NL-11-2187, HNP-ISI-ALT-14, Version 2.0 Temporary Non-Code Repair of Plant Service Water Piping

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HNP-ISI-ALT-14, Version 2.0 Temporary Non-Code Repair of Plant Service Water Piping
ML12027A047
Person / Time
Site: Hatch Southern Nuclear icon.png
Issue date: 10/27/2011
From: Ajluni M
Southern Nuclear Operating Co
To:
Office of Nuclear Reactor Regulation, Document Control Desk
References
HNP-ISI-ALT-14, Version 2.0, NL-11-2187
Download: ML12027A047 (56)
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{{#Wiki_filter:Mark J. Ajluni, P.E. Southern Nuclear Nuclear Licensing Director Operating Company, Inc. 40 Inverness Center Parkway Post Office Box 1295 Birmingham, Alabama 35201 Tel 205.992.7673 Fax 205.992.7885 October 27, 2011 SOUTHER.\\ COMPANY Docket Nos.: 50-321 NL-11-2187 U. S. Nuclear Regulatory Commission ATTN : Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant - Unit 1 HNP-ISI-AL T-14, Version 2.0 Temporary Non-Code Repair of Plant Service Water Piping Ladies and Gentlemen: Pursuant to 10 CFR 50.55a(a)(3)(ii), Southern Nuclear Operating Company (SNC) hereby requests Nuclear Regulatory Commission (NRC) approval of the enclosed Alternative HNP-ISI-ALT-14, Version 2, which proposes a temporary non-code repair for leaks discovered in the Hatch Nuclear Plant Unit 1 (H NP-1) Plant Service Water (PSW) System. During inspection of HNP-1 buried demineralized water transfer piping adjacent to the HNP-1 Reactor Building to address tritium leakage, two leaks were identified in a nearby run of PSW piping exposed by the excavation. An operability determination concluded that PSW system operability is maintained with respect to both structural stress and the potential for flow diversion due to leakage. This request supersedes Alternative HNP-ISI-ALT-14, Version 1.0, which was submitted by letter dated October 24,2011 (NL-11-2155). The technical and safety basis for the proposed Alternative has not changed from Version 1.0 to 2.0. Version 2.0 provides an improved structural safety margin and addresses technical difficulties encountered while making preparations for the repair. The principal areas of difference in Version 2.0, indicated by revision bars, are: 1) an expanded ultrasonic test (UT) thickness map for the #2 leak location, 2) a larger repair plate for the #2 leak location to ensure base metal of the desired thickness in the weld footprint, and 3) independence of piping structural integrity from river water temperature. Despite these differences, the SNC need date for approval has not changed since SNC desires to recover margin in the degraded piping areas as soon as possible, while achieving the required pre-heat temperature for the repair welds remains the limiting concern, as described below.

U. S. Nuclear Regulatory Commission NL-11-2187 Page 2 As discussed in Enclosure 1, the proposed non-code repair meets most of the requirements for a "full code repair"; however, to perform a repair/replacement activity, IWA-4412 of the 2001 Edition of the ASME Section XI Code with Addenda through 2003, requires that "defect removal be accomplished in accordance with the requirements of IWA-4420." Removing the defects would require that the system be taken out of service, necessitating a plant shutdown. In order to preclude a shutdown, SNC proposes to leave the defects in service and perform a temporary non-code repair requiring NRC approval. SNC requests NRC approval of HNP-ISI-ALT-14, Version 2, by Thursday, October 27, 2011 to support repairs scheduled to begin Friday, October 28, 2011. The SNC need date is based on the plant's ability to start work and on the fact that the required minimum preheat temperature for the repair welding is 60°F, while the current river temperature is 66°F and expected to trend down. If approved, the non-code repair would remain in place until the next refueling outage (scheduled for February 2012) or until the next cold shutdown of sufficient duration to perform the repair/replacement activity, whichever comes first. A similar temporary non-code repair was approved for HNP previously (reference NRC SER dated January 14, 2011 for HNP-ISI-ALT-10). The excavations where the leaks in the PSW piping were observed are located in the Protected Area of the plant adjacent to the Unit 1 Reactor Building, and are surrounded by concrete or steel structures on 75% of the access pathway. The Protected Area is a heavily controlled, low-traffic environment, and metal barriers placed to increase awareness of the excavation site will also prevent smaller vehicles (such as golf carts) from reaching the excavation. In addition, the excavation site is covered by grating material evaluated to meet missile protection criteria for the exposed pipe. The details of the proposed alternative are contained in Enclosure 1 to this letter. Documentation of Engineering Judgment (DOEJ)-HRSNC341 070-S001 and DOEJ-HRSI\\lC341070-S002, provided as Enclosures 2 and 4, address the PSW piping leaks with respect to ASME Section XI Code Cases N-513-2 and N-513-3. provides DOEJ-HRSNC341 070-M001, which addresses the potential for PSW flow diversion due to the observed pipe degradation.

U. S. Nuclear Regulatory Commission NL-11-2187 Page 3 This letter contains no I\\lRC commitments. If you have any questions, please contact B. D. McKinney at (205) 992-5982. Respectfully submitted, M. J. Ajluni Nuclear Licensing Director MJAlDWD

Enclosures:

1. Alternative HNP-ISI-ALT-14, Version 2.0, Temporary Non-Code Repair of Plant Service Water Piping
2. Documentation of Engineering Judgment (DOEJ)

HRSNC341 070-S001, Evaluation of Plant Service Water Pipe Leaks per ASME Code Case N-513-2 and N-513-3

3. Documentation of Engineering Judgment (DOEJ)

HRSNC341 070-M001, Evaluation of Unit 1 Plant Service Water (PSW) Flow with Pipe Degradation

4. Documentation of Engineering Judgment (DOEJ)

HRSNC341070-S002, Evaluation of HNP Unit 1 Plant Service Water Pipe Leaks per ASME Code Case N-513-2 and N-513-3 cc: Southern Nuclear Operating Company Mr. S. E. Kuczynski, Chairman, President & CEO Mr. D. G. Bost, Chief Nuclear Officer Mr. D. R. Madison, Vice President - Hatch Ms. P. M. Marino, Vice President - Engineering RTYPE: CHA02.004 U. S. Nuclear Regulatory Commission Mr. V. M. McCree, Regional Administrator Mr. W.C. Gleaves, NRR Senior Project Manager - Hatch Mr. E. D. Morris, Senior Resident Inspector - Hatch

