ML18093A474

Forwards Util Technical Dept Engineering Evaluation Re Thermocouple Column Leak Discovered on 871009.Evaluation Describes Affected Areas,Cleanup & Corrective Actions. Support Analysis & Documentation Included

ML18093A474
Person / Time
Site:	Salem
Issue date:	10/26/1987
From:	Miltenberger S Public Service Enterprise Group
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
NLR-N87202, NUDOCS 8710300071
Download: ML18093A474 (88)

Text

{{#Wiki_filter:-1 __ -_-_------=---=------_-_-__ -_-_-_-----~~~ -- -- - i' Public Service Electric and Gas Company Steven E. Miltenberger Vice President - Publlc Service Electric and Gas Company P.O. Box 236, Hancocks Bridge, NJ 08038 609 339-4199 Nuclear Operations October 26, 1987 NLR-N87202 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen: THERMOCOUPLE:COLUMN'LEAK SALEM GENERATING STATION UNIT NO. 2 DOCKET NO. 50-311 Attached is the Public Service Electric and Gas Company (PSE&G), Salem Technical Department Engineering evaluation concerning the Unit 2 Thermocouple Column Leak which was discovered on August 9, 1987. The evaluation describes the affected areas, the clean up and the corrective actions performed to address the issue. The support analysis and documentation is included. If you have any questions concerning this evaluation, please contact us. 8710300071 871026 PDR ADOCK 05000311 P P~R Attachment C Mr. D. c. Fischer Sincerely, USNRC Licensing Project Manager Mr. T. J. Kenny USNRC Senior Resident Inspector Mr. w. T. Russell, Administrator USNRC Region I Mr. D. M. Scott, Chief Bureau of Nuclear Engineering Department of Environmental Protection 380 Scotch Road Trenton, NJ 08628 \\ I~

UTIL TECHNICAL DEPARTMENT ENGINEERING EVALUTION CONCERNING THERMOCOUPLE COLUMN LEAK DISCOVERED ON 10/09/87 >?*,' NOTICE - THE ATTACHED FILES ARE OFFICIAL RECORDS OF THE DIVISION OF DOCUMENT CONTROL. THEY HAVE BEEN CHARGED TO YOU FOR A LIMITED TIME PERIOD AND MUST BE RETURNED TO THE RECORDS FACILITY BRANCH 016. PLEASE DO NOT SEND DOCUMENTS CHARGED OUT THROUGH THE MAIL. REMOVAL OF ANY PAGE(S) FROM DOCUMENT FOR REPRODUCTION MUST BE REFERRED TO'FILE PERSONNEL. DEADLINE RETURN DATE RECORDS FACILITY BRANCH

1.0 Executive Summary 2.0 Description of Thermocouple Instrumentation Port Column Assembly 3.0 Event Identification 3.1 Containment Airborne Monitor Levels 3.2 Summary of Airborne Monitor Levels 3.3 Reactor Coolant System Leak Rate 4.0 Affected Areas 4.1 Lower Canopy Seal 4.2 Reactor Vessel Head 4.3 CRDM Counterbore in Vessel Head 4.4 Leaking Thermocouple Column Counter Bore Wastage 4.5 Reactor Vessel and Reactor Head Flange. 4.6 Reactor Vessel Studs 4.7 CRDM Ventilation Fan Shroud and Duct Work 4.8 Mirror Insulation 4.9 CRDM Columns and Non-Affect Thermocouple Columns 5.0 Clean-Up Operations 5.1 General boric acid clean-up and acceptance criteria 6.0 Westinghouse's Calculations and Evaluations 7.0 Weld Repair 8.0 Final Inspection 9.0 Potential Failure Mechanisms 10.0 Boric Acid Corrosion Information Survey 11.0 Future,Actions 12.0 Safety Evaluation 13.0 Event Chronology 14.0 Ref*e:z::ences 15.0 Attachments RE~UL~JORY DOCKET f!LE COPY

