ML20127K565

Rev 0 to Topical Rept TR-092, Evaluation of Leak in Oyster Creek Core Spray Sys

ML20127K565
Person / Time
Site:	Oyster Creek
Issue date:	01/18/1993
From:	Abramovici J, Furia R, Jerko D GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20127J767	List: ML20127J763 ML20127K565
References
TR-092, TR-092-R00, TR-92, TR-92-R, NUDOCS 9301260158
Download: ML20127K565 (23)

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{{#Wiki_filter:............ 9 TR-092 Rev. O Page 1 EVALUATION OF LEAK IN OYSTER CREEK CORE SPRAY BYSTEM TR-092 (REV 0) BA 323602 Prepared by: J. D. Abramovici R. V. Furia D. G. Jerko J. Johnson (MPR) J. P. Logatto J. Noutell (MPR) S. Schwartz January 15, 1993 Approved by: f~ [Y 'k /- /B -93 Dire C6r - thrCTdar I alysis and Fuel Date / / / bk ') 97 Director - Erfginfeering and Design Dato GPU Nuclear 1 Upper Pond Road Parsippany, New Jersey 07054 9301260150 930119 PDR ADOCK 05000219 Y PDR

L TR-092 Rev. O Page 2 ABSTRACT During the 1992 refueling outage (14R) at. Oyster Creek, a leak was detected in the core spray annulus piping. A number of' evaluations were performed to characterize the leak and determine if' Oyster Creek can continue to operate with the leak. The following conclusions were reached: (1)' The observed leak is an isolated weld defect that has opened a small path within the pipe inside diameter and the remaining portion of the weld is intact and will-maintain the integrity of the piping. (2) The leak will not prevent the core spray system from performing its intended function under all design basis conditions and (3) the plant can-ope.vate safely for at least the next operating cycle,

.9 -TR-092 Rev. 0 Page 3 -Table of Contents Pace 1.0 Introduction 4 2.0 Core Spray System Description 4 2.1 Core Spray Spargers and Annulus' Piping 4 2.2 Core Spray System Flow Requirecents 5 3.0 Inspections and Results 5 3.1 - Inspection Techniques 5 3.2 14R Inspection Results 6 3.3 Previous Inspections and Results 6 4.0 Evaluations 6 4.1 Defect Evaluation 6 4.2 Structural Evaluation 8 4.3 Hydraulic Evaluation 12 -5. 0 Conclusions 13-6.0 References 14 APPENDIX A - Previous Surveillance Tests 16 Attachments Figure 1 - Core Spray Lower Piping Unit-18 Figure 21 - Core Spray Piping-Down Leg 19 Figure 3 - Core Spray Pipe Coupling 20 Figure 4 - Core Spray System Flow Rates 21 Figure 5 - Leak Path For Weld 9efect 22 ,7 Figure 6 - !!ydraulic Model of Core Spray Inlet Piping 23

=__ 1 TR-092 Rev. O page 4

1.0 INTRODUCTION

During the oyster Creek 14R refueling outage In-Service Inspection (ISI) of the reactor vessel internals, a leak was detected in the core spray system I piping between the vessel wall and core shroud in the annulus area (Figure 1). The leak was detected during an air test on system I when bubbles were observed ccming from a weld on the down leg portion of the annulus piping (Figure 2). A video camera was abin to determine that the leak was coming from a weld defect in a fillet weld on a pipe coupling. The identificat. ion of the leak raises the following concerns: the cause of the leak, the loss of flow to the sparger through the leak, the structural integrity of the weld with the defect, and the potential for further degradation of the coupling or weld. This report provides the technical evaluation that addresses these concerns. The evaluation addresses the structural integrity, material condition, Icahage flow and performance of the core spray system with the Icak. It is concluded that the leak will not prevent the core spray syctem from performing itn intended function under all design basis conditions and i that the plant can be operated safely with the leak in the system for at least the next operating cycle. i 2.0 CORE DPRAY SYSTEM DESCRIPTION 2.1 core spray spargers and Annulus Piping l The Oyster Creek reactor vessel contains two independent core spray sparger assemblics. Each core spray sparger assembly l consists of two 180' segments of formed 3-1/2 inch Schedule 40S stainless steel piping, each of which contains 56 spray nozzles (112 nozzles total per sparger ring assembly). Each sparger is fed from an independent penetration through the reactor vessel. The pipe is-6" schedule 40 stainless steel l from the reactor vessel nozzle in a down leg _to a 6" standard i weight "T" located next to the shroud and below the spargers. On either side of the "T" is a 6 x 5 inch eccentric reducer. Five inch schedule 40 stainlesu steel piping is then routed in either direction around the outside of the shroud for about 90* to a riser where it penetrates the shroud connecting to the two sparger segments. (Figure 1). The piping from the vessel penetration to the shroud penetration is known as the annulus piping. During the original installation, a coupling t C.

