ML14148A492

ISI Program Alternative FNP-ISI-ALT-16. Version 1

ML14148A492
Person / Time
Site:	Farley
Issue date:	05/28/2014
From:	Pierce C Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-14-07-15
Download: ML14148A492 (16)

Text

{{#Wiki_filter:Charles R. Pierce Southern Nuclear Regulatory Affairs Direclor Operating Company, Inc. 40 Inverness Center Parkway Post Office Box 1295 Birmingtiam, Aiat}ania 35201 Tel 2D5.992.787Z Fax 205.992.7601 SOUTHERN.^ COMPANY May 28, 2014 Docket Nos.: 50-348 NL-14-0715 50-364 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Joseph M. Farley Nuclear Plant - Units 1 and 2 ISI Program Alternative FNP-ISI-ALT-16. Version 1 Ladies and Gentlemen; In accordance with the provisions of 10 CFR 50.55a(a)(3)(ii), Southern Nuclear Operating Company (SNC) hereby requests NRC approval of an American Society of Mechanical Engineers (ASME) Section XI code alternative to address the leakage examination for the Class 2 reactor vessel flange leak-off lines for both Units 1 and 2 of the Joseph M. Farley Nuclear Plant (Farley), The System Leakage Test of Class 2 pressure retaining components is cited in Table IWC-2500-1, Examination Category C-H, Item No. C7,10, and required to be performed during each inspection period. Table lWC-2500-1 requires that system leakage tests be conducted at a pressure not less than the pressure corresponding to 100% rated reactor pow/er in accordance with IWC-5221. This requested alternative (FNP-ISI-ALT-16, Version 1,0) is necessitated by the impracticality of conformance to requirements of Table IWC-2500-1 of the ASME Section XI Code. The proposed alternative examination for Class 2 reactor vessel flange leak-off piping pressure test at Farley is predicated on the ASME Code Case N-805 which has yet to be approved by the NRC. However, the NRC has previously determined that similar requests from the commercial nuclear industry are acceptable for these circumstances. Enclosed is a detailed discussion of the need and justification for the use of an alternate approach for the Class 2 leakage testing of the subject vessel flange leak-off lines needed for conformance to IWC-5221 of the Code. In order to support restart of Farley Unit 2 following the 2R23 Refueling Outage, expedited approval of the proposed alternative is requested by October 1, 2014.

U. S. Nuclear Regulatory Commission NL-14-0715 Page 2 This letter contains no NRC commitments. If you have any questions, please contact Ken McElroy at (205) 992-7369. Sincerely, ^ C.R. Pierce Regulatory Affairs Director CRP/TWS/lac

Enclosure:

Proposed Alternative in accordance with 10 CFR 50.55a(a)(3)(ii) cc: Southern Nuclear Operating Company Mr. S. E, Kuczynski, Chairman, President & CEO Mr. D. G. Bost, Executive Vice President & Chief Nuclear Officer Ms. C. A. Gayheart, Vice President - Farley Mr. B. L. Ivey, Vice President - Regulatory Affairs Mr. D. R. Madison, Vice President - Fleet Operations Mr. B. J. Adams, Vice President - Engineering RTYPE: CFA04.054 U. S. Nuclear Regulatory Commission Mr. V. M. McCree, Regional Administrator Mr. S. A. Williams, NRR Project Manager - Farley Mr. P. K. Niebaum, Senior Resident Inspector - Farley Mr. J. R. Sowa, Resident Inspector - Farley

ENCLOSURE JOSEPH M. FARLEY NUCLEAR PLANT ISI PROGRAM ALTERNATIVE FNP-ISI-ALT-16, VERSION 1 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(ii)

SOUTHERN NUCLEAR OPERATING COMPANY - FNP-1 and FNP-2 FNP-ISI-ALT-16, VERSION 1.0 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(ii) Plant Site-Unit: Joseph M. Farley Nuclear Plant (FNP) - Units 1 and 2 Interval-Interval Dates: 4"" IS! - December 1, 2007 through November 30, 2017 Requested Date for Approval and Basis: Approval is requested by October 1,2014 to permit performance of the proposed alternative pressure test during the 23rd Refueling Outage of FNP Unit 2. This proposed alternative will remain applicable through the remainder of the fourth 181 Interval. ASME Code Components Affected: FNP Units: Unit 1 and Unit 2 ASI^E Code Class: Code Class 2