Edwin I. Hatch Nuclear Plant - Unit 1 HNP-ISI-ALT-14, Version 2.0 Temporary Non-Code Repair of Plant Service Water Piping Alternative HNP-ISI-ALT-14, Version 2.0 Temporary Non-Code Repair of Plant Service Water Piping Hatch Nuclear Plant - Unit 1 Alternative HNP-ISI-AL T-14, Version 2.0 Temporary Non-Code Repair of Plant Service Water Piping UNIT: COMPONENT: SYSTEM: ASME CODE CLASS: FUNCTION: CODE REQUIREMENT: Hatch Unit 1 This unit is in the fourth lSI interval which ends on December 31,2015. 1 O-inch Nominal Pipe Size (NPS) carbon steel piping with a nominal wall thickness of 0.365-inch. Plant Service Water (PSW) The PSW system was built to the requirements of ANSI B31.1, Power Piping Code. The portion of PSW containing this piping is treated as Class 3 for Section XI purposes. This 10-inch diameter piping is the supply header for the Unit-1, Division II, Reactor Building loads listed below: RHR and Core Spray Pump Room Coolers 1T41 B003A1B RHR Pump Seal Coolers 1 E11 B002B/D HPCI Pump Room Coolers 1T41 B005A1B CRD Pump Room Coolers 1T41B001A1B Main Control Room HVAC Condensing Units 1 Z41 B008B/C Two leaks are located on the straight run of buried pipe adjacent to the Unit-1 Reactor Building and were identified by Maintenance personnel during the buried piping inspections. This piping was uncovered initially to address suspected leakage coming from Unit-1 buried piping. To perform a repair/replacement activity, IWA-4412 of the 2001 Edition of ASME Section XI with Addenda through 2003 requires that "defect removal shall be accomplished in accordance with the requirements of IWA-4420." The defects will not be removed during PSW system operation because of the significant increase in the leak rate that would be incurred by removal of the degraded material. Therefore, a modification is proposed which is considered a "temporary non-code repair," necessitating this alternative. See the Proposed Temporary Non-Code Repair section of this alternative for more details. E1-1 Hatch Nuclear Plant - Unit 1 Alternative HNP-ISI-AL T-14, Version 2.0 Temporary Non-Code Repair of Plant Service Water Piping ALTERNATIVE REQUIREMENT: POSITIVE FLAW DETECTION DURING PLANT OPERATION: HARDSHIP OF REPAIR: Southern Nuclear Operating Company (SNC) has applied ASME Code Case N-S13-3, "Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping" for the two observed leaks. However, as a conservative measure, SNC has determined that a temporary non-Code repair will be implemented as soon as possible as described herein. A permanent repair will be implemented during the next refueling outage, which is currently scheduled to begin in February 2012, or during the next cold shutdown of sufficient duration. Compliance with the specified requirements of the Section XI Code would result in hardship without a compensating increase in the level of quality and safety; therefore, approval of this alternative per 10 CFR SO.SSa(a)(3)(ii) should be granted. On October 21,2011, two through-wall leaks were discovered in the PSW system. The initial leakage is documented in Hatch Condition Report 364491. HNP-1 Technical Specifications (TS) 3.7.2 requires that two PSW subsystems and one UHS (Ultimate Heat Sink) be operable. Performing an ASME Code repair at this location during power operation would require that Division II of PSW be taken out of service. With a division of PSW out-of-service, TS 3.7.2 Condition E requires that the PSW subsystem be restored to Operable status within 72 hours. While the Technical Specification provides 72 hours for repair, doing so would result in the loss of one train of emergency cooling components during the repair window. In addition, isolation and draining of a PSW loop during power operation is complex and would expend a significant portion of the 72 hours allowed. Shutting the plant down to perform a Code repair vs. using the proposed temporary non-code repair is considered by SNC to be a hardship. E1-2 Hatch Nuclear Plant - Unit 1 Alternative HNP-ISI-AL T-14, Version 2.0 Temporary Non-Code Repair of Plant Service Water Piping DEGRADATION MECHANISM: FLAW SIZING: EVALUATION APPROACH AND RESULTS: The exact cause of the degradation has not been confirmed, as it is internal to the pipe. However, based on the degradation pattern, the cause is expected to be localized corrosion. Additional areas of this piping were examined and found to have no degradation. This data, along with the required broadness examinations of ASME Code Case N-S13-3, provides assurance as to a lack of potential additional broadness issues. Detailed ultrasonic testing (UT) measurements were obtained around the area of the two leaks to better understand the scope of the degradation (See Figures 1 and 2 for Locations #1 and #2, respectively). The UT measurements were compared to the published minimum wall thickness of 0.078-inches (includes corrosion allowance) taken from Bechtel calculation number 020 (VOL 999 BIN 999) for Hatch Unit 1 PSW. At Location #1, the pipe wall thickness was found to be less than 0.200-inch in a circular shape that is 1-1/8-inches in diameter. At Location #2, the pipe wall thickness was found to be less than 0.200-inch in an elliptical shape that is 2-1/4-inches on the major axis and 2-inches on the minor axis. However, an acceptable reading could not be obtained on the piping above the flaw at Location #2. The rest of the piping in the examination grid was found to have a wall thickness greater than 0.200-inches. Subsequently, UT thickness measurements were performed that showed additional localized thinning above minimum wall (see Proposed Temporary Non-Code Repair below). For details, see Documentation of Engineering Judgment (DOEJ)-HRSNC341 070-S001 and -S002 as provided in Enclosures 2 and 4. Because PSW is functioning in an operable but degraded condition, the following issues as identified below were addressed to ensure that no harm to plant safety or public health exists. Once the proposed temporary non-code repairs are made, any potential adverse effects due to leakage would be mitigated. Flaw Evaluation: A flaw evaluation was conducted in accordance with Section 3.0 of Code Case N-S13-3 to evaluate the leak. The Code Case N-S13-3 flaw evaluation determined that structural integrity is being maintained. Stress Analysis: The added weight of the two plates to be welded to the affected piping (See Proposed Temporary Non-Code Repair below) was reviewed, and did not impact the stress analysis calculations. E1-3 Hatch Nuclear Plant - Unit 1 Alternative HNP-ISI-AL T-14, Version 2.0 Temporary Non-Code Repair of Plant Service Water Piping Flow Diversion: An analysis was performed to estimate the leakage from the piping based on the area of the flaw size that was below the thinnest measurable wall thickness, or 0.200 inches. Although the current leakage area is smaller than the flawed area that is below 0.200 inches, the analysis conservatively assumed that the area of the leakage would be equal to the flawed area below 0.200 inches. Based on the ultrasonic thickness readings, Location #1 was assumed to be 1.125 inches in diameter and Location #2 was assumed to be elliptical with a 2.25 inch major axis and a 2 inch minor axis. Conservatively this was modeled by assuming two 2" x 3" holes which is modeled as a 2.45" diameter hole. The model was run for this case and additionally for the loss of inventory from a 3.97" diameter hole in Division II of the PSW system. The results were then evaluated against the design flows to safety-related components during a LOCA using the PROTO_FLO model (2007 benchmark update). The results of this evaluation showed that with a 3.97-inch diameter hole in the 1O-inch line, that all safety-related components would receive adequate PSW flow during a LOCA. The details are described in DOEJ-HRSNC341070-M001, which is provided in Enclosure 3. Therefore, with the worst case leak due to loss of material from the existing location, the PSW system would still be capable of providing the required cooling to all components. Water Temperature: In Version 1.0 of this alternative, DOEJ-HRSCN340170 S001 (Enclosure 2) indicated that the pipe temperature must remain above 46°F. Because of the limitations associated with this evaluation, a more detailed approach which is not temperature dependent was used as allowed in Code Case 1\\J-513-3. This subsequent evaluation, DOEJ-HRSCI\\J340170-S002 (Enclosure 4) noted that the piping will remain structurally sound without temperature limitations and will meet the B31.1 Code requirements. Spraying: The leak locations were considered for impact on other components. There is no equipment in this area that could be affected by these leaks. This information provides a reasonable expectation that this condition would not affect ability of the PSW systems, or other components located in the area to perform as designed. Flooding: With respect to the potential for flooding due to excessive leakage into this area, there is only piping and no equipment in the excavated pit. This provides reasonable assurance that the components in this area would be capable of performing the necessary design functions in the event of flooding. Therefore, the amount of leakage into the area will not affect the operability determination of the PSW system. E1-4 Hatch Nuclear Plant - Unit 1 Alternative HNP-ISI-ALT-14, Version 2:0 Temporary Non-Code Repair of Plant Service Water Piping AUGMENTED EXAMINATIONS: Flaw Growth Rate: As stated previously, the cause of the degradation is believed to be from localized corrosion. If further degradation were to occur on this area of the piping, it would be minimal and gradual with respect to the time frame for the next opportunity for piping repair (next refueling outage or until the next cold shutdown of sufficient time to perform the repair/replacement). This assumption is further justified by the fact that the piping with the degradation is original plant piping, and has been in service for approximately 36 years. There is reasonable assurance that the calculations and evaluations associated with the current degradation would remain valid until a Code- ' compliant repair/replacement is performed. The daily rounds and the ongoing ultrasonic examinations performed on a 30-day frequency will enable Hatch to verify that structural integrity is maintained. Based on the above discussion, SNC has determined that the structural integrity of the PSW piping at this location is being maintained and will continue to be maintained until a Code-compliant repair/replacement is performed. To determine the extent of condition, five sample points, as specified by Code Case N-S13-3, will be examined using ultrasonic thickness techniques. If any of these examinations identify piping with thickness measurements below the required minimum wall thickness, the condition will be documented in a condition report and this operability determination will be re-evaluated. This will meet the guidance of Code Case N-S13-3. The five sample points will be at the following locations: Point 1 - Scan 2 feet of piping downstream of valve 1 P41 F380A

  • . Point 2 - Scan 2 feet of piping downstream of valve 1 P41 F380B Point 3 - Scan the 8 feet area in excavation #1 as previously directed by the Buried Pipe Program Point 4 - Scan 2 feet of piping between valve 1 P41 F066 and the wall penetration Point S - Scan 2 feet of piping between valve 1 P41 F067 and the wall penetration.

The UT thickness examinations for the five sample points identified above are expected to be completed prior to November 20,2011. E1-S Hatch Nuclear Plant - Unit 1 Alternative HNP-ISI-AL T-14, Version 2.0 Temporary Non-Code Repair of Plant Service Water Piping PROPOSED TEMPORARY NON-CODE REPAIR: Several repair/replacement activities were evaluated and it is proposed that the addition of two contoured plates to the affected sections of piping by means of welding be made to isolate the leaks (see Figure 4). This option allows the welding on the two attachments to be located in an area with minimal degradation, ensuring a structurally sound load path while minimizing the risk of "burn-through" and increased leakage. The design will also ensure that the configuration of the repair will allow continued wall thickness monitoring of the region by ultrasonic examination to ensure that future degradation will not adversely impact the structural capability of the repaired section. The degraded piping is 10-inch, Schedule 40 (0.365-inch nominal wall), seamless carbon steel piping. The repair plates will be constructed from either plate or pipe; the Hatch site plans on using plate. In either case, the thickness of the repair plates will be 0.365-inch nominal wall from P-I\\lo.1 carbon steel material having an allowable stress of 15,000 psi up to 650°F. If it is determined that plate will not work, piping will be used. Plate #1 covering Location #1 This location is essentially at the 12 o'clock position. The size of this plate was based on inputs from the ultrasonic thickness measurements taken as requested by the SNC Corporate Stress Group. The UT examiner was asked to find where the wall thickness measured at least 0.200-inches and at least 0.300-inches away from each leaking location in four directions. The examiner was able to get the requested eight ultrasonic measurements. As noted herein, copies of the test report and thickness measurement results are enclosed as Figures 1 A and 1 B for the initial UT examinations of Locations #1 and #2. Based on these measurements (ref. Figure 2), a 3-inch by 3-inch plate will be positioned over Location #1 as shown in Figure 4. Plate #2 covering Location #2 This location is at the 7 o'clock position looking south. The size of this plate was based on inputs from the UT measurements taken as requested by the SNC Corporate Stress Group (ref. Figure 3). The UT examiner was asked to find where the wall thickness measured at least O.200-inches and at least O.300-inches away from each leaking location in four directions. The examiner was able to get seven UT readings; however, a measurement could not be obtained for the ultrasonic point for the 0.200-inch location at the "upper" side for Location #2 because of ID surface irregularities. After finalization of Version E1-6 Hatch Nuclear Plant - Unit 1. Alternative HNP-ISI-ALT-14, Version 2.0 Temporary Non-Code Repair of Plant Service Water Piping 1.0 of this alternative, Hatch site personnel were verifying the area for placement of the plate to cover Location #2. During preparatory activities, additional UT examinations were performed and additional pipe wall thinning in the weld area around the footprint of the plate covering Location #2 was discovered. Refer to Figures 1 C and 10 for the UT test report and thickness measurements, respectively, for the additional UT thickness examinations performed for Location #2. As a result of the additional UT examinations and the results thereof, it was determined that the plate for Location #2 had to be*re-designed. The size of the plate was changed to 3-inch by 13-3/8-inches as shown in Figure 4. Additional details from Figure 4 are provided in the enclosed Sketch 1, which includes that area between the edges of Plate #2 and the adjacent areas of piping in which the wall thickness is at least 0.300-inches. An additional 1/2 inch of base material beyond the area covered by Plate #2 was scanned t6 verify sound base material in the attachment weld footprint. As noted above, 10 surface irregularities limited the UT thickness measurements. Although this cannot be confirmed, SNC is of the opinion that a better representation of the thickness at Location #2 is depicted in the enclosed Sketch 2. All welders and welding procedure specifications shall be qualified for groove welding in accordance with the ASME Section XI Code. The new pressure boundary will now be located at the reinforcing plate attaching weld. The welding process to be used for attaching the reinforcing plate will be the shielded metal arc welding (SMAW) process. If rejectable indications are identified during performance of nondestructive examination, the indications will be removed and the attachment weld repaired in accordance with applicable provisions of ASME Section XI and ANSI 831.1. The welding is to be performed with water in the line and with the system pressurized to approximately 120 psig. SNC believes that this will not create any problems based on the following factors:

  • Welding with water in a pipe is performed frequently in the industry and, as discussed above, the water temperature meets the 60°F minimum preheat.
  • The measurements noted in Figures 2 and 3 indicate that the welding will be performed on thicknesses ranging from 0.200-inch to 0.300-inch thick.
  • With the water in the system acting as a heat sink, the resulting heat affected zone of the piping base material caused by the welding should be relatively shallow.