..___.______._._-_m_____...__ t n l TR-092 Rev. O Page 5 (Figure 3) was used to allow the field installation of annulus piping to the penetration connection. This was accomplished by allowing for vertical adjustments and final. fit up field fillet weld. The leak was observed from this field weld, r 2.2 Core Spray System Flow Requirements Ea'ch of Oyster Creek's two core spray systems contain a core spray sparger, two main pumps and two booster pumps. The l currer.t ECCS analysia for OC is based upon a core spray flow of 3400 gpm (1 main and 1 booster pump) from one system and-2200 gpm (1 main pump) from the other system at a vessel pressure of 110 psig. To compensate for a previously found crack in the sparger, the flow requirements for system II are 3640 gpm when the system is the two pump contributor and 2360 gpm if it is the single pump contributor. Based on core spray system tests, the actual flow rates exceed the design basis, by 500 gpm f or system I and 300' gpm for system II. The curve of fIcw versus pressure used in the ECCS analysis is shown-in figure 4. 3.0 INSPECTIONS AND TESTS 3.1 Inspection Techniques A visual inspection of the core spray system annulus piping is performed each outage utilizing a video camera. During 14R the visual examination for the core spray annulus piping was performed utilizing General Electric's (GE) " Firefly," a remotely operated vehicle. All accessible areas of the piping were inspected in accordance with article IWA-2211 (VT-1) of reference 2. When the visual examination is completed,.an air test is performed on each core spray system sparger. Because of the configuration of the core spray piping, the upper spargers (System-II) with downward pointing nozzles should fill completely with air while the lower sparger with upward pointing nozzles (System I) will only partially fill with air. The down leg between the reactor vessel penetration and the horizontal circumferential pipe run in the annulus should fill completely with air. All other piping will, at best, fill only partially with air'or just pass air bubbles along its upper conterline inside surface.

p ^ TR-092 Rev. O Page 6 3.2 14R Inspection Results During the 14R outage air test of system I, a steady stream of air bubbles was observed from the annulus piping. The leak is located in a coupling that connects two ends of the 6" schedule 40 pipe in the down leg between the reactor vessel penetration and the circumferential pipe run. A closer video examination determined that the leak was from the weld joint identified as L-3A. Additional visual

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aminations were performed to better locate and characterice the leakage.

~ This additional inspection confirmed that the indication in the L-3 A fillet weld was circular and approximately one-eighth of an inch in diameter. Wold L-3A was 100% VT-1 inspected per ref erence 2 and was determined to be acceptable except for the 1/8" diameter round weld defect believed to be from the 3 original construction which is now 1 caking. A uimilar inspection was also performed on the system II annulus piping. No bubbles were observed from the annulus region. However, in the same wold (U-3A) as the system I weld (L-3A) and at the same relative location of the piping to the vessel wall as the defect in weld L-3A, two linear indication were detected. They were characterized as splits or tears between wold passes that appear to be construction related. 3.3 Previous Inspection Results The annulus piping has been inspected in each of the oyster Creek outages since 1978. No relevant indications have been found. Air tests were perf ormed in each of the outages except for the 11R outage and no leakage was observed from the annulus region. Appendix A provides a summary of the inspections and results since 1978. 4.0 EVALUATIONS 4.1 Defect Evaluation The visual inspection of the defect in weld L-3A shows it to be a rounded hole in the 1/4 inch fillet weld in the core spray annulus pipe coupling. The weld is a field weld and the def ect appears in an area of limited access where the coupling is closest to the reactor vessel wall. The air test showed that the defect bubbled at a relatively high rate, consistent with a fairly open leakage path completely through the weld such as might occur if a slag inclusion was cleared out to create a leak path. The leak path is shown in Figure 5. I