References:

ASME Section XI, Table IWC-2500-1 and IWC-5220 Examination Category: C-H (All Pressure Retaining Components) Item Number: C7.10 Component: Reactor Pressure Vessel Flange Leak-off Piping with NPS 3/8", 3/4", and 1"

Description:

3/8" Tubing is A-213 Gr 304 and has a minimum wall thickness of 0.065" Design Pressure is 2485 psig at 650T 3/4" and 1" Pipe is SA-376 Type 304 or 316 and Sch. 160 Design Pressure is 2485 psig at 650T

Applicable Code Edition and Addenda

ASME Section Xi, 2001 Edition through the 2003 Addenda Applicable Code Requirements: System Leakage Test of Class 2 pressure retaining components per Table IWC-2500-1, Examination Category C-H (Item No. C7.10) are to occur each inspection period. Paragraph IWC-5221 indicates that system leakage tests are to be conducted at the system pressure obtained while the system, or portion of E-1

SOUTHERN NUCLEAR OPERATING COMPANY - FNP-1 and FNP-2 FNP-ISI-ALT-16, VERSION 1.0 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(li) the system, is in service performing its normal operating function or at the system pressure developed during a test conducted to verify system operability. Per IWC-5222(a), the pressure retaining boundary includes the portion of the system required to operate or support the safety function up to and including the first normally closed valve. Background and Reason for Request: Southern Nuclear recently became aware of enforcement actions associated with failure of licensees to perform or adequately perform system leakage tests on the reactor vessel flange leak-off piping at a pressure less than that required by ASME Section XI, Subarticle IWC-5220. The NRG provided additional information regarding these instances in Information Notice 2014-02 (IN 2014-02), Failure to Properly Test Reactor Vessel Flange Leak-off Lines, dated February 25,2014, with the expectation that licensees would review the information for applicability to their facilities and consider actions, as appropriate, to avoid similar problems. Southern Nuclear Operating Company (SNO) performed a review of the applicable operational experience (OE) and IN 2014-02, along with the design of the FNP reactor vessel head flange leak-off piping and determined that FNP is also subject to the same limitations that resulted in the instances cited in the OE and IN 2014-02. The FNP reactor vessel flange leak-off piping is part of the reactor coolant pressure boundary leak detection system and is designed in accordance with the requirements of Regulatory Guide 1.45, Revision 0, dated May 1973, as specified in Appendix 3A of the FNP Final Safety Analysis Report (FSAR). Per Section 3A of the FNP FSAR, the design function of the reactor coolant boundary leakage detection system is to forewarn the operator of minor leakage that may develop during normal operation; however, it does not perform a safeguard function and is not required to operate during or after a seismic event. The RPV flange and head at FNP Units 1 and 2 are sealed by two hollow, metallic o-rings. The o-rings are made of a silver-plated, nickel-chromlum-lron alloy and are installed between the mating surfaces of the reactor vessel flange and the closure head to prevent reactor coolant leakage between the vessel and the closure head. The o-rings are self-energized by reactor coolant pressure acting through the slots on the inside of the o-rings causing them to expand. Two grooves are machined into the reactor vessel flange for the o-rings and the corresponding retainer rings. Two additional grooves are machined into the closure head flange for the retainer rings, which are used to support the o-rings during their installation and removal. The o-rings are fastened to the closure head by a mechanical connection to facilitate removal. One o-ring is sufficient to create the seal but the second serves as a backup. Seal leakage detection is facilitated by means of two leak-off connections: one between the inner and outer o-rings and one outboard the outer o-ring. A E-2

