ML20086J676
| ML20086J676 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 07/13/1995 |
| From: | Marsh W SOUTHERN CALIFORNIA EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20086J679 | List: |
| References | |
| GL-90-05, GL-90-5, NUDOCS 9507190172 | |
| Download: ML20086J676 (20) | |
Text
{{#Wiki_filter:F ] I feV5 Southem Califomia Edison Company l P. O. BOX 128 SAN CLEMEN'lI , CAUFOHNIA92674 4128 WALTER C. MARSH 1EuPHcmE l U. S. Nuclear Regulatory Commission . Attention: Document Control Desk Washington, D.C. 20555 Gentlemen: i
Subject:
Docket No. 50-362 l Second Ten-Year Interval Inservice Inspection Program i San Onofre Nuclear Generating Station , Unit 3 l Gentlemen: i This letter submits Relief Request (RR) D-2-01 (enclosed) for NRC approval. I RR D-2-01 requests relief from the requirement to perform immediate repairs in J accordance with ASME Section XI, IWA 4000. This request is made using the
< criteria in Generic Letter (GL) 90-05 for Class 3 moderate energy piping and the guidance provided by the NRC Project Manager. ;
A through-wall leak was discovered in the line from the boric acid makeup l pumps to the charging pump suction in San Onofre Unit 3 on June 19, 1995. This piping is Quality Class II, ASME Section III Class 3, and the system is a moderate energy system. Using the "Through-Wall Flaw" approach described in 2 GL 90-05, the suction line to the charging pump is detennined to be operable . in accordance with the guidance of GL 91-18. Unit 3 is scheduled to shutdown r for a refueling outage on July 22, 1995, 33 days from the discovery of the leak. Therefore, the purpose of this relief request is to defer the Code repair until this refueling outage. In the interim, periodic inspections are being performed in accordance with GL 90-05 to insure this line remains operable. If you have any questions, please call me. j
. Sincerely,
& G/
Enclosure
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i fl ' 9507190172 9507:3 PDR ADOCK 05000362 NI I Q PDR - 8 _ m
(- _ _ i wd:W?l Document Control Desk
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cc: L. J. Callan, Regional Administrator, NRC Region IV A. B. Beach, Director, Division of Reactor Projects, Region IV K. E. Perkins, Jr., Director, Walnut Creek Field Office, NRC Region IV J. A. Sloan, NRC Senior Resident Inspector, San Onofre Units 2 & 3 M. B. Fields, NRC Project Manager, San Onofre Units 2 and 3 e l l L ]
RELIEF REQUEST D-2-01 San Onofre Nuclear Generating Station Unit 3, Docket 50-362 Page 1 of 9 SYSTEM: Boric Acid Make-up (BAMU) COMP 0NENT/ AREA: BAMU pump discharge to the charging pump suction, 3" line upstream of check valve S31208MUO83 which is unisolable. REPAIR CATEGORY: IWA-4000 CODE CLASS: 3 FUNCTION: Boron injection flow path. REPAIR REQUIREMENT: Removal of the flaw and subsequent weld repair. BASIS FOR RELIEF: The following basis for relief satisfies the criteria specified in Generic Letter 90-05, " Guidance for Performing Temporary Non-Code Repair of ASME Code Class 1, 2, and 3 Piping." The guidance provided by the NRC Project Manager for Units 2 and 3 was used to prepare this basis.
- 1. A Code repair involves the replacement of a section of schedule 10 stainless steel pipe which is attached to an anchored pipe support and runs through both the pipe support and a 2' thick wall. Completion of this repair during the 72-hour Technical Specification Action Statement was not feasible in the time immediately following the discovery of the flaw.
Additionally, completion of the total repair within 72 hours is questionable because of the restricted access for the line replacement and the associated freeze seal. With the refueling outage scheduled to start 33 days from the date the flaw was discovered, Edison evaluated the flaw to determine acceptability until the refueling outage.