E1-7 Hatch Nuclear Plant - Unit 1 Alternative HNP-ISI-ALT-14, Version 2.0 Temporary Non-Code Repair of Plant Service Water Piping Since only the inner 0.200-inch of the base material is required for pressure containment, welding on 0.200-inch thick to 0.300-inch thick base material would not be expected to encroach upon the Code required minimum wall thickness and should have no impact on the load bearing capability of the piping during the welding process. The schedule for the application of the two plates is limited by river water. ternperature. As noted above, this water is used as a heat sink for the welding process. The present water temperature is 66°F and is expected to trend downward. In order to enhance the welding process, SNC wishes to perform the temporary non-Code repair of the affected piping before significantly lower temperatures occur. The completed welds will be VT examined per ANSI 831.1 and any indications evaluated per the requirements of ANSI 831.1. A pressure test will then be performed as required by IWA-4540 of the Section XI Code. The pressure test with be accompanied by a visual VT-2 examination. Additionally, a liquid penetrant examination will be performed in accordance SNC procedure NMP-ES-024-301. The examination will be performed no less than 48 hours after completion of the weld to ensure no delayed cracking occurs. (This examination is consistent with the requirements for weld overlay repair examinations made on P-No. 1 material using ASME Code Case N-661 1, which has been accepted for use in Regulatory Guide 1.147). I\\IMP-ES-024 301 provides techniques and acceptance criteria to be used for the performance of Liquid Penetrant Examinations at the Hatch, Farley, and Vogtle nuclear plants. Indications will be evaluated per the following procedural acceptance criteria:

1. Relevant indications are indications which result from imperfections. Only indications with major dimensions greater than 1 /16-inch shall be considered relevant imperfections.
2. Imperfections producing the following indications are unacceptable:

Any cracks or linear indications. Rounded indications with dimensions greater than 3/16-inch. Four or more rounded indications in a line separated by 1 /16-inch or less edge to-edge. Ten or more rounded indications in any six square inch area with the major dimension of this area not to exceed six inches with the dimension taken in the most unfavorable location relative to the indications being evaluated. E1-8 Hatch Nuclear Plant - Unit 1 Alternative HNP-ISI-AL T-14, Version 2.0 Temporary Non-Code Repair of Plant Service Water Piping CODE CASE N-S13-3 ACTION PLAN: STATUS: ALTERNATIVE DURATION

3. An operating system VT -2 pressure test will then be performed as required by IWA-4540 of the Section XI Code.

The following actions will be performed by SNC for this component until the proposed temporary non-code repair is performed: Site personnel will perform daily rounds to identify further degradation of the affected area as evidenced by a significant increase in the leakage rate. If a significant increase in leakage is detected an ultrasonic examination will be performed to assure that the criteria used to evaluate the structural integrity remains valid.

  • The area will be ultrasonically examined on a 30 day frequency to assure that unexpected degradation is not occurring and that the structural integrity of the piping is being maintained.

The following actions will be subsequently performed by SNC in the time period after the temporary non-code repair is made until the ASME Section XI repair/replacement is performed:

  • An ASME Section XI repair/replacement will be performed before the completion of the Hatch Unit 1 1 R25 refueling outage currently scheduled to begin in February 2012 or during the next cold shutdown judged to be of sufficient time to perform the repair/replacement, whichever occurs first.
  • Site personnel will perform daily rounds to identify any signs that additional degradation is occurring.
  • The area around the temporary repair will be ultrasonically examined on a 30-day frequency to assure that degradation outside of the repaired area is not occurring and that the structural integrity of the piping is being maintained.

This alternative is awaiting NRC approval. This alternative will remain in effect until an ASME Section XI Code repair/replacement is performed during the refueling outage scheduled to beginning in February 2012 or during the next cold shutdown of sufficient duration to perform the repair/replacement, whichever occurs first. E1-9 Hatch Nuclear Plant - Unit 1 Alternative HNP-ISI-ALT-14, Version 2.0 Temporary Non-Code Repair of Plant Service Water Piping Nuclear . $OUlH&IIH.\\ Ma~ent COMPANY Ins.truction UltrasonIc Thlcim&&s Report l'4NW"'~_L...Io' Plant I Vnt/. l7<r-k. t] I Southern Nuclear Operating Company NMP-E&-024*5t 1 Utlrasonic Thickne.ss Examination Component: I~SO I Drawing No,; . o "\\?:::'<L

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.::> I MHz t3: A-Scan o Motered I DIgital 0 Spot ,El Continuous /,/ Couplant: <;C"..... ~,..c:'." d.... / t*t }.~.,.,,),,~ err:- ir', ') f2,. <: Cable Type 1 Length: " r " EXAMINATION RESULTS J <ae '.'\\... ' (""l- ....~ . j~ Examiner. ILevel: _u)4 ,l,I )4 LeV$1 1Dale: 0 I'Z'i) I I ~ Figure 1A -10/21/11 Ultrasonic Thickness Test Report - Page 1 of 2 E1-10 Hatch Nuclear Plant - Unit 1 Alternative HNP-ISI-ALT-14, Version 2.0 Temporary Non-Code Repair of Plant Service Water Piping tialen 1 PSW BurJ!!d Plpe-l0!21/11 Hatr::h 1-*1.0" Plant Service WatN UT around areas of leak:ag~s located in excavation number 2, wh<ere 9 ff'et of piPf! was exr.osed. Reference S-{)()631/ H 11139, Leakage from area nomber 1-is loc~ted top dead c-enter of the pipe at... 2 feet from the north erod of the dirt wall. The area was griddP.<d I 4a for UT FA!: inspe<;{ion. l eakage from area nUIY\\ber 2 - is Iocatf'd ~ 2-1/2 ff'et from the north dirt wal'l itt 7:00 when facing north. The area was grf-dded F40 for IJT FA!:. iftSpectl'oli. East Side 1.25" (.3) ,625 (,2) North End 1.37S{.31.5(0).. .625(.2) 1,37S{.3) South End .50(.21 1,75(,3) West Sid1! Area 1 Up 1.5(.3) None(.2) North EOO 3(.3) 1.0(.2) + 1.25(.2) 2.51.3) South EM sPI 1.250(.3) Down + !! c;entN oHM lei!"k Inches from the le.ak (thic~ne'5~ at that locationI North and South is tne longltudhHl1 u *.is of th~ pipe. Most other 3r<'as were difficult to obtain UT readings of.100 or Ie-55. .The e:a:amination was performed usln& a USN 60 with a 5.0 MHl3/8~ duallransduCf!f 'IIh@re a one Inch screen ranse W3S estdbll~ed, Figure 1 B-1 0/21/11 Ultrasonic Thickness Test Report - Page 2 of 2 E1-11 Hatch Nuclear Plant - Unit 1 Alternative HNP-ISI-ALT-14, Version 2.0 Temporary Non-Code Repair of Plant Service Water Piping Southem Nuclear Operating Com~a~ Nuclear NMp*ES*024*511 Ultrasonic Thickness Examination SOUlNON~ Management Version 3.0 COMPANY Procedure Instruction Page 15 of 17 Ultrasonic Thickness Report Southern.Nuclear Operating Company -,P:;:-ln'T,. bYA L~t::2L~u~n r4Iti-='~1..'=-_--r"""'"'::--:-::;:-----;-----:c:--_I-D_a_te...ll1o:>::.cL..-~:2UiI'OOl:.-:.J\\~1--1-1~~eet Number: Component: lisa I Drawing No.: {r/'.Di~ it!:::".~ ~ ( o'3.\\ System ~. r:::-lW Examination Area and Location: ~.....- A +-l J>.~j "l"",,,.>~. A..J:....,sj } Description of Item Examined: P / / Calibration Standard Serial Number: Instrument Manufacturer: Model Number: Serial Number: ~hl t:\\ (&0 ! \\ <~ \\~ Sound Path Screen Distance.; Iiransducer Type: Size: Manufacturer: \\," ~~"''"<.~';.c. > ~\\ .~\\~( Smallest Screen Division: Serial Number Frequency: Sl\\.\\~ Z::. '1'31-{&/f .02." MHz Procedure: tVM? - r.:">.()~"\\ -5\\\\ .e A-Scan o Metered I Digital Remarks: tV.+ o Spot eg Continuous Couplant: c:..,..... ~.k d.... l~;). ),$'""1$ Cable Type I Length: S Level: Level:.I I /I'/'ij I~ I f Reviewed 6'y:~ ff Le~ ~ ~~~~ Figure 1 C - 10/26/11 Ultrasonic Thickness Test Report - Page 1 of 2 E1-12 Hatch Nuclear Plant - Unit 1 Alternative HNP-ISI-ALT-14, Version 2.0 Temporary Non-Code Repair of Plant Service Water Piping 0 C, '"<I __,0 flO z~ ~...J _Q. 2-d W -j lL <{o )I" Vi o~ z ~~

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1) x-Location of the Indication
2) The numbers represent the distance of the ultrasonic thickness measurement from the leak and the numbers in parentheses are the thicknesses of the base material. All measurements are in inches.

Figure 2 - Leak Location 1 E1-14 Hatch Nuclear Plant - Unit 1 Alternative HNP-ISI-AL T-14, Version 2.0 Temporary Non-Code Repair of Plant Service Water Piping Location 2 Up 1.5(.3) None(.2) North 3.0(.3) 1.0(.2) X 1.25(.2) 2.5(.3) South 0.5(.2) 1.25(.3) Down Note:

1) x-Location of the Indication
2) The numbers represent the distance of the ultrasonic thickness measurement from the leak and the numbers in parentheses are the thicknesses of the base material. All measurements are in inches.