SOUTHERN NUCLEAR OPERATING COMPANY - FNP-1 and FNP-2 FNP-ISI-ALT-16, VERSION 1.0 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(M) manual isolation valve is installed on each leak-off connection to allow isolation of the inner or outer leak-off lines in the unlikely event of failure of either of the o-rings. Downstream of the manual valves, the leak-off lines connect to a common line equipped with: (1) an air-operated isolation valve (AOV) that allows operators in the main control room to isolate leakage from the reactor vessel flange o-rings, and (2) a resistance temperature detector (RTD) that annunciates in the main control room in the event an elevated temperature is detected in the leak-off piping. The leak-off piping is located inside the biological shield and sections of this piping are considered to be inaccessible due to its proximity to other piping and equipment in the vicinity of the RPV flange that are known to be inaccessible. A depiction of the typical RPV head flange leak-off piping configuration is shown in Attachment 1. As indicated earlier, IWC-5221 states the system leakage test shall be conducted at the system pressure obtained while the system, or portion of the system, is in service performing its normal operating function or at the system pressure developed during a test conducted to verify system operability. IWC-5222(a) indicates that the pressure retaining boundary includes only those portions of the system required to operate or support the safety function up to and including the first normally closed valve (including a safety or relief valve) or valve capable of automatic closure when the safety function is required. The configuration of the reactor vessel leak-off lines is such that the piping is not capable of being pressure tested at normal reactor operating pressure (2235 psig) without: (1) changes to the reactor vessel head flange leak-off piping design, or (2) intentionally failing the inner and outer RPV flange o-rings. Neither of these options are considered viable and the bases for this determination are provided below. Design changes to facilitate code compliant test In order to perform a Code compliant system leakage test, a design change would be required to: (1) install a new test connection to which a pump skid could be attached, or (2) modify the existing branch connection to replace valve QV018. Valve QV018 is a Kerotest valve and would have to be replaced due to its design that precludes pressurization in the reverse direction. A Code compliant system leakage test would be ideally performed during a routine refueling outage with the head removed in order to: (1) allow any trapped air to escape through the holes on the vessel flange, and (2) to preclude equalization of the inner o-ring by subjecting it to reverse pressurization. In order to perform the test with the head removed, threaded plugs would be required to be installed in both the inner and outer taps to act as the pressure boundary. Use of plugs for this purpose poses an additional concern associated with the potential introduction of foreign material into the vessel during installation. E-3

SOUTHERN NUCLEAR OPERATING COMPANY - FNP-1 and FNP-2 FNP-ISI-ALT-16, VERSION 1.0 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(ll) Although not an ideal lest condition, performing the test with the vessel head installed would require that a vent be added to allow any air to escape when pressurizing the reactor vessel flange leak-off piping. The vent line would likely need to be added close to the reactor vessel flange where dose rates are elevated (approximately 230 mrem/hr). In addition, the outer tap would need to be plugged as well since it is not located between the inner and outer o-ring and Is open to the atmosphere. As such, the radiological exposure required to implementthe design changes necessary to facilitate pressure testing of the reactor vessel flange leak-off piping and the actual performance of the pressure test required by ASME Section XI, Subarticle IWC-5220 during each inspection period, represents a hardship without a compensating increase in the level of quality or safety beyond the proposed alternative described below. Intentionally failino the RPV flange o-rinos The option to intentionally fail the RPV flange inner o-ring to establish normal operating pressure and temperature on the leak-off piping is not considered a viable option due to the increased dose that would result from the need to replace the inner o-ring. Replacement of the inner o-ring following the pressure test requires removal and reinstallation of the reactor vessel head a second time during an outage, regardless as to whether the test was perfonned at the beginning of the outage or at the end of the outage. The typical dose associated with removal of the reactor vessel head and reinstallation based on previous FNP outages is approximately 3.8 REM. To meet the requirements of ASME Section XI, Subarticle IWC-5220, a second removal and reinstallation of the reactor vessel head would be required and result in an additional radiological dose of 3.8 REM. Removal of the reactor vessel head is a time and labor intensive activitythat includes de-tensioning the RPV studs, disconnecting the various cables and head vent piping, and moving the RPV head to the RPV head stand. Removal and reinstallation of the RPV head a second time during the outage for the purpose of performing the pressure test required by ASME Section XI, Subarticle IWC-5220, results in an increased dose associated with the outage of approximately 3.8 REM and an increased probability of occurrence of a heavy load drop accident. As such, compliance with the requirements of ASME Section XI, Subarticle IWC-5220 represents a hardship without a compensating increase in the level of quality or safety beyond the proposed alternative described below. Proposed Alternative and Basis for Use: In accordance with the provisions of 10 CFR 50.55a(a)(3)(ii), FNP proposes to examine the accessible portions of the Class 2 reactor vessel head flange leak-off piping once per inspection period. Accessible portions of the leak-off piping shall be examined using the VT-2 visual examination method and will be performed by certified VT-2 examiners. The test shall be conducted at ambient conditions after the refueling cavity has been flooded to its minimum water level E-4