1.0 EXECUTIVE

SUMMARY

On August 7, 1987, No. 2 Unit was removed from service to replace one of three, single phase, main power transformers. Station Management was concerned over the fact that Noble Gas Radiation Monitor Channel 2Rl2A Count Rate had doubled (4K to 9Kcpm) in May, 1987, and that a small, persistent leak apparently existed in Containment from the Reactor Coolant System (RCS). It was decided to form a team of Station System Engineers to conduct a detailed area by area inspection of the reactor coolant system in Containment. During this inspection on August 8, 1987, boric acid residue was observed on the reactor vessel head control rod drive mechanism (CROM) ventilation shrouding. At first it was suspected that the boric acid might be from a flange leak on the reactor vessel.level system (RVLIS). On August 9, 1987 a Station Planner entered the Containment to scope out the initial inspection and repair. On August 10, 1987 a System Engineer and Maintenance personnel entered Containment to remove a section of the CROM ventilation shroud to inspect the area. At this time it was discovered the leak was from the No. 5 thermocouple instrument column. Three (3) pin hole leaks were discovered spraying outwards from the lower canopy seal weld. Removal of the reactor vessel head mirror insulation revealed boric acid build up around studs No. 22 through 26. Removal of the boric acid revealed eleven (11) depressions on the reactor vessel head due to apparent corrosion. Boric Acid was evident on the side of the vessel head down to the flange gap area. Westinghouse was contacted for support. Removal of the CRDM ventilation shrouding began to better inspect the thermocouple column. On 8/12/87 a project team was formed under the direction of the 5tation Technical Manager. Westinghouse (consultant) proposed welding a split canopy around the leaking seal weld. The split canopy is designed ASME Code Section III, Code Class 1 and is considered the new seal pressure boundary for the RCS. Clean-up operations began and the Station prepared to remove Reactor Vessel studs for inspection and eval-uation of the 11 depressions found beneath the boric acid pile.

2.0 In summary, 17 Reactor Vessel studs were removed, inspected, cleaned and reinstalled 16 in the vicinity, of the boric acid build up and one 180° from the boric acid deposits as an added precaution. The depressions on the head were cleaned followed by ultrasonic tests and magnetic particle tests. The split canopy weld repair was c,ompleted and the boric acid clean up was accepted. The following se~tions will disc~ss in detail the resolution of each item. Deficiency Reports (DR) were used to document resolutions. Safety Eval-uations included in the DR disposition assure there are no USQs involved with the "repair" or "use as is" disposition. Sections of this report involving deficiency resolution reference the Deficiency Report number next to the Section title. It was PSE&G's conclusion that the issues associated with this event had been resolved and did not prevent return to normal operations. DESCRIPTION OF THERMOCOUPLE INSTRUMENTATION PORT COLUMN ASSEMBLY Salem Unit 2 is equipped with incore thermocouples which measure fuel assembly coolant outlet temperatures at preselected core locations. The thermocouples penetrate the Reactor Vessel closure head by utilizing five instrumentation port column assemblies. A total of 65 thermocouples pass thru these five columns. These instrumentation port column assemblies are pressure-retaining assemblies which can be disassembled to permit removal of the reactor head from the reactor vessel, for activities such as refueling. (See Attachment 1)

3.0. EVENT IDENTIFICATION 3.1 Containment Airborne Monitor Levels

• containment airborne levels for iodine and noble gas increased on February 7, 1987.

Particulates had already been at a high level. This increase was due to increased fuel clad leakage as well as a suspected small RCS leak into the Containment which was below the lgprn unidentified Technical Specification limit. Through March, April and May the levels remained consistent with reactor power changes (i.e., increases or decreases in monitor levels are attributed to power swings and/or operation of the fan coil units). In May a period of steady power resulted in a slow increase of iodine and noble gases. The 2R11A particulate reading remained steady at approximately 30,000 to 40*,ooo cprn. The period of March through May indicates that the small leak persisted without increase. On May 28th, an increase in the noble gas channel (2R12A) occurred roughly doubling the reading (4,000 to 9,000 cpm). At this time, the original RCS leak could have increased or a new leak could have started. A power decrease from May 29th to May 31st masked the particulate reading causing it to decrease. However on return to power on June 1st, the particulate monitor 2R11A increased to a somewhat higher value supporting the conclusion of a leakage increase on May 28th. This trend continued with the particulate monitor reaching just over 70,000 cpm on June 20th. This appears to be the result of increasing Unit 2 to 100% power earlier that same day. On July 12th when the unit returned to service from a turbine repair outage, a series of power swings between 100% to 60% were experienced which resulted in a slow increase of noble gases and iodine levels including return of particulates to the 70,000 cprn range observed in June. 3.2

SUMMARY

OF CONTAINMENT AIRBORNE MONITOR LEVELS

1.