Figure 3 - Leak Location 2 E1-15 Edwin I. Hatch - Unit 1 Alternative HNP-ISI-ALT-14, Version 2.0 Temporary Non-Code Repair of Plant Service Water Piping I--"" F--I -~' [. ~-* .J~ i-----------,--'c------- \\--'-- --1 1----:-------,-*:t ----------1 NQltTU Nott.$.l I '.... *0:.f\\Q ...' -c.~ I"~ c;onPlfl r ~ ot \\P

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Edwin I. Hatch Nuclear Plant - Unit 1 HNP-ISI-AL T-14, Version 2.0 Temporary Non-Code Repair of Plant Service Water Piping Documentation of Engineering Judgment (DOEJ)-HRSNC341 070-S001, Evaluation of Plant Service Water Pipe Leaks per ASME Code Case N-S13-2 and N-S13-3

Southern Nuclear Operating Company DOCUMENTATION OF ENGINEERING JUDGMENT DOEJ-HRSNC341070-S001 Evaluation of Plant Service Water Pipe Leaks per ASME Code Case N-S13-2 andN-513-3 Version Record Version Originator/Date Reviewer/Date No. Signature Signature 1 An Nguyen I October 22, 2011 Y. Jani / October 22,2011 ....d. 2 An Nguyen / October 24, 2011~;J. Y. Jani / October 24,2011 ~ Oocumenlalion of Engineering Judgment E2 - 1

DOEJ-H RSNC341 070-S001 Southern Nuclear Operating Company

Purpose:

The purpose of this DOE] is to support RER SNC341 070. The scope for this DOE] is the evaluation of the piping structural integrity for the plant service water system. This piping has developed through wall seepage and UT inspection has been performed. Design Inputs (Reference NMP-ES-042):

1. S00631.
2. S00779.
3. Attachment to E-mail from Kevin White to An Nguyen, 10/21/JI (Attachment I shows the sketch redrawn in Temporary Non-Code Repair plan)
4. Exposed Piping Evaluation (Attachment 3)

References:

I. Code Case N-513-2 and N-513-3.

2. ASME Section Xl, 2003 (Code of Record).
3. ASME Section XI, 2010.
4. RASEARCH Results for N-513-2 and N-5J3-3 (Attachment 2)

Assumptions: In this evaluation, a representative flaw geometry enveloping the geometry of the two flaws was used. From the UT report (attachment I), the flaw can be characterized as 2.5 inch in the circumferential direction and 2.25 inch in the longitudinal direction. The minimum pipe wall thickness outside of the flaw is at least 0.2 inch. Due to the size of the flaw and the nominal thickness of the pipe wall, the evaluation is limited to temperature higher than the upper shelf temperature ofcarbon steel. In this case, for the thickness of0.365 inch, the upper shelf temperature is determined from Table C-8321-2 of Reference 3 as 45.6°F. The design pressure and design temperature are 180 psig and 125°F, respectively. The piping system was classified as buried pipe. As such, this piping system does not have stress calculation. Now the pipe is exposed in the pit. The exposed piping was evaluated as shown in attachment 3 of this DOE]. Since the temperature of piping system is low (125°F), no secondary stress evaluation is required. Evaluation: This evaluation is in accordance with Code Case N-513-2 and 513-3. The difference between the two versions of the code case is not applicable to this case as discussed in reference 4. The exposed pipe span is approximately 12 ft span. The natural frequency is calculated to be -40 hz; hence, there is no concern for seismic. A conservative value of 1500psi for bending stress was used for primary longitudinal stress in the code case calculation. Frequency calculation and primary stress due to weight and seismic are shown in attachment 3. NMP-ES-039-002. Version 2.0 Documentation of Engineering Judgment, E2 - 2

DOEJ-HRSNC341070-S001 Southern Nuclear Operating Company Circumferential Flaw Calculation: ThiS spreadsheet provides an evaluation of pipe wall flaws, including through wall, per Section XI, Appendix C and is primarily focussed on Code Case N-513-2, for Class 2 or 3 piping only (service level 8 controls) Color indicates cells requiring inputs Color indicates output or result information Constant pi= 3.14159 Nominal Condition NPS= '. WOQO :.' NPS OD=JO:Z50.*. Pipe OD, inch tnom=., 0_365 '. Pipe nominal thickness (in) Snom= 29.904 Pipe Section modulus, in"3 StressJ>ri=

.:l_~OCl ::

primary nominal stress Sb (ksi) moment_pri= 44.856 primary moment, (in-kip) Pressure= ~ :j:Pf~ '\\; pressure, ksi Stress_sec'" :> 0.000,..,: Secondary Stress, (ksi) momenCsec= 0.000 secondary moment (in-kip) Safety Factors per C-2621 SFb= ' : 2.300 ' safety factor for bending stress: 2.3 (A), 2 (8), 1.6 (C)/ 1.4 (D) SFm= 2.700.' safety factor for membrane stress: 2.7 (A), 2.4 (6)/ 1.8 (C), 1.3 (D) AS Found Condition tpipe= 0.200. corrode pipe thickness, in. Rm= 5.28 mean pipe radius, in. Scorr= 17.16 corrode pipe section modulus, in" 3 Sigma_b= 2.61 adjusted bending stress, kSi slgma_e= 0.00 adjusted bending stress, ksi Rm / tpipe= 26.38 Compare R/t to 20 per Code Case N-S13-2, Section 1-2. Flaw Sizing L_circ= . 2.50 circ flaw length, in theta = 0.24 half flaw angle, rad theta/pi = 0.08 C_= 1.25 flaw half length C_ to be used in C-7000, in. Jjc= 300.00 J_ic (in-lbf/in"2) - OP TEMP> Upper Shell Temperature E= 2,94E+O E, ksi 2 NMp*ES-039-002. Version 2.0 Documentation of Engineenng Judgment. E2 - 3

DOEJ-HRSNC341070-S001 Southern Nuclear Operating Company E-prime= sigma_y= sigma_u= calculate SrJlrime sigma_m= beta= sigma_bJlrime= theta+ beta = phi= sigma_m_prime= (sigma_b+sigma_e)/sigma_m= S_RJlrime= calculate KjJlrime per C 4311 Appendix I of N-513-2 Am= Bm= Cm= Ab= Bb= Cb= Fm= Fb= K_ib_C4000= K_i_C4000= KJJlrime= SC= SCreening PYoceoure: J NMP*ES *039*002. Version 2.0 4 3.23E+0 4 27.10 60.00 2.42 1.36 29.72 1.60 0.12 23.03 1.08 0.09 18.95 -48.20 72.36 15.84 -37.56 44.91 1.33 1.27 6.35 3.24 9.59 0.10 1.11 Use C6000 EJlrime= E / (1-nu"2) yield strength, ksi ultimate strength, ksi sigma_m = pD/4t (ksi) beta angle (rad) in figure 1 of code case limit load primary bending stress Check for short crack: theta +beta < pi phi;: =ASIN(0.5*SIN(tI1eta>>, rad. sigma_m_prime= sigma:,.y*(l-theta/pi-2*phi/pi), ksi check for (sigma_b+sigma_e) / sigma_m > 1. S_RJlrime=(Sigma_b+slgma_e)/sigma_bJlrime Am = -2.02917+1.67763*(Rm/tpipe) 0.07987*(Rm/tpipe)" 2 +0.00 176*( Rm/tpipe)" 3 Bm = 7.09987-4.42394*Rm/tpipe+0.21036*(Rm/tpipe),,2 0.00463*(Rm/tpipe)" 3 Cm = 7.79661+5.16676*(Rm/tpipe) 0.24577*(Rm/tpipe)" 2+0.00541 *(Rm/tpipe)" 3 Ab =-3.26543+ 1. 52784*(Rm_oveUpipe) 0.072698*(Rm_oveUpipe)" 2+0.00 16011*(Rm_oveUplpe)"3 Bb = 11.36322-3.91412*(Rm/tpipe)+0.18619*(Rm/tpipe)" 2 0.004099* (Rm/tpipe)" 3 Cb =-3.18609+3.84763*(Rm/tpipe) 0.18304*(Rm/tpipe)" 2+0.00403 *( Rm/tpipe)" 3 calculate per Code Case Appendix I Calculate per Code case Appendix I KJm=sigma_m*(pi*C_)"0.5*Fm K_ib=( momeflCpri+momencsec)/(2*pi*Rm" 2*tpipe)*(pi*C_),,0. 5' Fb KJ=K_im+K_ib K_r_prime=( 1000* K_i" 2/(E_prime" ]_ic>>"0.5 Screening Criteria SC:: K_r_prime/S]Jlrime 5'C<O.2Use (5050 0.2<SC<1.8 Use C6000 Documenlation of Englnee(ing Judgmenl. E2 - 4

DOE]-HRSNC341 070-500 I Southern Nuc1ear Operating Company ' For SC< 0.2 Use C5000 (to be provided) For 0.2< SC < 1.8, Use C6000 Calulation sigma_c_b per C-5320 sigma_f= beta = slgma_c_b= Calculation sigma_c_m per C-5222 sigma_c_m= S_t per C-5322= Calculation per C-6320 Z= S_c= S_t= Calculate per C-2612 sigma_b_over_S_c= For SC > 1.8 Use C7000 Fm_ClOOO= Fb_C7000= K_im= K_ib= K_ir= K_I_ClOOO= K_C= U_C7000/K_C= 43.55 1.37 47.77 19.25 37_00 13.70 1.38 13.30 9.95 0.20 OK 0.24 OK 1.33 1.27 17.16 15.18 0.00 32.34 9845 0.33 OK SC> l.B Use 0000 Flow stress = (sigma-y+sigma_u)/2 beta angle from figure 1 bending stress at collapse allowable bending stress per C-5320 membrane stress at collapse= sigma_f*( 1-(theta/pi)-2*phi/pi) allowable membrane stress per C-5320 :: sigma_c_m/SFm load multiplier for ductile flaw S_c= 1/SFb*(sigma_c_b/Z-slgma_e)-sigma_m*( 1-1/(Z"'SFm>> S_t '" sigma_c_m/Z/SFm Fm = 1+Am*(theta/pl)" 1.5+Bm*(theta/pi)"2.5+Cm*(theta/pi)"3.5 Fb = 1+Ab*(theta/pi)" 1.5+ Bb*(theta/pi)" 2.5+Cb*(theta/pi)" 3. 5 K_im = SFm*Fm_C7000*sigma_m*(pl*C_Y'0.5 K_ib = (SFb*Sigma_b+sigma_e)*Fb_ClOOO*(pl*C_)"0.5 Residual stress intensity K_I_ClOOO :: K_im+K_ib+K_ir K_C :: (J_ic-E..prime/1000)"O.5 K_I_ClOOO < KJ :==> OK NMP*ES*039*002, Version 20 Documenrarion of Engineering Judgment. E2 - 5