F l RELIEF REQUEST D-2-01 ! San Onofre Nuclear Generating Station ; Unit 3. Docket 50-362 Page 2 of 9 i
- 2. Description of the characteristics and size ot' the flaw l
- a. On June 19, 1995 plant engineering a personnel noted leakage from BAMU line i S31218MLO31. A single through-wall crack ;
initiating from the toe of a solid anchor i attachment weld was identified. The crack is a 9/16" long linear indication, running i lengthwise 30 to 40 degrees off horizontal l of the pipe centerline (A drawing is in the NDE results contained in Attachment 3.). There are no eroded areas around the i flaw. There are no additional detectable .; flaws in the area surrounding the crack. i The adjacent wall thickness of the schedule 10S stainless steel pipe was l measured to be approximately 0.130", as determined by ultrasonic examination. A-total area approximating a 4" wide band : around the pipe, encompassing the crack, an adjacent support attachment weld, and a downstream elbow weld, was examined. , t
- b. The approximate location of the leak is ,
provided in Figures 1, 2, and 3 in , Attachment 1. I
- c. Visual, dye penetrant, dye penetrant !
developer " blot testing," and radiographic imaging were used to characterize and ! evaluate the flaw. Ultrasonic examination
- was used to determine wall thickness. t Blot testing consists of a proceduralized l method for using developer spray to evaluate wet indications where dye penetrant cannot be used effectively.
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RELIEF REQUEST D-2-01 l San Onofre Nuclear Generating Station l Unit 3 Docket 50-362 i Page 3 of 9_ j l l 1
- 3. A description of the pipe and system i
- a. S31218MLO31, the line upstream of !
S31208MUO83, is part of the Boric Acid Mix ! and Storage System (Start-up System l designation BGB). This line is part of l the system which supplies boric acid (4021 i ppm to 6119 ppm) from the discharge of the l Boric Acid Makeup Pumps Pl74 and P175 to i the suction of Charging Pumps P190, P191, and P192. This is one of three safety i related sources of boric acid to the l suction of the charging pumps. The other two sources, that open on a safety injection actuation signal, are the , gravity feeds from both boric acid tanks l and the refueling water storage tank. ; i
- b. Design Pressure: 150 psig j l
Operating Pressure: 100 psig with the boric acid pump operating, and ! approximately 50 psig under normal ; conditions. Design Temperature: 1750F f Operating Temperature: ambient [
- c. Pipe Material: ASME SA376 GR T304 (ASTM A376); seamless austenitic stainless steel !
pipe Diameter: 3" (nominal) 3.500"(outer) i 3.260"(inner) , Schedule: 10S
)
Liner Material: N/A l l Wall Thickness: 0.120" { i 1 i l
l l l RELIEF REQUEST D-2-01 San Onofre Nuclear Generating Station Unit 3. Docket 50-362 Page 4 of 9 i i i
- d. Copies of Process and Instrumentation :
Drawings (P& ids) 40125AS03, 40125B503, 40124BS03 are in Attachment 2. j
- 4. We experienced a crack in a Unit 2 Boric Acid l line at a similar weld location and -
configuration in 1991. The failure mechanism was identified as Intergranular Stress Corrosion ! Cracking (IGSCC) based on metallurgical examination. The root cause was sensitization in the heat affected zone (HAZ) resulting from ' the use of 304 SS (high carbon) and the thin j section (schedule 10, thickness 1/8"). This ; combination is vulnerable to overheating, i enlarging of the heat affected zone, and i sensitization of the material. The damaged ! section was replaced with schedule 40 SS. l The root cause of the present crack is l uncertain. It is thought that IGSCC is a strong l primary candidate cause. Residual stresses in : the HAZ of the thin wall pipe coupled with I cyclic fatigue caused by system vibration ; inherent to the operating triplex positive l displacement charging pumps are possible i contributing causes. The root cause will be ' determined after detailed metallurgical i examination of the cracked section following l pipe replacement, j
- 5. There is no safety significance associated with l this crack, as catastrophic failure is not considered likely based on fracture mechanics :
analysis. In addition, the refueling water ' storage tanks and the gravity feed lines from the boric acid tanks are available to provide ; borated water to the reactor core'via the ; charging pumps if needed for safe shutdown. !