-_ = TR-092 Rev. O Page 7 The characteristics of the defect are consistent with a weld defect in a portion of the weld that was difficult to access. The defect likely contained slag and/or lack of fusion and opened up af ter approximately twenty years service with normal cyclic stresses. This is the most probable cause for the Icak. other postulated causes resulting in a through wall defect are fatigue or intergranular stress corrosion cracking (IGSCC) originating on the pipe or weld inside surface and joining with the visible defect in the weld. liowever, these latter mechanisms are considered to be highly unlikely based on the following observations. l The high bubbling rate is consistent with a rounded hole with significant flow area. Air leakage through a tight fatigue or IGSCC crack in part of the weld or pipe wall would be expected to be much less than observed. Cyclic service loads are small and would not be auf ficient to cause a f atigue crack to initiate. Fatigue crack growth of a small initial crack is predicted to be negligible (see Section 2.3.2). l llo ICSCC has over been observed in-any of the Oyster Creek annulus piping. 11 0 confirmed IGSCC has been observed in any part of the core spray system since 1978. If IGSCC were to appear, it would be expected to occur in L the weld heat affected zone (llAZ) adjacent to the weld rather than in the weld itself. Weld metal is known to l be more resistant to IGSCC than base metal. There is no evidence of IGSCC-in the weld liAZ. If ICSCC has occurred, it had to originate in the pipe inside surface in a crevice area at the root of the socket weld. Such crevice-assisted IGSCC requires the l presence of oxygenated water in the immediate vicinity of the crevice. 1:owever, the pipe internal environment near the coupling was stagnant before the leak occurred and contained little or no oxygen. The nearest source-of oxygen was from the core spray sparger approximately twenty feet (48 pipe diameters) away. Based on the above observations, it is concluded that the defect is a weld slag inclusion that opened up after the slag dissolved or after normal fatigue cycles caused the slag to loosen. While it is possible that -crevice assisted --IGSCC . - -.,. - -... - - - - - -., -. - -.. - - - - - - _ ~. - _. -_ - - - _.-

TR-092 Rev. O page 8 i originating on the inside surface has occurred, this is considered highly unlikely. Since it is not possible to prove the absence of IGSCC, an ICSCC crack was postulated to exist. Crack growth was considnred on a worst case basis to determine the arowth during one operating cycle (approximately two years). Crack growth rates f or IGSCC were taken f rom ITUREG 0313 and applied i to a postulate.1 0.50 inch through wall crack in the coupling with an assumed yield level stress field in the weld heat affected zone. Under these conditions the circumferential growth was calculated to be approximately 3.0 inches in a two year period. This crack length is structurally acceptable since the 1/4" field weld with this defect is still stronger than the 3/16 inch shop weld in the coupling. Further, the leak rate through the calculated 3.0 inch IGSCC flaw is less a than 1 gpm, far below existing flow margins in the systems. The indications on weld U-3A have been characterized as splits or tears between veld passes. The splits or tears have the 1 same reddish color indicating that those are not new. If they were new indications, a chiny metal or reflection would 1 probably be evident. The indications on weld U-3A are in the same location as the defect on weld L-3A facing the reactor vessel wall where the welding would be difficult. ICSCC, as stated above, is not expected to occur in the veld material. The conclusion in that these indications are construction defects and the weld is acceptable. i 4.2 Structural Evaluation The overriding f actor in the structural evaluation of the core spray annulus piping is thst the axial load capacity of the coupling is controlled by a 3/16 inch shop weld on the lower coupling sleeve (Figure 3). Thus, the 3/16 inch weld controls the axial strength of the coupling for all normal and postulated accident loads. Because of the weld size and diameter differences, the 1/4 inch weld with the leak could-lose approximately 31 percent of its circumference (e.g., about 7 inches) and not reduce the axial _ capacity of the coup 1ing. 6 The above not withstandi'ig, finite element stress analyses l were performed to determine the stresses 'in tho' 1/4 inch i fillet weld with the leak during normal operating and_ core spray injection conditions. The core spray annulus piping was modeled from the reactor vessel core spray nozzle where the thermal sleeve is rolled into the nozzle safe end to the upper 1

TR-092 Rev. O Page 9 core shroud where the 5 inch pipe is wolded to the shroud (Figure 1). The annulus piping was assumed to be built-in at the core spray nozzle and upper core shroud and supported in the vertical direction only at the pipe supports attached to the outside diameter of the shroud. The loads consisted of: 1. Deadweight of the pipe (including water inside the pipe). 2. Drag loads due to recirculation flow (19,000 lb/s) in the annulus between the reactor vessel and the upper core shroud. 3. Thermal loads due to differential thermal expansion between the annulus piping attachment - points on the reactor vessel and upper core shroud, and thermal growth of the inlet piping. During normal operation the reactor

vessel, upper core shroud, and annulus piping were assumed to be at 550*F.