DOEJ-HRSN C341070-S00 I Southern Nuclear Operating Company Ax ial Flaw Calculation: Constant pi= Nominal Condition NPS= 00= tnom= Snom; Pressure= S Cmin As Found Condition tpipe= Rm= sigma_h= Rm / tpipe= Flaw Sizing L_axial= . c_axial= lambda= F_= sigma_y= sigma_u= Sigma_f= SFmaxial; Sigma_L= Calculation per C-4312 Q..'" K_i_axial= Kj-prime_axial = S_r.J)rime_axial= SC_axial= Screening Procedure: 3.14159 10.000 10.750 0.365 29.904 0.180 0.064 0.200 5.275 4.748 26.375

~~

NPS Pipe OD, inch Pipe nominal thickness (in) Pipe Section modulus, in"3 pressure, ksi Allowable design stress per Section II, Part D, Table 1A per Code case, Eq'n 4 corroded/degraded pipe thickness, in. mean pipe radius, in. hoop membrane stress, ksi Compare R/t to 20 per Code Case N-513-2, Section r-2. axial flaw length, in 1.125~ClIf<:rilckl~llgth.. ifl. 1.095 1.603 27.100 60.000 27.100 1.000 14.306 0.145 0.175 0.829 Use (6000 yield strength, ksi ultimate strength, ksi safety factor for membrane stress: 2.7 (A), 2.4 (8), 1.8 (C), 1.3 (D) Sigma L is defined as yield strength in this case =(Pressure* Rm/tpipe)*(pi*c_axial)"O. 5*F _ =( 1000*KJ_axial"2/(E.J)rime*J_ic>>" 0.5 =Pressure"Rm/tpipe/Sigma_L Screening Criteria SC=K_r _prime/S_R.J)rime SC<0.2 Use C5000 0.2<SC< 1.8 Use (7000 in lieu C6000, SInce under preparation per C-6420 SC> 1.8 Use (7000 Cille>jl3tiOf"l per (-5400 (Not a':aiiaole for through \\'::111) L_all= 5.269 Code case equation 1 5 NMP*ES-039*002. Version 2.0 Documentalion of Engineering JUdgment. E2 - 6

DOEJ*HRSNC341070*S001 Southern Nuclear Operating Company Calculate per (-7400 for flaw length L_axial K_Im_axial= 38.625 K_Im_axial = K_i_axial-SFmaxial K_c= 98.450 K_Im_axial < K_c OK 0.392

== Conclusion:== Current flaw configuration meets the criteria for temporary acceptance of flaws in moderate energy class 3 piping system. This evaluation is in accordance with Code Case N*513-2 and N* 513-3. Hence, the following compensatory actions are also required: Compensatory Measures

  • Daily monitoring of leakage for noticeable changes
  • UT - at least monthly based on no noticeable leakage change
  • PSW supply temperature monitoring (river). Minimal acceptable temperature is 46deg.

Projection of temperature for 30 days should ensure minimum temperature is not challenged. List of Attachments: I. UT Results.

  • 2: RASEARCHResuIts **
3. Exposed Pipe Evaluation.

6 NMP*ES-039-002. Version 2.0 Documentarion or Engineering Judgment. E2*7

Location 1 East 1.25(.3) North 1.375(.3) 0.50(.2) 0.626(.2) X 0.625(.2) 1.375(.3) South 0.50(.2) 1.75(.3) West Note:

1) X-Location of the Indication
2) The numbers represent the distance of the ultrasonic thickness measurement from the leak and the numbers In parentheses are the thicknesses of the base material. All measurements are in inches.

Location 2 Up 1.5(.3) None(.2) North 3.0(.3) 1.0(.2) X 1.25(.2) 2.5(.3) South 0.5(.2) 1.25(.3) Down Note:

1) X-Location of the Indication
2) The numbers represent the distance of the ultrasonic thickness measurement from the leak and the numbers in parentheses are the thicknesses of the base material. All measurements are in inches.

DOEJ-HRSNC341070-S001 - Attachment 1-1/1 E2 - 8

DOEJ-HRSNC341070-S001 Altachment 2 1/3 Nguyen, An N. From: Nguyen, An N. Sent: Friday, October 21.2011 7:46 PM To: Retherford, Rebecca Sue Cc: Edwards, James A. (Jim - SNC): Agold. James M.

Subject:

RE: Code Case N-513-3 Thank you. Below is the commentary on N-S13-3. An Nguyen, PE Telephone' 8-992-7307 Gatherl'd from Rasl:arch NUC Files\\Revision-NuclearCases.wpd (7/ 16/2010)

'=... :=.~;..~;:.: =~~::.:.:.====~:..::..:::..::..:== :.:.:_.::.:=. =:.:;.=====::.:=====:.::;:=====.z~.:.-=.:=.:.:=-==:::: =.:======---=::::-:..:= ===::=:=.:.====.;:.::;=-~=.=.;::.=.====

unuu#########################ff###########~########### Codc Case Revisions N-S IJ-J (07-S8) (07-1 J()J) Evaluation Criteria f()r Temporary Acceptance of Flaws in Moderate Energy Class ~ or 3 Piping. Section XI, Divisiori I TECHNICAL Thi s n:visioll provitks sigllificant clarifit:Lltions rt!g:arding evaluation of through-wall, Ilonplunar tlU\\VS, which arc the tlaw type Illost communly disposirioned using this eLIse. The acceptabi lity criterion fix the prior brancil rcinfom:ll1ent evaluation approach (based on ASME Section III Class 2 and J rules) was ambiguous. This revisioll spct:ifies that the reqllired area of reinforcement is to be caklllatl:d in aCL:ordance with Class I rules, and proves IlC\\V (acceptance criteria for this approach. Also, the depth at which a through-wall, nonplallilr flaw i~ characteriz..::d for planar evaluation in both thc axial and ein:ul11f..::rential directions is made less restrictive ill the propu~eJ revi sion. (0 accolillt for NDE capabilitil.!s. A new I:quatioll is introdllced to addrl:ss the potential for pressure blowout iran arca larger (han thc curren! through-wall, nonplanar leak is evaluated to provide a bOllnding analysis. N-5lJ-~ (04-S I) (0('01-249) (i\\cccptablt:; - Regulatory (Juide 1.147 - Rev, 15) 1-: V;lILlatillll Critcria fllr Temporary Acccptance of Flaws in Moderate Energy Cla~s 2 or 3 Piping. Sediun XI. Division I TEC1INfCAL This rev ision ;Idds a procedure for cvaluation of non-plallar throllgh-wallllaws in moderate Cllcrgy piping. Scrvice cxpcricncc has shown thaI SOIllC piping "utTcr degradation fr\\)ll1lloll-planar tlaws.,)lIch as pitting and E2 - 9

DOEJ-HR5NC341070-5001 AUachmenl 2 213 Illiuobiological attack. where local illcollsequenri;illcakage can OCCllr. SOllle Owners have llS\\;*tI N-513-1 as guidallce for evaluation ofllon-plunur leaking. fluws. but relief requests from Codc requiremcllts were still rClluired.lJec(luse tile scope ofN-513-1 was limited by section 3.0 of the Cuse. This revision extcnds the Case tu cover u II lypes of non-planar tlaws. The analysis procedures have been expanded to (J(Jdress the general case ofthrough-\\vall degradation. This revision also include~ the illlprovetl flaw evaluation pmcedurcs for piping ,HH:d to SeetiolJ XI. Appentlix C. in the 2002 Addellda. N-5\\3-1 (98-SI2) (13COO-572) Evaluatiun Criteria for Temporary Acceptance of Flaws ill Moderate Energy Cluss 2l1r 3 Piping, Section XI. Division I TECHNICAL The ('USI.! has been cxpamlcd to penni! applicatioll to Class 2 moderate energy piping. Thc analysis procedures have been expanded to atltlrc$s degratlation mechanisms, slleh as stress-corrosion cracking. that Illay be all issue for Class 2 piping. N-513 (95-S10) (97-208) (CollditionallyAccepwble-ReglilutoryGuide l.1*n-Rev. 14) Evaluatiml Criteria for Temporary Acceptance of FI<lwS in Class 3 Pipillg. Section Xl. Divisioll NE\\V CASE This Case provitles li)r the teillporary accept<lnce of flrlws, includillg through-wall (lerlking) tlu\\vs ill low alld modcrate cnergy Class J pipillg. providing that the conditions of the Case un! satisfied. Acceptance criteria arc based on the same margins as conUlincd in Appendices C and H and Case N-480. The problem with thc euse is that the provisions an~ morc restrictive than the current requirements ill Section III nnd Scctioll Xl. The Case applies only to Class J ClHllpOnCn(s, hut it requires the usc of a Class I typc stress an:dysis to justify the delay of the replacemcnt. Thc Casc is not needcd. because I:llrrent Code requiremcnts provide rules that call be used for Illore cconulll ical evaluations anti repairs. (f~cgulatory Condition (I) Specific surety r~ld()rS in par<lgr,lph 4.0 Olust be s<lti,tieJ. (2) Code Cas\\! N-513 may nor be u[l[llicd to : (~I) Componellts oth.:r than pipe anti tuhe. tel Threaded connections eillploying nOIlS(fLIClUral seal welds for leakage pn:venlioll (through ~cal weltl Icabge is not a structural tlaw; thrcadintcgrity must be maintaincJ). E2 - 10

DOEJ-HRSNCJ41070-S001 Altachmenl2 3/3 (d) Degraded sockd welds.) From: Retherford, Rebecca Sue Sent: Friday, Odobr 21, 2011 5:50 PM To: Nguyen, An N. Cc: Edwards, James A_ (Jim - SNC); Ago/d, James M.