SOUTHERN NUCLEAR OPERATING COMPANY - FNP-1 and FNP-2 FNP-ISI-ALT-16, VERSION 1.0 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(il) for refueling operations of 23 ft above the top of the RPVflange for at least four (4) hours. A static pressure of approximately 10 psig is expected to be experienced at the top of the RPV flange with a minimum of 23 ft of borated water above. In order to eliminate any air that may be trapped in the piping, FNP will flush the leak-off piping prior to performing the visual examination. There are no vents near the top elevation of the pipingthat would allow venting of air in the pipe. The only location to permitventing without performing a design modification is valve QV018, which is located on the lowest section of the piping. This valve is located downstream of the two manual isolation valves and can be seen in Attachments 1,2 and 3. To ensure the piping is water solid prior to the beginning of the four hour pressurization hold time, the blindflange of the QV018 valve will be removed and the QV018 valve will be opened allowing water to flow from the piping. The flow of water will be allowed for approximately five minutes. Once a steady stream of water is obtained, the QV018 valve will be closed and the blind flange will be reinstalled. This evolution is expected to purge the piping of air since this will draw water from the refueling cavity to flush out air. In addition, the time it takes to fill the refueling cavityto 23 ft above the RPVflange and the flow of water through the pipe, coupled with air's buoyancy, will allow air to exit the piping through the top of the RPV flange, which is open to the atmosphere. With this process, there is the potential for small pockets of air to be trapped in the piping. However, if a through-wall indication exists at the same location of an air pocket, eventually the air would leak out, allowing the water to leak out as well. The preceding process for filling the RPV flange leak-off piping will help ensure the piping is water solid prior to the four hour pressurization hold time. The Class 2 portion of piping originating from the reactor vessel flange and terminating at the last boundary isolation valve (AOV QV019) is required to be examined. An excerpt from the Unit 1 and Unit 2 Reactor Coolant System (RCS) P&ID is shown in Attachments 2 and 3, respectively, depicting the reactor vessel leak-off piping subject to the proposed alternate examination method. For segments of the line that are inaccessible for direct VT-2 visual inspection, examination will include inspection of the surrounding areas below the line for evidence of leakage as permitted by IWA-5241 (b) of the ASME Section XI Code, 2001 Edition through 2003 Addenda. It is not clear which sections of the leak-off piping are fully accessible for visual examination since this piping has never been examined individually; however, it has been included in the scope of the Class 1 leakage exam performed at the end of each refueling outage. Therefore, visual examination of the piping has been historically performed in some manner. There are approximately 99 linear feet of pipe on Unit 1 and 137 linear feet of pipe on Unit 2 that make up this leak-off piping. Approximately 66 linear feet of pipe on Unit 1 and 98 linear feet of pipe on Unit 2 are visually accessible, either E-5

SOUTHERN NUCLEAR OPERATING COMPANY - FNP-1 and FNP-2 FNP-ISt-ALT-16, VERSION 1.0 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(ll) by direct or remote visual examination. The sections of piping that is believed to be completely inaccessible for visual examination during refueling cavity flood up conditions starts where the piping connects to the reactor vessel and continues to where the primary shield wall ends. This section of piping is considered inaccessible for reasons: (1) access to this piping is not available while the refueling cavity is flooded (i.e., access to this piping is only accessible through the "sandbox covers" which are in the refueling cavity), (2) a section of piping is located between the reactor vessel and the interior of the primary shield wall, making it inaccessible, and (3) the final section goes through the primary shield wall with the RGB hot leg piping, through the restraint rings. Attachments 4 and 5 are provided to illustrate the reactor vessel flange leak-off piping layout. Attachment 4 provides a section view of the area under the "sandbox cover" that is considered inaccessible when the refueling cavity is flooded while Attachment 5 provides a plan view of both the accessible and inaccessible sections of the reactor vessel head flange leak-off piping system. A VT-2 visual examination of the accessible areas will be performed on the piping subjected to the static pressure from the head of water when the reactor cavity is filled for at least four hours. The proposed alternative examination will be perfonned within the frequency specified by table IWC-2500-1 for a System Leakage Test (once each inspection period). As stated previously, the reactor vessel head flange leak-off piping does not perform a safeguard function and as such, is not required to function for safe shutdown of the plant or decay heat removal. The reactor vessel flange leak-off piping is essentially a leakage collection/detection system and would only function as a Class 2 pressure boundary in the unlikely event that the inner o-ring fails, or for the outer line in the event both the inner and outer o-rings fail, thereby pressurizing the piping. Ifthe inner and/or outer o-rings were to fail, the piping would be exposed to a maximum pressure of 2235 psig, which is below the design pressure (2485 psig) for the leak-off piping. If the flange leak-off piping were to fail in conjunction with failure of both the inner and outer reactor vessel head o-rings, leakage would be identified by either;

