgia State official was notified of the proposed i ssuance of the amendment. The State official had no comments.

6.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installatio n or use of a facility component located within the restricted area as defined in 10 C FR Part 20, Standards for Protection Against Radiation. The staff has determined that t he amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant incre ase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideratio n, and there has been no public comment on such finding as published in the Federal Register on September 3, 2021 (86 FR 49572). And as stated above, the NRC is making a final no sign ificant hazards consideration determination for this license amendment. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c) (9). Pursuant to 10 CFR 51.22(b),

no environmental impact statemen t or environmental assessment need be prepared in connection with the issuance of the amendment.

Because the exemption is necessary to allow the changes propose d in the license amendment, and because the exemption does not authorize any activities oth er than those proposed in the license amendment, the environmental consideration for the exem ption is identical to that of the license amendment. Accordingly, th e exemption meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs to be prepared in connection with the issuance of the exemption.

7.0 CONCLUSION

The staff has determined that pursuant to Section VIII.A.4 of A ppendix D to 10 CFR Part 52, the exemption (1) is authorized by law, (2) presents no undue risk to the public health and safety, (3) is consistent with the common defense and security, (4) pre sents special circumstances, and (5) does not significantly reduce the level of safety at the li censees facility. Also, the staff has determined that the special circumstances for the exemption out weigh any decrease in safety that may result from a reduction in standardization caused by t he exemption. Therefore, the staff grants the licensee an exemption from the Tier 1 informat ion requested by the licensee.

The staff has concluded, based on the considerations discussed in Section 3.1 that there is reasonable assurance that: (1) the health and safety of the pu blic will not be endangered by the proposed changes, (2) the changes are in compliance with the Co mmissions regulations, and (3) the issuance of the amendment will not be inimical to the c ommon defense and security or to the health and safety of the public. Therefore, the staff find s the changes proposed in this license amendment acceptable.

8.0 REFERENCES

1. Southern Nuclear Operating Company, Vogtle Electric Generati ng Plant Units 3 and 4, Request for License Amendment and Exemption: Clarification of ITAAC Regarding Invessel Components (LAR 21-001), August 24, 2021 (ADAMS Acces sion No.

ML21236A305).

2. AP1000 Design Control Document, Revision 19, June 13, 2011 ( ADAMS Accession No. ML11171A500).

3. Combined License NPF-91 for Vogtle Electric Generating Plant Unit 3, Southern Nuclear Operating Company (ADAMS Accession No. ML14100A106).

4. Combined License NPF-92 for Vogtle Electric Generating Plant Unit 4, Southern Nuclear Operating Company (ADAMS Accession No. ML14100A135).

5. Vogtle Electric Generating Plant Units 3 and 4, Updated Fina l Safety Analysis Report, Revision 10, June 14, 2021 (ADAMS Accession No. ML21179A130).

6. Nuclear Energy Institute (NEI) 08-01, Industry Guideline fo r the ITAAC Closure Process Under 10 CFR Part 52, Revision 5 - Corrected, dated June 30, 2 014 (ADAMS Accession No. ML14182A158).

7. Regulatory Guide 1.215, Guidance for ITAAC Closure Under 10 CFR Part 52, Revision 2, July 20, 2015 (ADAMS Accession No. ML15105A447).

8. WCAP-17226, Revision 2, Assessment of Potential Interaction between the Core Exit Thermocouple Signals and the Self-Powered Detector Signals in t he AP1000TM In-core Instrumentation System, dated August 25, 2010 (ADAMS Accession No. ML102390521).

9. NUREG-1793, Supplement 2, Final Safety Evaluation Report Re lated to Certification of the AP1000 Standard Plant Design Docket No.52-006, dated August 5, 2011 (ADAMS Accession No. ML112061231).

10. IEEE 603-1991, IEEE Standard Criteria for Safety Systems f or Nuclear Power Generating Stations.