Subject:

FIN: Code Case N-S13-3 An : N-S13-3 is the version approved for use at Hatch and reflected in the lSI Plan Volume 1. Technically, version 3 evaluation is essentially the same as in N-S13-2. The NRC requirement is the requirement is that the permanent repair be done in the next refueling outage. Copy attached. Rebecca From: Retherford, Rebecca Sue Sent: Friday, October 21,2011 04:13 PM To: Altizer, J. Mike

Subject:

Code Case N-S13-3 Mike: Code Case N-5 B-3 is attached. This code case is referenced in the Hatch lSI Plan, Vol. 1 as acceptable for use at Hatch. Rebecca E2 - 11

Frequency of Exposed Pipe

Purpose:

To delermine the nalUral rrequit:nt.:y of Ihe NPS 10, slandard wall. simply supported, 12 fool long. IIlt!rlia := 160.';\\noJ In mass := (40.5 + :14.2) ft 2 It E Inenia rad w:=- --- : 24tJ.6:18-= mass s~c It:n w '".19.7:11111 Prmary slress due to weight (I g):' 2 mass g lell 1 Momcnt :=

= 1.:145x 10" ft Ihr 8

MomenlDia slress.-

5W.68Ipsi 21ncrlia

== Conclusion:== no seismic load required if hending stress of 1500psi is used. DOEJ-HRSNC341070-S001 Att:lchment 3 - III E2 - 12

Edwin I. Hatch Nuclear Plant - Unit 1 HNP-ISI-ALT-14, Version 2.0 Temporary Non-Code Repair of Plant Service Water Piping Documentation of Engineering Judgment (DOEJ)-HRSNC341070-M001, Evaluation of Unit 1 Plant Service Water (PSW) Flow with Pipe Degradation

Southern Nuclear Operating Company DOCUMENTATION OF ENGINEERING JUDGMENT DOEJ-HRSNC341070-M001 Evaluation of Unit 1 Plant Service Water (PSW) Flow with Pipe Degradation Version Record Version Originator/Date Reviewer/Date No. Signature Signature J'. 1.0 Scott Kirk ~_ ~ :J Steve Berryhill Skt;::p & l"(U/O(~o/, Ad:t .~ iu h. Z)z.Ol, U NMP*ES-039-002. Version 2.0 Documentation 01 Engineering Judgment E3 - 1

DOEJ-HRSNC341070-MOOI Southern Nuclear Operating Company

Purpose:

Unit 1 Plant Service Water (PSW) piping has been inspected with degradation discovered. The purpose of this evaluation is to use the PROTO-FLO hydraulic model to simulate through-wall leaks and determine if safety-related components will still receive design PSW flow. Design Inputs (Reference NMP-ES-042):

1. The safety-related components that receive PSW flow post-LOCA are identified in Reference 1 with the design flowrates listed. These components and flowrates are provided in Table 1 below.

References:

1. SMNH-02-012, version 5, Generate Unit 1 Plant Service Water (PSW) PROTO-FLO Database for Latest Test Data
2. RER SNC119724, Sequence 02, Evaluate 9rF River Temperature
3. RER 1100341001, Sequence 03, Main Control Room Air Conditioning PSW Flow Evaluation Assumptions:
  • The PROTO-FLO hydraulic model contained in Reference 1, which was bench marked in 2007, is an adequate representation of the current Unit 1 PSW system for the purpose of this evaluation.
  • For conservatism, the LOCA case is used as defined in Reference 1 (i.e., Technical Specification minimum river level of 60.7 feet MSL, single failure of a diesel generator, all reactor building loads in service, etc.).
  • For conservatism, the river temperature is assumed to be 9r.
  • For conservatism, assume the Division II PSW strainer has a 125 gpm packing leak (reference Attachment 3).

Evaluation: Reference 1 contains the PROTO-FLO file PSW HATCH UNIT 1 2007 LOCA.PDB. This file was used as the basis for this DOEJ. Two new files were created to simulate holes in the 10* Division II supply header to the reactor building. The first case simulates two holes, each equivalent to 2" x 3" (reference Attachment 4), in the Division II supply header to the reactor building, and determines if design flow is still provided to safety-related components that receive PSW. The second case determines the maximum hole size the header can withstand and still provide design flow to safety-related components that receive PSW. For both cases, a 125 gpm packing leak is assumed as flow out of Node 0020. Case 1 (PSW HATCH UNIT 1 2007 LOCA with break.POB) Pipe section 101 in the model is the 10" Division II supply header to the reactor building. In PROTO-FLO, a hole is modeled at a node. In order to model two different holes, pipe section 101.1 and Node 0270A were added since only one hole can be modeled at a single node. The original length of piping was maintained by placing 600 feet in pipe section 101 and 4.2 feet in pipe section 101.1. At Nodes 0270 and 0270A. a hole was modeled on the Nodal Flow tab. For a hole 2" x 3", Area for ellipse = rrRtRz = rr(2I2)(3/2) = 4.71 in2 [Total through-wall leakage area = (2)(4.71 )=9.42 in2.J NMP*ES*039*002. Version 2.0 Documentation of Engineering Judgment. E3 - 2

DOEJ*HRSNC341070*MOOl Southern Nuclear Operating Company Since PROTO-FLO models circular holes, determine equivalent diameter of a circular hole: (rrD2)/4 = 4.71 D =2.45" With two 2.45" holes in the Division II supply header to the reactor building, PROTO-FLO predicts the following flows to safety-related components: Table 1 Design Flow Predicted Flow Component Pipe (GPM) jGPMJ 1 E11C001 A 906 4 8.3 1 E11C001 B 920 4 8.9 1E11C001C 914 4 8.4 1 E11C001D 926 4 9.0 1P41COO1A 38 2 4.7 1P41COO18 53 2 5.1 1P41COO1C 43 2 4.8 1P41COO1D 59 2 5.1 1T41B002A 603 100 217 1T41B002B 612 100 212 1T41B003A 733 100 188 1T41B003B 740 100 185 1T41B004A 672 25 40 1T41B004B 669 25 36 1T41B005A 721 25 38 1T41B005B 724 25 33 1 R43S001 A 139 700 844 1R43S001C 128 700 762 1Z41 B008A (Div. I) 1318 120 100 1Z41 B008C (Div. II) 1332 120 74 All of the safety-related components receive design flow except the control room HVAC units. The Division I HVAC unit (1 Z41-B008A) receives approximately the same flow as indicated in Reference 1. This flowrate was determined to be acceptable as discussed in Reference 1. The Division II HVAC unit (1 Z41-B008C) receives 74 gpm which is less than design. This flowrate is judged to be acceptable because:

  • The current model (Reference 1) has not been revised since control room HVAC pipe replacement and cleaning; therefore the predicted flow is underestimated.
  • Reference 2 evaluated the control room HVAC units for gr water and determined the minimum acceptable flow to be 75 gpm. Current river temperature is 66°; therefore, 74 gpm is judged to be acceptable.
  • Unit 2 provides backup flow, and credit could be taken for the Unit 2 PSW system to perform its safety function of supplying adequate flow to the control room HVAC units.

2 NMP*ES*039*002. Version 2.0 Documentation of Engineering Judgment. E3 - 3

DOEJ-HRSNC341070-MOOl Southern Nuclear Operating Company Case 2 (PSW HATCH UNIT 12007 LOCA with maximum breakPDB) Pipe section 101 in the model is the 10" Division II supply header to the reactor building. A single hole is modeled at Node 0270. Since the control room HVAC units are already receiving less than design flow in Case 1, and those flows have been evaluated'for acceptability, the hole size will continue to be increased until the next safety-related component reaches its design flow. By trial and error, with a single 3.97" diameter hole in the Division II supply header to the reactor building, PROTO-FLO predicts the following flows to safety-related components: Table 2 Design Flow Predicted Flow Component Pipe (GPM) (GPM) 1E11 C001A 906 4 8.3 1E11C001B 920 4 8.6 1E11C001 C 914 4 8.4 1E11 C001 D 926 4 8.6 1P41C001A 38 2 4.7 1P41C001B 53 2 4.9 1P41 C001C 43 2 4..8 1P41C001D 59 2 4.9 1T41B002A 603 100 217 1T41B002B 612 100 213 1T418003A 733 100 170 1T41B0038 740 100 168 1T418004A 672 25 41 1T41B0048 669 25 36 1T41B005A 721 25 34 1T4180058 724 25 30 1R43S001A 139 700 844 1R43S001C 128 700 725 1Z41B008A (Div. I) 1318 120 100 1Z41 B008C (Div. II) 1332 120 64 All of the safety-related components receive design flow except the control room HVAC units. The Division I HVAC unit (1Z41-B008A) receives approximately the same flow as indicated in Reference 1. This flowrate was determined to be acceptable as discussed in Reference 1. The Division /I HVAC unit (1Z41-B008C) receives 64 gpm which is less than design. This flow rate is judged to be acceptable because:

  • The current model (Reference 1) has not been revised since control room HVAC pipe replacement and cleaning; therefore the predicted flow is underestimated.
  • Reference 3 evaluated the control room HVAC units for reduced flow and determined the minimum acceptable flow to be 63.9 gpm at a maximum water temperature of 91.8° 3

NMP*ES-039-002. Version 2.0 Documentation ot Engineering Judgment. E3 - 4

DOEJ-HRSNC341070-MOOI Southern Nuclear Operating Company cooling water (with margin limitations as discussed in Reference 3). Current river temperature is less than 91.8°; therefore, 64 gpm is judged to be acceptable.

  • Unit 2 provides backup flow, and credit could be taken for the Unit 2 PSW system to perform its safety function of supplying adequate flow to the control room HVAC units.