During core spray injection the reactor vessel and upper core shroud were assumed to be-at 550'F while the core spray annulus piping was assumed to be at 200*F due to cooling from the incoming core spray flow. (Note: These temperatures correspond to the approximate conditions early in the core spray injection transient. Later in the injection transient the temperature conditions - and resulting stresses are less severe.) 4. Pressure load during core spray. injection. The pressure load is the dif ference between annulus pipe pressure and the reactor pressure and was assumed to be 35 psi-based )n hydraulic analyses described in Section 4.3 below. i 5. Seismic load. The analyses were based on a censervative static seismic load of-Sg assumed to act simultaneously in the three orthogonal directions. The following load combinations were evaluated. For normal operation: 1. Deadweight + Drag Load 41 Normal Thermal Load 2. Deadweight + Drag Load + Normal Thermal Load + Seismic L

4 TR-092 Rev. O Page 10 For core spray injection: 3. Deadweight + Accident Thermal Load + Pressure 4. Deadweight + Accident Thermal Load + Pressure + Seismic Results of the finite element stress analyses of the core spray annulus piping are tabulated in Table 1. In this table, the calculated shear stress in the 1/4 inch fillet weld at the field coupling is tabulated for each of the above loading combinations. These stresses are based on a full 360' weld. The reduction in static strength of the veld due to the observed defect (1/8 inch hole) is considered to be negligible. Table 1 Calculated Shear Stress in 1/4 Inch Fillet Wald at Field Coupling Load Combination Shear Stress (ksi) Rofinal operation 1. D+Fo+T 0.8 o 1 2. D+Fo+To+ E 4.1 C_ ore Spray Injectign 3. D+T+P 3.2 4. D + T, + P_+ E 5.8 As shown in Table 1, the calculated shear stresses _-in the 1/4 inch fillet weld during normal operation and core spray injection are well below typical ASME Code allowables -for shear in a fillet weld (19 ksi at 550'F). Therefore, the l System I core spray annulus piping _ is. considered to' be structurally adequate in its present condition. Further', at-the calculated stress levels during-normal operation, fatigue 1 initiation and growth would be negligible. 7 Finite _ element stress analyses were also performed-to _ determine the stresses in the 1/4 inch fillet weld during the blowdown portion of a large break LOCA, but prior to onset'of (k

TR-092 Rev. O Page 11 core spray flow. The highest fluid velocities and drag forces on the core spray annulus piping would be due to a complete instantaneous circumferential break of a recirculation line. The drag force on the core spray annulus piping during a large break LOCA was conservatively calculated for a recirculation 4 line break with a maximum flow rate of 45,000 lb/s. The drag fdrce was calculated to vary from approximately 32 lb/ft for naturated water conditions to 655 lb/ft for saturated steam. (!!ote: The drag force during normal operation is approximately 13 lb/ft.) The actual quality of the fluid in the annulus varies during the blowdown. A drag load of 655 lb/ft corresponding to saturated steam conditions was used in the bounding calculations. The loading combination evaluated was: 5. Deadweight + LOCA Drag Load + tlormal Thermal Load For the above loading combination, the shear stress in the 1/4 inch fillet weld was calculated to be 2.2 ksi assuming a full 360' weld. The required length of weld to _ withstand the applied loads at ultimate stress levels was calculated to be 8.5 inches (136' of the pipe circumference). In addition to the stress analyses described above for LOCA blowdown loads, a hypothetical case was calculated in which the 1/4 inch fillet weld was assumed to have totally failed. Finite element analyses were performed to determine the relative displacement of the_ 6-inch pipes at the field coupling assuming the 1/4 inch fillet weld was cracked through wall 360* around its circumference. The loading was the same as described above, except the model of the inlet piping was decoupled at the field coupling. For this hypothetical

case, the lower pipe segment _ was calculated to displace downward 0.5 inches relative to-the upper segment.

