View source for NL-13-2590, Units 1 and 2, ISI Program Alternative VEGP-ISI-ALT-10, Version 1

{{Adams
| number = ML14016A488
| issue date = 01/16/2014
| title = Units 1 and 2, ISI Program Alternative VEGP-ISI-ALT-10, Version 1
| author name = Pierce C
| author affiliation = Southern Co, Southern Nuclear Operating Co, Inc
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000424, 05000425
| license number = 
| contact person = 
| case reference number = NL-13-2590
| document type = Letter type:NL
| page count = 21
}}

=Text=
{{#Wiki_filter:Charles R. Pierce             Southern Nuclear Regulatory Affairs Director   Operating Company, Inc.
40 Inverness Center Parkway Post Office Box 1295 Birmingham. Alabama 35201 Tel 205.992.7872 Fax 205.992.7601 SOUTHERN << \
January 16, 2014                                                                COMPANY Docket Nos.: 50-424                                                        NL-13-2590 50-425 U. S. Nuclear Regulatory Commission ATTN : Document Control Desk Washington , D. C. 20555-0001 Vogtle Electric Generating Plant- Units 1 and 2 lSI Program Alternative VEGP-ISI-ALT-10, 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 for the leakage examination of the Class 2 reactor vessel flange leak-off piping for both Units 1 and 2 of the Vogtle Electric Generating Plant (VEGP). This 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 is required to be performed once an inspection period. Table IWC-2500-1, subparagraph IWC-5221 indicates that system leakage tests shall be conducted at the system pressure obtained while the system, or portion of the system, is in service performing its normal operation function or at the system pressure developed during a test conducted to verify system operability. The requested alternative (VEGP-ISI-ALT-10, Version 1.0) is necessitated by the impracticality of conformance to these specifications of the subject Code.
The proposed alternative examination for Class 2 reactor vessel flange leak-off piping pressure test at VEGP is predicated on the ASME Code Case N-805 which has yet to be approved by the NRC. However, the NRC has previously approved the methods described in ASME Code Case N-805 in response to similar requests from operators of commercial reactors in lieu of testing at pressures corresponding to 100% rated reactor power. of this letter provides 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 piping in lieu of the methods described in IWC-5221 of the Code. An expedited approval of the use of the proposed alternative is requested by March 16, 2014, to support plant restart following the VEGP 1R18 Refueling Outage which is currently scheduled to begin March 16, 2014.
Enclosures 2 and 3 contain piping isometric drawings. It is important to note that the originals of these drawings contain the following statement:

U. S. Nuclear Regulatory Commission NL-13-2590 Page2 "This document contains proprietary, confidential , and/or trade secret information of the subsidiaries of the Southern Company or of third parties. It is intended for use only by employees of, or authorized contractors of, the Southern Company.
Unauthorized possession , use, distribution, copying , dissemination, or disclosure of any portion hereof is prohibited."
SNC does not request exclusion from the public domain via 10 CFR 2.390 for these drawings. For Enclosures 2 and 3 of this letter, the drawings have been modified to not include the above statement.
If you have any questions regarding this request, please contact Mr. G. K.
McElroy at (205) 992-7369.
Respectfully submitted ,
C. R. Pierce Regulatory Affairs Director CRP/CLN/Iac

==Enclosures:==
: 1. Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(ii)
: 2. VEGP Isometric Drawing 1K4-1201-054-04
: 3. VEGP Isometric Drawing 2K4-1201-054-04 cc:   Southern Nuclear Operating Company Mr. S. E. Kuczynski, Chairman , President & CEO Mr. D. G. Bost, Executive Vice President & Chief Nuclear Officer Mr. T. E. Tynan , Vice President- Vogtle Mr. B. L. lvey, Vice President- Regulatory Affairs Mr. B. J. Adams, Vice President- Fleet Operations RType: CVC7000 U.S. Nuclear Regulatory Commission Mr. V. M. McCree, Regional Administrator Mr. R. E. Martin, NRR Senior Project Manager- Vogtle Mr. L. M. Cain , Senior Resident Inspector- Vogtle