Determine maximum throu~h-walileakage area based on 3.97" hole: Areamax = (n02)/4 = n(3.97 )/4 =12.38 in2

== Conclusion:== With two 2" x 3" holes (modeled as 2.45" diameter holes for a total through-wall leakage area of 9.42 in2) in the PSW Division" supply header to the reactor building, PROTO-FLO predicts all safety-related components will receive design flow, with the exception of the control room HVAC units. By judgement, the control room HVAC units will receive suHicient flow to perform their safety function. With a maximum through-wall leakage area ot 12.38 in2 in the Division II supply header to the reactor building (modeled as a single 3.97" diameter hole), PROTO-FLO predicts all safety related components will receive design flow, with the exception of the control room HVAC units. By jUdgement, the control room HVAC units will receive suHicient flow to perform their safety function. List of Attachments:

1. PROTO-FLO summary report for Case 1
2. PROTO-FLO summary report for Case 2
3. Email from Eric King to Scott Kirk, October 22,2011, with strainer packing leak flow
4. Email from Eric King to Scott Kirk, October 21, 2011, with pipe hole sizes 4

NMP*ES*039*002, Version 20 Documentation of Engineering Judgment, E3 - 5

10/2212011 13:22 PROTO-FLO 4.60 by Proto-Power Corporation - Serial IIPFL-I 016 Page I of I Southern Nuelear-P:IModclsl!latchISMNfI-02-012IPSW HATCH UNIT 12007 LOCA with break.POB - Version 4.6 DOEJ-HRSNC341070-MOOI - Allachmenl I Flow Summary Report ConVCfR<nce: P,cssurc=1.0E*5 Sum <PI.OE*! friclion-1.0E*6 FCV= I 0E-4 PCV. 1.0E*3 TCmll<l'l!lurc-5.0E-3 Flow Summary Title Diameter Flow Minimwn flow NPSHA NPSH (in) (gpm) (gpm) (ft) Ratio Pipe 38 0.493 4.74 2.0 Pipe 43 0.493 4.75 2.0 Pipe 53 0.493 5.07 2.0 Pipe 59 0.493 5.08 2.0 Pipe 128 6.065 762.35 700.0 Pipe 139 6.065 844.42 700.0 Pipe 603 2.469 216.69 100.0 Pipe612 2.469 212.30 100.0 Pipe 669 1.939 35.92 25.0 Pipe 672 1.939 40.46 25.0 Pipe 721 1.278 37.94 25.0 Pipe 724 1.278 32.86 25.0 Pipe 733 2.469 187.78 100.0 Pipe 740 2.41i9 I R5.43

  • 100.0 Pipe 906 0.493 8.33 4.0 Pipe914 0.493 8.35 4.0 Pipe 920 0.493 8.94 4.0 Pipe 926 0.493 8.99 4.0 Pipe 1318 3.068 99.62
    • < 120.0 Pipe 1332 3.068 743~
    • <: 120.0 II RevLTIc Flow Th",u~h Check Valve

&& Pll mp Flow i< Pas\\ End ofCllrvc

    • Flow Below Minimum SS NPSH Available Below NPSH Required E3 - 6

J 0/22/20J I 13:39 PROTO-FLO 4.60 by Proto-Power Corporation - Serial #PFL-1016 Page I of I Southern Nuclear - P:\\MODELSIHA TCH\\S~fNH-02-0 12\\PSW HATCH UNIT J 2007 LOCA WITH MAXIMUM BREAK.PDB - DOEJ-HRSNC341070-Mool - At!~elunen! 2 Flow Summary Report COnYC,"..,,' C: Pres...,,,,,, t.OE*S Sum Q-I.OE*2 Friclion-I.OE*6 Fev= I.or.... PCV~t.OE*] Tempe""tu..=S.OE.) Flow Summary Title Diameter Flow Minimwn Flow NPSHA NPSH (in) (gpm) (gpm) (n) Ratio Pipe 38 0.493 4.74 2.0 Pipe 43 0.493 4.75 2.0 Pipe 53 0.493 485 2.0 Pipe 59 0.493 4.87 2.0 Pipe 128 6.065 725.44 700.0 Pipe 139 6.065 844.33 700.0 Pipe 603 2.469 216.91 100.0 Pipe 612 2.469 212.51 100.0 Pipe 669 1.939 35.96 25.0 Pipc 672 1.939 40.50 25.0 Pipe 721 1.278 34.30 25.0 Pipe 724 1.278 29.71 25.0 Pipe 733 2.469 169.98 100.0 Pipe 740 2.41\\9 167.R5 1000 Pipe 906 0.493 8.33 4.0 Pipe914 0.493 8.35 4.0 Pipe 920 0.491 8.57 4.0 Pipe 926 0.493 8.62 4.0 Pipe 1318 3.068 99.70

    • <: 120.0 Pipe 1332 3.068 63.92
    • <: 120.0

" Reverse Flow Through Check Valve && Pump Flow is PosC End of Curve

    • Flow Below Minimum S5 ~PSH Available Below NPSH Required E3 - 7

DOEJ*HRSNC341070*M001 Attachmenl3 Page I Kirk, Scott From: King, John Eric Sent: Saturday, October 22, 2011 9:29 AM To: Kirk, Scott

Subject:

PSW Leakage Calculations 2 Holes have been identified in the Unit 1 Division" PSW Reactor Building Header Hole 1: Hole l is round with 1-1/8" diameter Flow Q = .6"A"sqrt((2*g"P)/p) Pressure P = 180 Ibfjin"2 Density p = 0.03657 Ibf/in"3 Gravity g = 32.174 ft/s"2 Area Hole 1 A = pi"[d"2)/4 = 0.99402 in"2 Hole 1 Diameter d = 1.125 in Conversions 12 in/ft 0.004329 gal/ln"3 60 s/min Hole.1 Flow Q= 302 gpm Hole 2: Hole 2 is elliptical with a major axis of 2-1/4" and a minor axis of 2" Flow Q = .6* A "sqrt((2 "g" PI/pI Pressure P = 180 Ibfjin"2 Density p = 0.03657 Ibf/in"3 Gravity g = 32.174 ft/s" 2 Area of 1 hole A = pi*a"b. 3.534292 in"2 Major Axis/2 a = 1.125 in. Minor Axis/2 b = 1 in Conversions 12 in/ft 0.004329 gal/in"3 60 s/min Hole 2 Flow Q= 1074 gpm A packing leak on PSW strainer 1P41D1038 The hole is round with a 2-1/2" diameter The shaft is round with a 2.393" diameter Flow Q= .6*A *sqrt( (2' g" P)/p) Pressure P= 180 Ibf/in'-'2 Den sity p= 0.03657 Ibf/in'-'3 Gravity g= 32.174 ft/s" 2 Hole Area Ah = pl*(dh '-' 2)/4 4.908739 in'-'2 Shaft Area As= pi'(ds"2)/4 = 4.497543 leak Area AI = Ah*As => 0.411196 E3 - 8

DOEJ*HRSNC341 070*MOO 1 AUachment 3 Page 2 Hole Diameter dh= 2.5 in Shaft Diameter ds = 2.393 in Conversions 12 in/ft 0.004329 galjinJ\\3 60 s/min Strainer Flow Q= 125 gpm All Leakage Considered Total Flow Q:: 1501 gpm Leakage Impacts: Pump Capacity Qp= 8500 gpm From flow model Required Flow Qr= 4428 gpm From H16012 Leakage Flow QI= 1501 gpm From above Flow Margin Qm= Qp-Qr 4072 gpm Acceptable if Qm > QI Is Qm > QI? Yes E3 - 9

DOEJ*HRSNC341070*MOOI Altachmenl4 Page' Kirk, Scott From: King, John Eric Sent: Friday, October 21, 2011 1:45 PM To: Kirk, Scott

Subject:

Unit 1 Division II PSW Leak Scott, Could you help me figure out the effect that a leak would have on the 10" Unit 1 Division II PSW line to the reactor building? There are 2 pits found on the pipe. Both pits have an oval shape and are 2" x 3" in size. There is currently through wall leaks at both pits. For the operability evaluation, we are assuming that the entire pit is a through wall leak. calculated the now from this size hole to be 2863 gpm. See the calculation below: (2) Holes Each hole is 2"x3" elliptical Flow Q = .6*A*sqrt((2*g*P)/p) Pressure P= 180 Ibf/inll2 Density p= 0.03657 Ibf/inll3 Gravity g= 32.174 ft/sll2 Area of 1 hole A= pi*a*b 4.712389 inll 2 Major Axls/2 a = 1.5 in Minor Axls/2 b= 1 in Conversions 12 in/ft 0.004329 galjinll3 60 s/min 1 Hole => Q= 1432 gpm 2 Holes => Q= 2863 gpm

Thanks, Eric E3 - 10

Edwin I. Hatch Nuclear Plant - Unit 1 HNP-ISI-ALT-14, Version 2.0 Temporary Non-Code Repair of Plant Service Water Piping Documentation of Engineering Judgment (DOEJ)-HRSNC341070-S002, Evaluation of HNP Unit 1 Plant Service Water Pipe Leaks per ASME Code Case N-513-2 and N-513-3

Southern Nuclear Operating Company DOCUMENTATION OF ENGINEERING JUDGMENT DOEJ-HRSNC341070-S002 Eyaluation_QfllNP_Unitl.PlantSeLyice_Water-Pipe Leaks per ASME Code Case N-513-2 and N 513-3 Version Record Version Originator/Date ReviewerlDate No. Signature ,(, Signature 1 An Nguyen / October 26, 2011 'II ~ S. H. Pellet / October2<J~1f::-I~.r NMP ES 030*00r, Vai5ion 2.0 E4 - 1 Documentation of Engineering Judgment

DOEJ-HRSNC341070-S002 Southern Nuclear Operating Company

Purpose:

The purpose of this DOEJ is to support RER SNC34 1070. The scope for this DOEJ is the evaluation of the piping structural integrity for the HNP Unit I plant service water system. This piping has developed through wall seepage and UT inspection has been performed. Design Inputs (Reference NMP-ES-042): I. S00631.

2. S00779.
3. Attachment to E-mail from Kevin White to An Nguyen, 10/21111 (Attachment I shows the sketch redrawn with additional data, 10-26-2011)
4. Exposed Piping Evaluation (Attachment 3)

References:

I. Code Case N-SI3-2 and N-S13-3.

2. ASME Section XI, 2003 (Code of Record).
3. ASME Section XI, 20 IO.
4. RASEARCH Results for N-S13-2 and N-S13-3 (Attachment 2)

Assumptions: In this evaluation, a representative tlaw geometry enveloping the geometry of the two tlaws was used. From the UT report (attachment I), the tlaw can be characterized as 2.S inch in the circumferential direction and 2.2S inch in the longitudinal direction. The minimum pipe wall thickness-outsid-of-lhe--fla w-and-wi thin-the-area-of-rei nforeemen Hs-at--Ieast- (};-1inc h-------- The design pressure and design temperature are 180 psig and 12soF, respecti vely. The piping system was classified as buried pipe. As such, this piping system does not have stress calculation. Now the pipe is exposed in the pit. The exposed piping was evaluated as shown in attachment 3 of this DOEJ. Since the temperature of piping system is low << ISO°F), no secondary stress evaluation is required. Evaluation: This evaluation is in accordance with Code Case N-SI3-2 and SI3-3. The difference between the two versions of the code case is not applicable to this case as discussed in reference 4. The exposed pipe span is approximately 12 ft long. The natural frequency is calculated to be -40 hz; hence, there is no concern for seismic. A conservati ve value of ISOOpsi for bending stress was considered for static (weight) loading. Frequency calculation and primary stress due to weight and seismic are shown in attachment 3. The approach is per paragraph 3.2(c) of the code case. This method uses the branch reinforcement m~Jhoq~s sp9WRtn ~na~b.m~nt 4ofl~i~QOEJ,. NMP*ES-039-002, Version 2.0 Documentation of Engineering Judgment. E4 - 2

DOEJ-HRSNC341070-S002 Southern Nuclear Operating Company Note: This approach does not require the temperature to be higher than the upper shelf temperature of the material as discussed in DOEJ-HRSNC341070-S00 I and, therefore, is not dependent on the temperature of the service water tlowing through the pipe. ft is noted that the 0.078 inch thick indication was found recently nearby the leak in Area 2, as shown in Attachment I. This indication does not encroach in the reinforcement area, and the remaining thickness is more than the minimum code thickness of 0.064. Hence, this indication has no effect in this evaluation.