The coupling was designed to have a nominal engagement of 1.0 inch. Based on the General Electric installation drawing and measurements from field video-tapes, the actual engagement is estimated to be about 0.9 inches. Since _the drag load assumed in these calculations is L considered very conservative, the coupling is not_ considered i likely to separate during the blowdown even if the weld were L to fail completely. These analyses demonstrate that the core spray system would be functional even if a substantial portion of the 1/4 inch i fillet weld were to become cracked, although we believe this l l

TR-092 Rev. O Page 12 in highly unlikely. 4.3 !!ydraulic Evaluation !!ydraulic analyses A were performed to determine the ef fect 8 of the observed defect in the 1/4 inch fillet veld on core spray system performance. The design basis flow rate for core spray System I (with the leak in the field coupling) is 3400 a gpm at a reactor pressure of 110 psig. Based on core spray system tests, the actual flow rates exceed the design basis flow rates by about 500 gpm for System I. A hydraulic resistance model of the core spray annulus piping, and flow nozzles was developed (Figure 6), i sparger,. calculations were performed at the design basis flow rate of 3400 gpm with an exit pressure of 110 psig for various cases as summarized below. 1. Case 1. No leak at the 1/4 inch fillet weld. Base case. 2. Case 2. A 0.125" diameter hole at the 1/4 inch fillet weld. This case represents the observed condition. 3. Case 3. A.3.0" x 0.0009" linear crack at the 1/4 inch fillet weld. This case represents an upper-limit IGSCC flaw and includes the flow from the crack and the observed defect. 4. Case 4. A 0.015" gap 360*-around the pipe at the field-coupling. This case represents complete failure of the. 1/4 inch fillet wold. Results of the calculations are summarized in Table 2. In this table, the calculated flow loss at the assumed defect-and the minimum nozzle flow rate (as a percentage of the nozzle flow rate for the base case) are tabulated. j L )

TR-092 Rev. O page 13 Table 2 Results of Hydraulic calculations Core Spray System I Flow Loss Hin Nozzle case Through Flow Defect (% of Base (gpm) Case) 1. Daso Case 0 100.00 2. Observed Defect 0.4-2.7* 99.92 3. IGSCC Crack plus Defect 0.6-2.9* 99.91 4. Complete Failure of Wold 46 98.64 Range of flows is due to assumed resistance of the crevices behind the Veld. The higher value assumes no resistance from the crevice. As shown in Table 2, the calculated flow losses are small compared to the available margin in core spray System I (500 gpm). Therefore, the flow loss through the observed defect in the 1/4 inch fillet weld, as well as worst case scenarios, is considered acceptable.

5.0 CONCLUSION

The observed leak is-considered to be an isolated-weld defect that, as a result of slag dissolution or years of cyclic stresses, has opened a small path with the pipe 1D through-gaps in the pipe coupling. The leakage flow is negligible and will not af fect the performance of the core spray system. The remaining portion of weld L-3A is-intact and. insures more-than suf ficient strength to maintain the integrity of the coupling for stresses under all design basis conditions. There is no evidence of IGSCC or on-going corrosion in the-leaking weld and there is-no history of IGSCC or crevice corrosion in other fillet welds that have been inspected at Oyster Creek exposed to similar environment and operating pressure.- Even in-the-unlikely event that IGSCC does occur,- there would' be no significant degradation in core spray system performance for at least the next operating cycle. The on-going In-Service Inspections will insure that the coupling will continue to be closely monitored to detect a change in the leak rate or new indications in the coupling. 'ga.*p.+ ,.#y ,y y y y7.. g ,m-q 9w -p 9 e9 -.. w---- in-s p -w-fi.,e g w -e.. 4 e 4 e

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6.0 REFERENCES

1. P. Wei and li. 11. Paustian, " Oyster Creek Nuclear Generating Station SAFER /CORECOOL/GESTR-LOCA Loss-of-Coolant Accident Analysis," NEDC-31462P dated August 1987 s 2. ASME Boiler and Pressure Vessels, Section XI, 1986 edition 3. GPU Nuclear Topical Report, TR-093, " Oyster Creek Core Spray System Inspection Program 14R Outage, " dated January 8, 1993. 4. Oyster Creek Updated FSAR, Figure 6.2-11, Update 7.12/92 5. MPR Calculation No. 83-176-Id1C-03 Rev. O, " Annulus Piping Wold Crack Growth and Leakage Aren," dated January 14,-1993. 6. MPR Calculation No. 83-176-RBK-01 Rev. O, " Determination of Loads on Core Spray Annulus Piping Field Connection," dated January 14, 1993. 7. MPR Calculation No. 83-176-RBK-02 Rev. O, " Determination of