Vogtle Electric Generating Plant lSI Program Alternative VEGP-ISI-ALT-10, Version 1 Enclosure 1 Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(ii)

SOUTHERN NUCLEAR OPERATING COMPANY- VEGP-1 and VEGP-2 VEGP-ISI-ALT-10, VERSION 1.0 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(ii)
Plant Site-Unit:
Vogtle Electric Generating Plant (VEGP) - Units 1 and 2 Interval-Interval Dates:
3rd lSI Interval- May 31, 2007 through May 30, 2017 Requested Date for Approval and Basis:
Approval is requested by March 16, 2014 to permit performance of the proposed alternative pressure test during the 181h Refueling Outage of VEGP Unit 1. This proposed alternative would also be applicable through the remainder of the third lSI Interval.
ASME Code Components Affected:
VEGP Units:                Unit 1 and Unit 2 ASME 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 SA-213 Gr TP304L or TP316L and has a minimum wall thickness of 0.065" Design Pressure is 2485 psig at 680'F 3/4" and 1" Pipe is SA-312 Gr TP304L and Sch. 160 Design Pressure is 2485 psig at 680'F

===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 the system, is in service performing its normal operating function or at the E1 - 1

SOUTHERN NUCLEAR OPERATING COMPANY- VEGP-1 and VEGP-2 VEGP-ISI-ALT-10, VERSION 1.0 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(ii) 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 determined, through the use of industry operating experience, that VEGP is susceptible to an issue being identified in the industry regarding compliance with ASME Section XI for examination of the reactor vessel flange leak-off piping. Until recently, VEGP was not aware that the visual examination being performed on the reactor pressure vessel (RPV) leak-off piping was not in accordance with Examination Category C-H of Table IWC-2500-1. SNC had recently learned through a review of industry violations for various nuclear utilities that VEG P was susceptible to receiving the same or similar violation due to the matter in which the RPV leak-off piping was being or not being examined. The issue of not examining the RPV leak-off piping is a relatively new issue within the industry and with the proposed alternative, VEGP is seeking to come back into full compliance with the Code.
The RPV flange and head at VEGP Units 1 and 2 are sealed by two metallic a-rings. These gaskets are of the hollow self-energizing type in which pressure of the fluid being sealed enters the interior of the gasket. The a-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. An inner and outer o-ring leak detection system alerts the control room of any leakage that may occur. One tap is machined in the vessel between the two a-rings. The second tap is located outside the outer o-ring. The two taps are piped to a resistance thermal detector that alarms in the control room if leakage occurs.
The alarm setpoint for both units is 17o*F. The inner and outer taps have manual isolation valves prior to joining into a common header. The outer tap is normally isolated. The common piping is routed to the reactor coolant drain tank (RCDT) . In the common line, there is an air operated isolation valve that can be closed by the control room in the event of a seal failure. To monitor for leakage outside the outer seal, local manual operation of the isolation valves is necessary to isolate the inner seal tap line and to un-isolate the outer seal tap line. The control room would then be required to re-open the common air operated isolation valve to monitor the outer o-ring seal integrity. A majority of the leak-off piping is located inside the bioshield/missile shield with a small portion located in the RPV annulus (lnservice Inspection Gallery). A depiction of the typical configuration of the piping in question 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 E1 -2