== Conclusion:== Current flaw configurations meet the criteria for temporary acceptance of flaws in moderate energy class 3 piping system. Therefore the pipe in both Areas I and 2 continues to meet Code requirements. This evaluation is in accordance with Code Case N-S13-2 and N-SI3-3. Hence, the following compensatory actions are also required: Compensatory Measures Daily monitoring of leakage for noticeable changes UT - at least monthly based on no noticeable leakage change List of Attachments: J. UT Results.

2. RASEARCH Results
3. -Exposed-Pipe-Evaluation:---
4. MathCAD calculation.

2 NMP-ES-039-002. Version 2.0 Documentation of Engineering Judgment. E4 - 3

DOEJ*HRSNC341070*S002 Attachmenl 1 1/1 UPDATED: 10126120 II Hatch 1 PSW Buried Pipe - 10/21/11 Hatch 1 - 10" Plant Service Water UT around areas of leakages located in excavation number 2, where 9 feet of pipe was exposed. Reference 5-00631/ Hl1139. Leakage from area number 1 - is located top dead center of the pipe at ~ 2 feet from the north end of the dirtwall. The area was gridded 148 for UT FAC inspection. Leakage from area number 2 - is located 1/2 feet from the north dirt wall at 7:00 when facing north. The area was gridded F40 for UT FAC inspection. East Side 1.25" (.3) .625 (.2) North End 1.375(.3).50(.2) + .625(.2) 1.375(.3) South End .50(.2) 1.75(.3) West Side Area 1 Up 1.5(.3) None(.2) North End 3(.3) 1.0(.2) + 1.25(.2) 2.5(.3) 3.25(078) 45(>.2) South

  • End

.5(.2) 1.25(.3) Down + :: center of the leak Inches from the leak (thickness at that location) . North andSl~th i5-the lQngi-tu-ainfrl s)(haf th!!* pipe:. Most other areas were difficult to obtain UT readings of.100 or less. The examination was performed using a USN 60 with a 5.0 MHz 3/8" dual transducer where a one inch screen range was established. E4 - 4

DOEJ*HR5NC341 070*5002 Altachmenl 2 113 Nguyen, An N. From: Nguyen, An N. Sent: Friday, October 21,2011 7:46 PM To: Retherford, Rebecca Sue Cc: Edwards, James A. (Jim - SNC); Agold, James M.

Subject:

RE: Code Case N-513-3 Thank you. Below is the commentary on N-513-3. An Nguyen, PE Telephone: 8-992-7307 Ci:lthered from Rnscarch NUC Fiks\\ Revisioll - Nuclear Cases.wpd (7/16/20 10)

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Code Case Revisions N-513-3 (07-S8) (07-1303) Evaluntion Criteria for Temporary Acceptance of Flaws in Moderate Ellergy Class 2 or} Piping, Section xr, Division I TECHNICAL Th is revision provides signi fic;-\\I1t clarifications regarding evaluation of through-wall, nonplanar tla\\vs, wh ich arc the flaw type mosr commonly dispositil)l1ed lIsing this Casco The acceptability criterion for the prior branch rcinfi.)f(;el11cnl evaluation approach (b;-\\scd on ASME Scction 1[1 Class 2 emd } rules) was amhiguous. This revision ~pecifies that the required area of reinforcement is 10 he calculated in accordance with Class I rules, and proves I1C\\V (acceptance crireria ti.)!" this approach. Also, thc depth at which a through-wall, nonplall;-\\r !law is Cl1aracterizcd for planar evaluation in both the axi;-i1 and circllill feren tial dircl.:tions is made less restrictive ill the proposed revision. to aCCOUJlt for NDE I.:apabilities. A new equation is introduced to address the potential for pressurc blowout if an area larger than the current through-wall, non planar leak is evaluated to provide a bOlilld illg analysis. N-513-2 W4-S I) (8C03-249) (Accept.lble - Regulatory Guide 1.147 - Rev. 15) Evaluatiol1 Criteria ti.1rTelllporary Accept;-\\Jlce of Flaws ill Modemte Energy Class 2 or 3 Piping. Section XI. Division I TECHNICAL This revision adds a procedure for cv;-\\luation of non-plnnar through-wall tbws in moderate encrgy piping. Service cxpcrience has shown that some piping suffer degradation frollllloll-rl'lnar flaws. slleh as pitting and E4 - 5

DOEJ-HRSNC341070-S002 Attachment 2 2/3 microbiological attack. where local inconsequential leakage ean occur. Some Owners have used N-SI3-1 as guidance for evaluut ion of non-planar leaking flaws. bur relief requests from Code requirements were still rcquired. because the scope ofN-SIJ-1 was limited by section 3.0 of the Case. This revision extends the Case to cover all types of non-planar tlaws. The analysis procedures h:lVe been expanded to address the general case of rhmugh-wall degradation. This revision also includes the improved !law evaluation procedures for piping added to Section Xl, Appendix C. in the 2002 Addenda. N-SI3-1 (n-S 12) (BCOO-572) Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping. Section Xl, Division I TECHNICAL The Case has been expanded to permit applicalion to Class 2 moderate energy piping. The analysis procedures have been expanded to address degradation mechanisms, slich as stress-corrosion cracking, that may he an issue for Class 2 piping. N-513 (95-S 10) (97-208) (Conditionally Acceptable - Regulatory Guide 1.147 - Rev. 14) Evaluation Criteria tor Temporary Acceptance of Flaws in Class 3 Piping. Section XI. Division I NEW CASE This Case provides for the temporary acceptance of !laws, including through-wall (leaking) tlaws in low and moderate energy Cla:;s 3 piping. providing that the conditions of the Case are satisfied. Acceptance criteria are based on the same margins as contained in Appendiccs C and H and Case N-4~0. The problem with the Case is that the provisions are more restrictive than the eurrent requirements in Section IJI and Section XI. The Case applie:; only to Class 3 components, hut it requires the use ofa Class I type stress an ~dysis to justify the delay of the replacement. The Case is not needed, because current Code requirements provide rules lhat can be Ilsed for more cconomical evaluati()fls and repairs. (Regu latory Condition t I) Specitic safety factors in paragraph 4.0 must he satisficd. (2) Codc Case N-5 tJ may not be applied to: (a) Components other than pipe and tube. (c) Threaded cOtlnections employing tlonslructural seal welds for lcabge prevention (through seal weld leakage is not a structural tlaw; threadinlcgrity must be maintaincd). 2 E4 - 6

DOEJ*HRSNC341 070-S002 Attachment 2 313 (d) Degraded socket welds.) From: Retherford, Rebecca Sue Sent: Friday, October 21,2011 5:50 PM To: Nguyen, An N. Cc: Edwards, James A. (Jim - SNC); Agold, James M.

Subject:

FW: Code case N-S13-3 An: N-S13-3 is the version approved for use at Hatch and reflected in the lSI Plan Volume 1. Technically, version 3 evaluation is essentially the same as in N-S13-2. The NRC requirement is the requirement is that the permanent repair be done in the next refueling outage. Copy attached. Rebecca From: Retherford, Rebecca Sue Sent: Friday, October 21,2011 04:13 PM To: Altizer, J. Mike

Subject:

Code case N-S13-3 Mike: Code Case N*S13-3 is attached. This code case is referenced in the Hatch lSI Plan, Vol. 1 as acceptable for use at Hatch. Rebecca 3 E4 -7

Frequency of Exposed Pipe Purposl.! : frequency To de(l.!rmin..: (he natural frequiency of the NPS 10, standard wall, simply supPol"led, 12 fOOl long. E := :n.f> t d'psi Inl.!rlia := 160.nll~ 10 mass := (40.5 + 14.2) ft . :? It E tncrtiJ rad w:=- --- = 24l).6.\\~ ') mass SeC iell w" :W.7.1Ihz Prmary stress uue to w\\!ighl (I g): mass g len-" 1 MOllwnt := = I J45x 10" fllbf R Momen( Diu Slress.- =5W.681psi 2Inen ia

== Conclusion:== no seismic load required if bending stress of 1500psi is used. DOEJ-HRSNC34J070-S001 Attachment J - 111 HRSNC341070-S002 E4 - 8

DOEJ-HRSNC341070*S002 Altachment 4 1/1 Code Case N-S13-3 Using Branch Reinforcement Method 0r-------------------------------------------------------------------

Purpose:

This MathCAD demonstrates the use if branch reinforcement method to address leakage per code cast: N-513-3. This evaluation is per paragraph 3.2(c) of the code case. Input: P := 180psi Design Pressure S := 15ksi Allowable stress tadj :== 0.2in Minimum thickness dett:mlint:d outside of flaw erircle of diameter dadj dadj := 2.25in Diamt:ler of flaw area D :== 10.75 in Outsidt: diameter of pipe PD

  • tmin*.;== *

. = O:064*in**** ** code minimumwall*thicknesS'** 2(S + AP) D - tadj Rm :==

== 5.275 in Mean pipe radius 2 I d' (tadj - tmin) dadjall - 1.5y Rm ta J

== 3,26 in tmm tc_avg .353dadj H

== 0.087 in

== Conclusion:== This approach allows a thinning area enclosed by a circle of diameter dadj = 2.25 in of diameter provided the remaining wall is thiekt:r than tadj== 0.2 in and the average thickness in the flaw arca is tc_ avg== 0.087 in. In this case, tc_avg is not applicable. Therefore, the through wall flaw identified in area I and 2 of attachment I to DOEJ-HRSNC*341070*S002 are acceptable and continue to mt:et the Code requirements. The /law in Area 2 that is 0.078 inch thick and 3.25 inch south of the leak is outside of the reinforcement area and the thickness exceeds tmin of 0.064 inch, Therefore, this flaw also acceptable. E4 - 9}}

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