• performance of an PCS water inventory balance in accordance with Technical Specification 3.4.13, PCS Operational Leakage, or
• via the RCS Leakage Detection Instrumentation required by Technical Specification 3.4.15.

Leakage from the reactor vessel head flange leak-off piping would be considered unidentified leakage and, as such, be limited to a rate of 1 gallon per minute (GPM) by Technical Specification 3.4.13. Per Technical Specification Bases 3.4.13,1 GPM of unidentified leakage is allowed as a reasonable minimum detectable amount that the containment air monitoring E-6

SOUTHERN NUCLEAR OPERATING COMPANY - FNP-1 and FNP-2 FNP-iSI-ALT-16, VERSION 1.0 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(li) and containment sump level monitoring equipment can detect in a reasonable time period. Accordingly, leal<age resulting from failure of the inner o-ring and the reactor vessel head flange leak-off piping would be monitored and result in entry into Limiting Condition of Operation 3.4.13 should it exceed the 1 GPM allowable threshold for unidentified RCS leakage. Further, significant leakage due to a through-wall leak of the reactor vessel flange leak-off piping would be expected to clearlyexhibit boric acid residue accumulation that would be discernible during the proposed alternate VT-2 visual examination that will be performed. Additionally, the static head developed with the leak detection line filled with water and the minimum allowable time the line is filled with water, as described in the proposed alternative, will allowfor the detection of gross indications in the piping. Based on the above, compliance with the requirements of ASME Section XI, Subarticle IWC-5220 represents a hardship or unusual difficulty that does not result in a compensating increase in level of quality or safety as described in 10 CFR 50.55a(a)(3)(ii). Use of the proposed alternative in lieu of the pressure test required by ASME Section XI, Subarticle IWC-5220, will provide an acceptable level of quality and safety based on the ability to detect leakage from the vessel head flange leak-off piping and the corresponding limitations of Technical Specification 3.4.13 described above. Duration of Proposed Alternative: The alternative is requested for the remainder of the current Fourth Inservice Inspection Interval, which began December 1,2007 and is scheduled to end on November 30, 2017 for both Unit 1 and 2. Precedents:

1. Arkansas Nuclear One, Unit 2, Fourth Inspection Interval Alternative, Request for Relief from American Society of Mechanical Engineers (ASME)

Code, Section XI - Request for Relief AN02-ISI-015, approved by the NRC in a letter dated June 27,2013 (ADAMS Accession No. ML13161A241)

2. Callaway Plant, Unit 1, Third Inspection Interval Alternative, Proposed Alternative to ASME Section XI Requirements for Leakage Testing of Reactor Pressure Vessel Head Flange Leakoff Lines (Relief Request I3R-14, approved by the NRC in a letter dated August 13, 2013 (ADAMS Accession No. ML13221A091)

3. Palo Verde Nuclear Generating Station, Units 1,2, and 3, Third Inspection Interval Alternative, Request for Relief from the American Society of Mechanical Engineers (ASME) Code, Section XI, Reactor Vessel Head Flange Seal Leak Detection Piping - Relief Request No. 49, approved by E-7

SOUTHERN NUCLEAR OPERATING COMPANY - FNP-1 and FNP-2 FNP-ISI-ALT-16, VERSION 1.0 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(li) the NRG in a letter dated April 4, 2013 (ADAMS Accession No. ML13085A254)