SOUTHERN NUCLEAR OPERATING COMPANY- VEGP-1 and VEGP-2 VEGP-ISI-AL T-1 0, VERSION 1.0 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(ii) 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), unless the inner and outer o-ring seals fail or are intentionally failed. The only other viable option would be a design change to perform a Code compliant system leakage test.
If a design change were to be implemented with the intent to hydrostatically pressurize the piping, a new test connection would need to be added for the pump skid to be attached. The new test connection would be required unless the valve on the branch line coming off the 3/8" line was to be replaced. This valve would have to be removed or replaced since these valves are Kerotest valves and may not permit reverse flow. This test would ideally be scheduled with the head removed to allow any trapped air to escape through the holes on the vessel flange. Performing the test with the head removed will require the installation of a threaded plug in both the inner and outer taps, (something to act as the pressure boundary) which poses a concern for the potential introduction of foreign material to the vessel internals during the implementation of the modification and any reoccurring installations. Although not an ideal test condition, if the test were to be performed with the vessel head installed, a vent would need to be added to allow any air to escape. The vent line would likely need to be added close to the reactor vessel flange where dose rates are elevated (approximately 100 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. The end result of these additions would require numerous man-hours in the installation of the modifications that would result in additional radiological exposure without a commensurate increase in the level of quality and safety.
The option to intentionally fail the inner seal of the vessel flange 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 seal.
The increase in accumulated dose due to the need to replace the failed o-rings would come from the requirement to remove and reinstall the vessel head a second time (because of the failed o-ring test) during the outage, regardless of when the pressure test would be performed. Reactor vessel head removal is a very intensive process that requires a significant amount of time due to the need to de-tension the RPV studs, disconnect various cables, disconnect head vent piping, and move the RPV head itself. Performing an additional RPV head removal and lift, coupled with RPV head reinstallation, results in additional exposure for potential errors to occur. Although a great deal of planning and attention is given to this work, it is nonetheless one of the most critical aspects in the execution of a refueling outage. The performance of an additional head removal and reinstallation is expected to result in an additional 3 REM of accumulated dose based on information from previous Vogtle outages.
E1 -3

SOUTHERN NUCLEAR OPERATING COMPANY- VEGP-1 and VEGP-2 VEGP-ISI-ALT-10, VERSION 1.0 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(ii)
The performance of the actual pressure test would take an estimated 20 to 30 minutes, at most, in either case. The quality of examinations in both cases would be the same. The evolution required to remove the vessel head and reinstall the head would take approximately 80 hours. The time it would take and the dose accumulated during head removal and reinstallation , along with the lack in difference in the quality of examination, does not result in an increase to the level of quality and safety that would justify the additional burden of an additional reactor vessel removal and installation.
Both operations would subject individuals performing the tasks to considerable dose without a commensurate increase to the health and safety of the public.
Accordingly, these options are not considered to be viable.
Proposed Alternative and Basis for Use:
In accordance with the provisions of 10 CFR 50.55a(a)(3)(ii), VEGP proposes to examine the Class 2 portion of the leak detection system consisting of the accessible portions of the RPV head flange o-ring leak-off piping once an inspection period. 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 for refueling operations of 23 ft above the top of the RPV flange 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.
The Class 2 portion of piping originating from the reactor vessel flange and terminating at the last boundary isolation valve (AOV HV8032) is required to be examined. Excerpts from the Unit 1 and Unit 2 Reactor Coolant System P&IDs are shown in Attachments 2 and 3, respectively. The reactor vessel leak-off piping subject to the proposed alternate examination method is highlighted in the two attachments.
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 under the Class 1 leakage exam performed at the end of every refueling outage.
Therefore, visual examination of the piping has been performed in some manner. There is an estimated 139 linear feet of pipe on unit 1 and 144 linear feet of pipe on unit 2 that make up this leak-off piping. An estimated 118 linear feet of pipe on unit 1 and 122 linear feet of pipe on unit 2 is visually accessible, either by direct or remote visual examination. The section of piping that is believed to be completely inaccessible for visual examination is where the piping connects to the reactor vessel. This section of piping is located above the annulus area of the reactor vessel, also referred to as the lnservice E1 -4