- a. Flooding i The crack is not projected to grow to ,
where a significant boric acid leak would > be created. The leakage rate is not ; expected to grow past the current ; magnitude of seepage or dripping.
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P175 FIGURE 3 l DRAWING NUWBIR DESCRIP110N N0 i JUNE 30, 1995 7:26 REFERENCES _ &iM OD E L II L E unrus ST AT!!s
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CHARGING PUAP , NO. 2 f i, [' * ':[ i[, URE 6 A AP",...~1.,. t.z - ; -t'e }--fX}- X -i F HF H l-g
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V \ / CHARGING PDAP CHARGING PUMP g l INLET DISCHARGE w ACCUJULATOR ACCUAULATOR Y W247 W219 f' 7 a REGENERATIVE HEAT EXCHANGER E063 d 008-2"--C-FE0 ) ) 40123ASO3 (D-1) ) cSD7MOl7ZrO OC II UNIT 3 SAN ON0FRE NUCLEAR GENERATING STATION REACTOR COOLANT CHEMICAL AND VOLUME CONTROL SYSTEM SOUTHERN CALIFORNIA EDIS0b $ESCRIPTION RE IRE GS DM CHK OTHER DATE f /lll'E- _. J. .[YV. REVISIONS scale: NONE DRAWN Bf:
ATTACHMENT 1 RELIEF REQUEST D-2-01 FIGURES 1, 2, AND 3
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m i l l l I CONCRETE WALL I PENETRATION 4% h 4.s s - 2.- i ISO VIEW FIGURE 1 S01JTHERN CALIFORNIA EDIS0N J=C-i
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11.00 = 1 SIDE VIEW 4 l 1 FIGURE 2 LINE NUMBER 031-3"-D-LLO i o 3-E- SOUTHERN CALIFORNIA EDISON i
ATTACHMENT 2 RELIEF REQUEST D-2-01 PalDs 40125AS03, 40125B503, 40124B503 r I
i. RELIEF REQUEST D-2-01 San Onofre Nuclear Generating Station Unit 3. Docket 50-362 Page 5 of 9
- b. Spray The crack is located in a valve room enclosed on three sides that contains no electrical components. The open side of the valve alley faces a concrete wall where there are also no electrical components. Although the crack is not projected to grow to where a spray would be created, a postulated spray would not directly impact any electrical components.
- c. Loss of flow Catastrophic failure of the line is not considered credible. The current dripping of the leak is insignificant based on system flow requirements.
- d. Other interactions There are no other interactions anticipated.
- e. System-failure consequences Fracture mechanics models do not project a line or system failure.
- f. The effect on safe shutdown capability The capability of the boric acid and charging system is not projected to be affected by this small crack.
- 6. Inspection of similar lines did not show any indications of cracking. The root cause is thought to be related to the welding of schedule 10 pipe (See Item 4 above). Based on the maximum allowable crack size, it is expected that any crack will show up as leakage and be evaluated well before it reaches critical crack size.
i l RELIEF REQUEST D-2-01 San Onofre Nuclear Generating Station Unit 3, Docket 50-362 Page 6 of 9
- 7. a. There is no overall degradation of the BAMU or charging systems. Technical analysis of the cracked line and apparent ;
root cause concluded the line will remain structurally sound for all design conditions. Additionally, it was determined that present and postulated ; leakage will not adversely impact system inventory requirements or degrade surrounding components due to leakage mitigating measures and the nature of the fl aw. The crack is stable, as determined by analysis and confirmed by ongoing surveillance. The flaw is predicted to remain stable for greater than the required remaining operating time of approximately 33 days from when the crack was discovered until Unit 3 will shutdown for a scheduled refueling.
- b. Five (5) augmented surveillance locations were visually examined on June 28, 1995, following discovery of the Unit 3 flaw on June 19, 1995. The augmented inspection scope was established as a visual inspection of these five additional locations. The plan included provision for Non Destructive Examination / Penetrant Test (NDE/PT) examination if visual inspection indicated leakage or cracking.