There apparently has been a persistent leak in containment since before February, 1987 as in-dicated by particulate monitor trends 2Rl1A. This leak did not exceed any unidentified RCS leak rate calculations that are performed once every 24 hours.

2.

The increased noble gas monitor and iodine monitor levels commencing February 6th were due to the Unit 2 fuel clad leak increase which has been previously documented. It therefore is unlikely that the thermocouple column leak occurred at this time.

3.

It cannot be clearly identified from this data when the leak in the reactor thermocouple column commenced.

4.

On May 28th, it appears there was an increase in the leak or a new leak developed. NOTE: During the forced outage of July 3 thru July 12 for turbine repairs, leaks were identified in the containment and repaired. When the unit was returned to power (July 11) there was a containment inspection performed behind the Bio-Shield. (This inspection is required when the Reactor Coolant System is at 1000 psi and is documented in IOPLEAKS-2). This inspection did not identify any significant leaks.

5.

The airborne radioactivity plots during the month of July did not indicate any increase in the RCS leak rate or any new leaks in the containment. 3.3 REACTOR COOLANT SYSTEM LEAK RATE No significant changes in RCS leak rate were observed since the beginning of July. Identified, unidentified, and containment sump pump runs were graphed and show no large increase. The highest leak rates during this period were: Containment Sump - .47 gpm (8/1/87) Unidentified Leakage - .29 gpm (7/8/87) Identified - .59 gprn (8/6/87) At no time did Salem Unit 2 operate in excess of the Technical Specification limit of 1 gpm unidentified leakage. (See Attachment 2) l 4.0 AFFECTED AREAS 4.1 Lower Canopy Seal Weld The lower canopy seal weld on thermocouple column

1. 5 had three pin-hole leaks.

Two were within 1/4 inch apart and the third about 1 1/2 inches away. All three leaks were on the outside circumference of the column i.e. blowing away from the reactor head.

4.2 The original procedure for making this weld utilized an automatic gas shielded tungsten arc process in which the torch moves around the fixed position weld joint by means of a motor driven carriage seal welds. These welds were performed in the field. (See Attachment 2A). REACTOR VESSEL HEAD (DR-SSP-87-107) Boric acid solution from the 3 pin hole leaks impinged on the CRDM ventilation shroud and dripped onto the vessel head. Visual examination after the removal of the mirrior insulation showed an accumulation of boric acid on the top of the closure head between studs #22 through #26. The closure head was found to have 11 depressions caused by the boric acid. These depressions were mapped. A full scale tracing is shown in. The dark solid area represents the shape of the deepest portion. The papers should be placed edge to edge and the tracings will follow the contour of the vessel head. One hypothesis on how the depressions were formed is the instrument port leak spray was cooled by the CRDM cooling air and ran down the inside diameter of the vent shroud support ring. The liquid dripped onto the closure head, evaporated leaving boric acid crystals, and was rewetted by the next drop of liquid off the shroud ring causing the boric acid to become more concentrated and attack the head in the pitted area. The boric acid then ran down the head causing the erosion grooves to develop before it evaporated. After a while the boric acid built up on the side of the head -where it dried with the continuing boric acid from the leak drying on top of the initial boric acid stream. Below is an Engineering Evaluation summary preformed by Westinghouse on these depressions (MEB-PCE-5142). Following cleaning of the reactor head dome area, the various depressions resulting from boric acid wastage were visually inspected, measured and mapped. (See Attachment 3) The maximum depth of the depression was determined to be 15/32 inch. Additionally, the closure dome (head) was measured ultrasonically and determined to be 7.6 inches.