SOUTHERN NUCLEAR OPERATING COMPANY- VEGP-1 and VEGP-2 VEGP-ISI-ALT-10, VERSION 1.0 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(ii)
Inspection Gallery, and is deemed inaccessible due to it being partially encased in concrete and located between the concrete backed primary liner plate and the reactor vessel. An excerpt from plant drawing 1X2D48J01 0 is shown in Attachment 4 depicting this area.
In an attempt to communicate which section or portion of the piping that is accessible for visual examination, including what will have to be visually examined using the allowance of IWA-5241(b), the plant isometric drawings have been included as Enclosures 2 and 3 for unit 1 and unit 2, respectively.
Only a small section of the leak-off piping on unit 2 is insulated while the unit 1 piping is not insulated. This section of piping is located inside the lnservice Inspection Gallery. Sections of the piping have been highlighted with various colors to aid in understanding accessibility. The color codes are explained within each enclosure.
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 performed within the frequency specified by table IWC-2500-1 for a System Leakage Test (once each inspection period).
The reactor vessel flange seal leak-off piping is essentially a leakage collection/detection system and would only function as a Class 2 pressure boundary if the inner o-ring fails, or for the outer line if both the inner and outer o-rings fail, thereby pressurizing the piping. If the 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) of the leak-off piping.
If the inner o-ring should leak during the operating cycle it will be identified through the annunciation of an alarm in the control room. Downstream of the air operated valve in the common line is a temperature indicator with a setpoint of 170.F. Should temperature in this piping increase to 17o*F or greater, an alarm in the main control will be actuated. Upon receiving an alarm, operators will take action in accordance with annunciator response procedures. Subsequent Operator Actions from the annunciator response procedure are to (1) determine the vessel leak-off temperature, (2) monitor the Reactor Coolant Drain Tank level, (3) if Reactor Vessel Flange leakage is indicated, close the common line AOV, (4) refer to Technical Specification LCO 3.4.13, RCS Operational LEAKAGE, (5) if Reactor Vessel Flange leakage is indicated, enter containment to determine if gasket is leaking, (6) if inner o-ring leak is confirmed, place outer o-ring leak detection in service.
In the event significant leakage past the inner and outer o-rings or a through-wall leak develops in the reactor vessel flange leak-off piping, this leakage would be detectable through various means. One mean to determine leakage is through the performance of the RCS water inventory balance surveillance. The frequency of this surveillance is once within 12 hours after achieving steady state operation and at least once every 24 hours in modes 1, 2, and 3 and every 72 hours in Mode 4. Results from the RCS water inventory balance are evaluated against procedure acceptance criteria and, if necessary, actions are E1 -5

SOUTHERN NUCLEAR OPERATING COMPANY- VEGP-1 and VEGP-2 VEGP-ISI-ALT-10, VERSION 1.0 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(ii) taken to address values outside the acceptance criteria. Another means of detection is to utilize the containment atmosphere radioactivity monitors. The containment airborne particulate radioactivity monitor draws an air sample from containment via a sample pump, which then passes through a particulate filter with detectors. Particulate activity can be correlated with the coolant fission and corrosion product activities. The containment atmosphere gaseous radioactivity monitor draws air continuously from the containment atmosphere through a gas monitor. This sample stream flows continuously through a fix shielded volume where its activity is monitored. Gaseous radioactivity can be correlated with the gaseous activity of the reactor coolant. With these two methods for detecting leakage outside the reactor coolant pressure boundary, reasonable assurance is obtained that significant leakage past the inner and outer a-rings or a through-wall leak that occurs while the plant is operating will be detected well ahead of posing any significant risks to safe operation of the plant.
Any significant leakage due to a through-wall leak of the piping or failure of the outer a-ring would be expected to clearly exhibit 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 time the line is filled with water will allow for the detection of any gross indications in the line.
Performing the pressure test with only static head pressure in the piping does not reduce the margin of safety in operating the plant or detecting leaks.
Regardless of how the piping is examined, leakage that could occur during plant operation would still be detected. This conclusion is arrived at by one scenario in which it is assumed the leak-off piping develops a 100 percent through-wall leak.
As indicated earlier, the leak-off piping is designed so the control room is notified of an inner seal or outer seal ring leak through the annunciation of a control room alarm by the use of a temperature indicator. If a 100 percent through-wall leak of the leak-off piping were to occur, the temperature indicator would still be expected to detect elevated temperature, as long as the piping has not completely separated or sheared off to the point in which no fluid could traverse to the temperature indicator. If this scenario has occurred, although unlikely, the leakage would be detected through the performance of the RCS water inventory balance.
Since there is reasonable assurance that the proposed alternate examination will detect gross indications of leakage should any exist from this piping and that the examinations will occur once an inspection period, SNC requests authorization to use the proposed alternative pursuant to 10 CFR 50.55a(a)(3)(ii) on the basis that compliance with the specified requirement would result in hardship or difficulty without a compensating increase in the level of quality and safety.
E1 -6