PT examination would have been followed by more sophisticated examination techniques if additional flaws were identified. The five additional locations were selected for visual examination based on the candidate root causes. Four of these augmented surveillance locations were in Unit 3 in the same general area of the , flawed line, with one of these on the same line approximately 3' upstream. A solid anchor support in Unit 2 on the BAMU line comparable to the Unit 3 flawed line was the fifth location. No evidence of leakage or degradation was observed.
RELIEF REQUEST 0-2-01 San Onofre Nuclear Generating Station Unit 3, Docket 50-362 Page 7 of 9
- c. Based on the candidate root cause(s), the selected locations were similar in configuration (solid anchor supports and guided supports with weld attachments to pipe, 304 SS, schedule 10 pipe) and system operating characteristics (temperature, vibration, and boric acid concentration).
Final selection and prioritization were based on technical review of the line stress isometric drawings and support detail drawings.
- d. The five locations for augmented inspection were visually examined on June 28, 1995. No indications were observed.
- e. Presently, no requirement exists for an expanded augmented inspection plan. This is based on the absence of indications in the examined locations described above.
- 8. A drip-bag liquid collection device is installed to collect and channel leakage away from the crack location to a suitable floor drain. The bag prevents water from dripping onto adjacent supports. Plant Equipment Operator shift rounds are used to maintain frequent monitoring of the crack leakage.
- 9. The ASME Code repair will be made during the Cycle 8 refueling outage. This outage is scheduled to begin on July 22, 1995.
- 10. The flaw evaluation operability assessment and I supporting calculations are in Attachment 3.
The safety assessment concluded the crack is acceptable based on an evaluation of the actual crack size versus a fracture mechanics analysis of the critical stable crack size. The actual crack size of 9/16 inch is acceptable when I compared to the fracture mechanics results, which indicate the conservative limiting crack size is a 1.5 inch circumferential crack. The i
q.. . . . . . l RELIEF REQUEST D-2-01 San Onofre Nuclear Generating Station Unit 3 Docket 50-362 Page 8 of 9 ! evaluation concluded that the existing crack will remain stable and will not propagate to failure undcr design basis considerations. The flaw size is not expected to reach the limiting size within the 33 day period until the scheduled refueling outage based on the existing margin (see calculation in Attachment 3). This will be validated by periodic monitoring.
- 11. The flaw is being monitored by visual examination, direct measurement, and leakage quantification. A " blot test" using an approved procedure is used to facilitate measurement of the crack.
- a. Plant Equipment Operators perform walkdowns each shift (two 12 hour shifts / day) which include monitoring the crack during their normal rounds.
Qualified NDE examiners currently perform daily inspection and evaluation of the crack to monitor size and leakage. The interval of the NDE examiners' inspection may be increased or decreased as deemed appropriate by the cognizant Engineering staff.
- b. Ongoing NDE " blot tests" are performed, currently on a daily frequency, to monitor crack size along with visual examination of the crack, associated leakage, and the drip bag device. As previously discussed, the " blot test" is performed by spraying dye penetrant developer over the prepared flaw surface and directly measuring liquid bleed-out. In addition to the blot testing, a second radiographic examination was performed on June 26, 1995, as a follow-up NDE examination. Other examinations will be performed as deemed necessary to aggressively monitor crack growth and leakage until the line is removed from service and repaired during the Unit 3 refueling outage scheduled to begin on July 22, 1995.
I I
RELIEF REQUEST D-2-01 San Onofre Nuclear Generating Station Unit 3, Docket 50-362 i Page 9 of 9 { l l INTERIM , ALTERNATIVE ; TO CODE REPAIR: The interim alternative to a Code repair is to leave ; as is until the Refueling Outage (scheduled to begin on July 22, 1995). This approach is supported by the ! flaw analysis discussed above and validated by the ! augmented inspection olan, also discussed above. j i h i i l l i l i l i l l l I I l l u , - , -e , _ ,-e, , --w-w4 - - - - gg---eo ,- - .--n-m y- + -- - ---- s e
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ATTACHMENT 3 RELIEF REQUEST D-2-01 OPERABILITY ASSESSMENT AND SUPPORTING CALCULATIONS
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