An engineering evaluation of the reported wastage areas was then performed using the current ASME code Section III head sizing and nozzle reinforcement criteria. The calculation (No. RMB-870815) assumed that the head dome adjacent to the wastage has a thickness of 7.00 inches which is the drawing minimum thickness. A general loss of material around the affected head penetration (Penetration No. 76) resulting in a 1/2 inch reduction in thickness was then assumed. The calculation determined that head thickness could be reduced to 6.5 inches and the ASME Section III requirements would still be met. In this manner both the 15/32 inch deep depression and an enlarged crevice around penetration no. 76 are shown to be acceptable. An additional calculation, included in calculation RMB-870815, was performed to evaluate the abnormal cooling of the affected area of the closure head dome during the period that the canopy seal was leaking. This calculation assumed that the outside surf ace of the head dome and head adapter was cooled to 212 (degrees f) by the condensate in the wet boric acid on the outer surf ace continually boiling away. The inside surface of the dome and head adapter were assumed to be operating at the vessel outlet temperature.of 607.4 (degrees f), and linear through-wall thermal gradients were conservatively assumed in both parts. The thermal stresses which were calculated assuming this extreme linear thermal gradient resulted in maximum ranges of primary plus secondary stress intensity which satisfied the ASME Section III Code (3SM) limit for the head dome, but exceeded (3SM) for the head adapter tube section. The maximum range of stress intensity in the dome was calculated to be 78.4 KSI compared to a 3SM limit

• of 80.1 KSI.

The maximum range in the head adapter was found to be 71.5 KSI compared to an allowable of 69.9 KSI. Since the 3SM limit was exceeded and the major part of the stresses was thermal bending a simplified elastic-plastic analysis was performed on the adapter tube in accordance with paragraph NB-3228.3 of Sedtion III. The simplified elastic-plastic analysis requires performance of a fatigue analysis with the peak stress ranges multiplied by a (KE) factor greater than or equal

to 1.0 depending on how much the 3SM limit was exceeded. Exceeding the 3SM limit is acceptable if the calculated over-all usage factor remains less than 1.0. The cyclic operation of the plant during the period of the canopy seal leak was determined by checking the plant operating records. Based on these records the adapter tube and the head dome were conservatively assumed to have seen twenty--heat-up/cool-down cycles and thirty-six plant loading/unloading cycles during the time frame from refueling in December, 1986 until shut-down in August 1987. These operating cycles augmented by the effect of the local cold-spot on the head were then considered in fatigue analysis of the dome and the adapter. The fatigue analysis of the head dome resulted in a usage factor of 0.012. Since the original design usage factor for the dome for the life of the plant is zero, the additional 0.012 usage factor due to the cold-spot is insignificant when compared to the remaining margin to the 1.0 design basis limit. A UT and MT was also performed on the affected areas of the vessel head. The UT shows the head was 7.6" thick in the sound metal area around the pits. The 7.4" thickness is near the base of the leaking penetration and probably represents the effect of the erosion wash we observed at the base of that penetration. The MT showed no recordable indications in the area of the pitting condition on the reactor vessel closure head. See Attachment #4 for Examination Records. 4.3 CRDM COUNTERBORE IN VESSEL HEAD After the vessel head was cleaned and vacuumed the CRDM counterbore crevices were inspected. Though some boric acid was found in the CRDM crevices in the area of the leaking thermocouple column, no vessel head wastage was observed. All crevices that had boric acid deposits were verified clean before shrouding reassembly.