SOUTHERN NUCLEAR OPERATING COMPANY- VEGP-1 and VEGP-2 VEGP-ISI-ALT-10, VERSION 1.0 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(ii)
Duration of Proposed Alternative:
The alternative is requested for the remainder of the current Third lnservice Inspection Interval, which began May 31, 2007 and is scheduled to end on May 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 13R-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 the NRC 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)

==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 D-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, E1 -7

SOUTHERN NUCLEAR OPERATING COMPANY- VEGP-1 and VEGP-2 VEGP-ISI-ALT-10, VERSION 1.0 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(ii)
Code Case N-805 has not been approved by the NRC and is not identified in Regulatory Guide 1.147, lnservice Inspection Code Case Acceptability, ASME Section XI, Division 1.
Status:
Pending NRC approval.
E1 -8

SOUTHERN NUCLEAR OPERATING COMPANY- VEGP-1 and VEGP-2 VEGP-ISI-AL T -10, 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 Non Code   Class 2 087 3/8" Reactor Coolant Drain Tank E1 -9

SOUTHERN NUCLEAR OPERATING COMPANY- VEGP-1 and VEGP-2 VEGP-ISI-AL T -10, VERSION 1.0 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(ii)
ATTACHMENT 2 Excerpt from VEGP Unit 1 P&ID 1X4DB111
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E1 - 10

SOUTHERN NUCLEAR OPERATING COMPANY- VEGP-1 and VEGP-2 VEGP-ISI-AL T -10, VERSION 1.0 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(ii)
ATTACHMENT 3 Excerpt from VEGP Unit 2 P&ID 2X4DB111
: 121 (G~e) (
SJS HOT lEG E1 - 11

SOUTHERN NUCLEAR OPERATING COMPANY- VEGP-1 and VEGP-2 VEGP-ISI-ALT -10, VERSION 1.0 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(ii)
ATTACHMENT 4 Excerpt from VEGP Drawing 1X2D48J010 (Note: Unit 2 Typical) 1./NIST~lJT        SEE SEC7. 2aS           ------~---
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: 61. 1&4- *11 Note: Portion of piping in red is considered inaccessible, portion of piping in blue is considered accessible.
E1 - 12

Vogtle Electric Generating Plant lSI Program Alternative VEGP-ISI-ALT-10, Version 1 Enclosure 2 VEGP Isometric Drawing 1K4-1201-054-04 to NL-13-2590 VEGP Isometric Drawing 1K4-1201-054-04 Explanation of Highlighted Colors Green: Depicts the section of piping that is contained in an inaccessible area above the lnservice Inspection Gallery/Reactor Vessel Annulus.
Lavender: Depicts the section of piping that is visible in the Reactor Vessel Annulus.
Red: Depicts the section of piping that penetrates or is located inside concrete wall. The enclosure indicates which walls are the Primary Shield walls or the Bioshield walls. This piping is visually inaccessible.
Violet: Depicts the section of piping that is visible and may have to use the allowance of IWA-5241(b). This section is located inside the Bioshield.
Blue: Depicts the section of piping that is located outside of the Bioshield and is capable of direct visual examination.

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                                               -------    ------------------------------------  to NL-13-2590 VEGP Isometric Drawing 2K4-1201-054-04 Explanation of Highlighted Colors Green: Depicts the section of piping that is contained in an inaccessible area above the lnservice Inspection Gallery/Reactor Vessel Annulus.
Lavender: Depicts the section of piping that is visible in the Reactor Vessel Annulus.
Lavender w/Red Stripe: Depicts the section of piping within the Reactor Vessel Annulus that is insulated.
Red: Depicts the section of piping that penetrates or is located inside concrete wall. The enclosure indicates which walls are the Primary Shield walls or the Bioshield walls. This piping is visually inaccessible.
Violet: Depicts the section of piping that is visible and may have to use the allowance of IWA-5241(b). This section is located inside the Bioshield.
Blue: Depicts the section of piping that is located outside of the Bioshield and is capable of direct visual examination.

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