Loads, Stresses and Deflections During Recirculation Line Break,"

dated January 14, 1993. 8. MPR Calculation No. 83-176-DRM-01 Rev. O, "liydraulic Resistance of Core Spray Piping Between Reactor Vessel and Lower Sparger Outlet," dated January 14, 1993. i 9. MPR Calculation No. 83-176-DRM-02 Rev. O, " Determination of Leak Rate Through Crack at Field Connectio!. in Core Spray Piping," dated January 14, 1993. 10. GPU Nuclear Topical Report, TR-080, " Oyster Clock Core Spray System Inspection Program - 13R Outage", dated April 8, 1991. 11. GPU Nuclear Topical Report, TR-054 (Rev. 1), " Oyster Creek Core Spray System Inspection Program 12R Outage", dated January 17, 1989. 12. GPU Nuclear Topical-Report,-TR-037, " Oyster Creek Core Spray System Inspection Program", dated August 22,: 1986. 13. GPU Nuclear Topical Report, TR-013, " Oyster Creek Core Spray System Inspection Program, Response to NRC I&E Bulletin No. 80-13", dated April 25, 1983. ,-,,,..-,v---.,---.-- =,,.-- n. .J. . - a,6 - +.,. - -. - - -. -.. + - - -.. - -. - - - - -. ~ ~

__.=..m______..____ 6 4 TR-092 Rev.:0 Page 15 l 14. Jeracy Central Power & Light company, Repair Proposal llo. 475-01, OCllGS Core Spray System Sparger Ropair, dated March 31, 1980. 15. Jersey Central Power & Light Corapany, Repair Proposal lio. 320-78-1, O C!f G S, Core Spray Sparger 2, dated llovember 15, 1978. 4 1 5 t i 1 l l' I l l t L

TR-092 Rev. O Page 16 APPENDIX A Previous Survoillance Tests (since 1978) 10 ng ou1Ago (11U1 visual inspection of the core spray system annulus piping was performed utilizing a video camera with underwater auxiliary lighting. Due to access restrictions, n hand held camera technique was used to perform the examination. All accessible areas of the ~ piping were inspected. No relevant indications were noted during the inspection or subsequent review of video tapes. Observation of the core spray systems during the air tests was performed by utilizing a hand held video camera with auxiliary underwater lighting. No alt bubbles were observed coming from the annulus piping. 11 HR_RtLta.ge L1_9_8 8 - 11011 visual inspection of the core spray system annulus piping was performed utilizing a hand held video camera with auxiliary lighting. All accessible areas of the piping were inspected. No relevant indications were noted during the inspection or subsequent review of the video tapes. Observation of the core spray system during the air test was ,j performed by utilizing a hand held video camera with auxiliary lights. No air bubbles were observed coming from the annulus pipinij. 12 1 1 R C*t t a_g_e (1986) The core spray annulus piping was inspected utilizing a hand held camera technique. All accessible areas of the piping were inspected. The visual examinations were performed by vendor personnel with the results evaluated by a certified Visual Level III. An independent overview of the results was performed by a independent certified Visual Level III. No indications that could be interpreted as crack-like were noted during the examination or post examination review. No air test was performed on the annulus piping. l3 10R Outaae (1983-1984) The core spray annulus piping was visually inspected and no recordable indications were identified.

2 TR-092 Rev. O page 17 A limited visual air test of the System II vertical annulus piping was performed. No indications were noted in the System II annulus piping. 1 An ultrasonic examination was attempted on selected welds (fourteen) on che six and five inch diameter annulus piping. Due to access restrictions, seven welds were only partially examined. These w' elds included U7 (263'), L7 (305'), L8 (293'), L9 (301'), U8 (336'), U17 (295'), and U18 (286.5'). No recordable indications were identified in any of the seven welds inspected. Following the completion of the 10R inspections, a review was-conducted of the results along with the 1978/1980 results, and selected video enhancements performed in 1982. This review. concluded that the annulus piping did not contain any indications and no cracks of structural significance existed. Further, the piping was otructurally capable of meeting its design function. I4 PR Outage L1 M Visual inspecLions were performed and video tapes made-of the-core spray piping within the reactor vessel between the inlet nozzle and the vesse] shroud. These tapes were reviewed by two qualified visual inspectors and two indications were classified as possible cracks. Both of these indications were on the 6" x 5" eccentric reducers of the System II piping. The larger of these two indications was classified by a third qualified inspector as marks mado during installation. Review of the 1978 tapes did not provide any' additional information since this inspection concentrated on the piping welds. The indications were later dispositioned as non-cracks (see 10R). l5 8.R Outage {.19781 A scheduled in-service inspection was performed of the reactor j internals. This inspection included a visual inspection of.the l accessible portions of the 5" inlet piping between the reactor i vessel core spray nozzles and the OD of the shroud. No visual indications were found on the annulus. piping. l f r .. ~. ~

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