4.4 LEAKING INSTRUMENT PORT COLUMN BORE WASTAGE (DR-SSP-87-107) 4.5 4.6 The leaking thermocouple column (#5) suffered wastage in its counterbore region of the reactor vessel head (See attachment #5). The same transients as applied to the dome analysis were applied to the instrument port counterbore evaluation. The simplified elastic plastic/fatigue analysis of the head adaptor tube section resulted in a usage factor of 0.002 inch which is also insignificant. REACTOR VESSEL AND REACTOR HEAD FLANGES Boric Acid ran down the side of the Reactor Head around studs 23, 24 and 25. The boric acid solution then ran into the gap between the vessel and reactor head flanges (see attachment #6). After the initial cleaning, a fiberscope wa~ used to inspect the gap, back to the stainless steel cladding. This* examination showed that boric acid had ran from stud #18 to #30, with the highest concentration around stud #24. Flange areas were cleaned again as each stud was removed. A tool was fabricated for this cleaning effort. A piece of 1/8" sheet metal was cut in a "J" shape. Cotton cloth was wrapped around it and it was inserted into the gap and dragged along the surface. Inspection of the flange areas with the fiberscope proved that this cleaning method 'removed all detectable boric acid. REACTOR VESSEL STUDS (DR-SSP-87-313) Studs numbered 17 thru 31, 34 and 54 were removed by procedure for examination of boric acid residue and/or damage. Only one stud was removed at any time per the Westinghouse recommendation. As each stud was removed, it was cleaned and neolubed. The stud hole and flange area was cleaned and verified to be free of boric acid. Photographs of each stud were taken before and after cleaning operations. Readings of elongation on 5 studs, with 2 studs on each side of stud being removed were taken prior to stud removal. After the reinstallation and tensioning the stud, elongation was verified as before on the 5 studs under consideration before next stud removal was initiated. Since initial conditions were restored every time, multiple stud removal, one after another, had no impact on the bolted joint in-tegrity.

4.7 Closure studs numbers 18 through 30 appear to have been affected by the boric acid deposition based on visual and fiberoptic inspections through the gap between the vessel and the closure head flanges. Studs numbers 23 through 27 were both visually and Magnetic Particle Test inspected. No significant corrosion, wastage damage was found on any of the studs and the MT inspections revealed no reportable indications. In the event that boric acid might remain on some of the studs when the plant returns to operations, calculations were performed to evaluate the effect of dry boric acid on a stud for six months of continued operations. Calculation (SLA-081887) concluded that a uniform reduction in stud shank diameter.of 0.032 inch could be tolerated and the ASME Section III sizing requirement would still be met. Since the maximum expected corrosion rate due to dry boric acid on carbon steel at temperatures around 500 (degrees-f) is only 0.020 inch per year, the maximum reduction in diameter for a six month period would be 0.020 inch which is less than the 0.032 inch. Therefore, the maximum postulated wastage of the stud shank area for the six month period would not affect the structural integrity of the stud. Calculation (SLA-081987) assumed that an inch of stud thread engagement in the vessel flange was lost due to boric acid corrosion. The calculation determined that there is 1.19 inch of excess thread engagement in the vessel flange during the application of the maximum stud tensile load during normal operation (end-of-heat-up). It is therefore concluded that boric acid corrosion.of the stud and flange threads for six months of continued operation would not affect the structural integrity of the vessel and closure studs. CRDM VENT FAN SHROUD AND DUCT WORK Boric acid solution from the 3 pin hole leaks sprayed on the CRDM shrouding. Only one section of the shroud was affected. This section was removed and cleaned of all boric acid. N6 evidence of any wastage was found.

No. 23 CRDM ventilation duct work which connects to this piece of shroud also had boric acid deposits on the inside of a 90° elbow. This deposit extended approximately three feet into the duct work. Again, the elbow was cleaned and had no evidence of wastage from boric acid. 4.8 MIRROR INSULATION (9DR-SSP-87-308) Five panels of mirror insulation were removed and inspected. Inspection results showed that only three panels had boric acid deposited on them, with the worst one being the one by stud #24. This thin film of boric acid was removed and it was determined that there would be no future concerns by reinstalling this insulation. 4.9 CROM ADAPTERS AND NON-AFFECTED THERMOCOUPLE COLUMNS (DR-SSP-87.314) A number CROM adapters around the leaking thermocouple column had small amounts of boric acid sprayed on them. All boric acid traces were removed using scotch brite and damp rags with demineralized water. A 500V megger and resistance checks were performed on 2 CRDM coil stacks (204, 2SA3). These coil housing were stained with iron oxide caused by the spray of the leak. All tests were satisfactory. Of the four thermocouple columns that were not affected, three* received surface NOE examinations Dye Penetrate Test and no indications found. Thermocouple column #3 could not be accessed and no PT was conducted. This column was visually inspect during the 102% hydro that was performed at the end of the canopy weld repair; ref DR-SSP-87.314. 5.0 CLEAN-UP OPERATIONS 5.1 General Boric Acid Clean-up and Acceptance Criteria All cleaning was done with stainless steel wire brush, scotch brite and damp rags. The entire head and flange area was vacuumed. The following acceptance criteria was used by Station System Engineers to accept the clean-up operation.