4. Diablo Canyon, Units 1 and 2, Third Inspection Interval Alternative, Request for Approval of an Alternative to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section XI Pressure Test Requirements for Class 1 Reactor Vessel Flange Leakoff Lines, approved by the NRC in a letter dated September 12, 2013 (ADAMS Accession No. ML13192A354)

5. Dresden, Units 2 and 3, Fifth Inspection Interval Alternative, Request for Relief for Exemption from Pressure Testing Reactor Pressure Vessel Head Flange Seal Leak Detection System, approved by the NRC in a letter dated September 30, 2013 (ADAMS Accession No. ML13258A003)

6. Vermont Yankee, Fourth Inspection Interval Alternative, Alternative to System Leakage Test for the Reactor Pressure Vessel Head Flange Leak-off Lines, approved by the NRC in a letter dated March 1, 2013 (ADAMS Accession No. ML13055A009)

7. Vogtle Electric Generating Plant, Units 1 and 2, Third Inspection Interval Alternative, Request for Alternative VEGP-ISI-ALT-10, Version 1, Regarding System Leakage Test of Reactor Pressure Vessel Flange Leak-off Piping, approved by the NRC in a letter dated March 19, 2014 (ADAMS Accession No. ML14078A331)

8. Edwin I. Hatch Nuclear Plant, Units 1 and 2, Safety Evaluation of Relief Request HNP-ISI-ALT-18 for the Fourth 10-Year Inservice Inspection Interval, dated Febmary 12, 2014 (ADAMS Accession No. ML14038A192).

References:

1. ASME Code Case N-805, Alternative to Class 1 Extended Boundary End of Interval or Class 2 System Leakage Testing of Reactor Vessel Flange O-ring Leak Detection System was issued to the 2010 Edition of the ASME Section XI Code and is listed in Supplement 6 for Code Cases. However, Code Case N-805 has not been approved by the NRC and is not identified in Regulatory Guide 1.147, Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1.

Status: Pending NRC approval. E-8

SOUTHERN NUCLEAR OPERATING COMPANY - FNP-1 and FNP-2 FNP-lSI-ALT-16, VERSION 1.0 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(ii) ATTACHMENT 1 "Simplified Schematic of RPV Flange Leak-off Line 0-Ring Seals I Non Code Class 2 K QV018 QV017B 41. QV019 3/8" QV017A Reactor Coolant Drain Tank

             *The schematic is a depiction of the Unit 1 piping. Unit 2 piping differs only by the name of the QV017 valves and the location of the 3/4" to 3/8" reducers.

E-9

SOUTHERN NUCLEAR OPERATING COMPANY - FNP-1 and FNP-2 FNP-ISI-ALT-16, VERSION 1.0 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(ij) ATTACHMENT 2 Excerpt from FNP Unit 1 P&ID D175037 ai-tisu i/i

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SOUTHERN NUCLEAR OPERATING COMPANY - FNP-1 and FNP-2 FNP-ISI-ALT-16, VERSION 1.0 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(ii) ATTACHMENT 3 Excerpt from FNP Unit 2 P&ID D205037 p} LOC. iirocMJ-10 Ch'C^U 10 niK riCMPTER 2-TW-40I J.076A 2-RC-2M1R a-TA780L 2-B032 Q-/019 V2 X i/fl' PPE TUK iilJiPTDi 6- CC 379A NOTE 26 1/a' PPE 10 Tiff fits 5/4 V

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SOUTHERN NUCLEAR OPERATING COMPANY - FNP-1 and FNP-2 FNP-ISI-ALT-ie, VERSION 1.0 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(ii) ATTACHMENT 4 Excerpt from FNP Drawing D176275, depicting inaccessible leak-off piping. (Note: Unit 2 Typical) roi ae*crof. Sandbox Cover

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SOUTHERN NUCLEAR OPERATING COMPANY - FNP-1 and FNP-2 FNP-ISI-ALT-16, VERSION 1.0 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(il) ATTACHMENT 5 Excerpt from FNP Drawing 0176277, depicting inaccessible leak-off piping. (Note: Unit 2 Typical) IjOOP2 LOOP Note: Inaccessible leak-off piping identified by clouds. E-13}}