Insure counterbore crevices in the head at the base of CRDM's are free of boric acid deposits to prevent accelerated corrosion at those locations due to galvanic reaction. Dry immediately after any wet cleaning to minimize rusting. Clean top of stud assemblies completely. Dry immediately after any wet cleaning to minimize rusting. No staining from boric acid permitted. Apply Neolube for protection after inspection results are acceptable. Clean head low alloy steel surface completely. Dry immediately after any wet cleaning. Some staining is acceptable on head, however, no boric acid deposits are permitted. Clean CRDM Adapters

a.

CRDM adapters that are adjacent to the leaking adapter and are able to be reached by hand should be cleaned using demineralized water and site approved scotch brite. Note the precautionary drying of the carbon steel head in item above.

b.

Cleaning of those CRDM adapters that are not accessible within arms reach and have some boric acid deposits, should be cleaned remotely using poles that have clean rags attached. Demineralized water can be used providing the precaution with respect to drying is maintained.

c.

Tightly adhering boric acid film can remain on the CRDM adapter until the next outage. The area between the closure flanges and the removed studs must be completely cleaned of boric acid deposits. No staining permitted on the cleaned studs. The studs must be relubricated with Neolube after inspection results are acceptable. Acceptance Criteria for terminating stud removal: when one stud on each side of stud #24 meets the following criteria:

a.

Opaque blotches of white (crust) boric acid are unacceptable.

b.

White film (smear) on surfaces (no depth) is acceptable.

c.

Widely scattered random crystals of white boric acid are acceptable and are probably indicative of residual boric acid on the flange during stud insertion.

d.

Normal light surface rust (red) in the area of the vessel flange gap is acceptable. This criteria is based on the hypothesis that the studs furthest from the leak should have boron residue the same or less than the last pulled stud.

7.0 WELD REPAIR (DR-SSP-87-298) A split canopy has been supplied by Westinghouse under NDPR No. PI-222971 as a Code Class 1 component (See attachment 7) and welded per Welding Procedure Specification NDWP-49-0. Detailed step by step instructions were provided for welding in Code Job Package S-87-168. The NDE on welding (PT per M9-IBV-01C) was successful. Hydrostatic test and documentation were performed during heatup (Mode 3) at 2300 psig (1.02 x Normal Operating Pressure 2235 psig) and temperatures greater than 500F per procedure TI-30. Four welds on the split canopy were be examined for leakage during the hydro test. Westinghouse provided a preliminary evaluation regarding code required stress analysis and was found acceptable by the Authorized Nuclear Inspector until a detailed Code Class 1 Report is provided by Westinghouse. This repair was completed under DR-SSP-87-298 and will be finalized under a minor design change. 8.0 FINAL INSPECTION The following areas were inspected and were free of boron residue: CRDM counterbore in head CRDM boric acid residue Closure head Depression area ori head Flange - top/between studs/nuts/washers/side face CROM vent shroud and internals CRDM vent pipe supports Cavity seal ledge CRDM duct work at Stud #19 Groove between flanges Shank and threads of studs removed. This final inspection was completed by Station System Engineers and Westinghouse. 9.0 POTENTIAL FAILURE MECHANISMS Only a postulation can be made at this time as to the root cause of the failure and subsequent leaks.

This postulation is based on the failure analysis of two identical welds from another plant that failed and leaked and from several repairs of identical and similar welds where it was possible to determine the cause of the leak. Leaks are usually associated with the purge hole which is used to introduce inert gas to the back side of the weld to prevent oxidation while the weld is made. Occasionally, the purge hole blows out when it is welded over and must be repaired as an additional operation to the original weld. There are two known cases of such repair welds failing and leaking after seventeen years of service. Occasionally, the purge hole does not blow out but is not fused properly when welded over. There are several suspected cases of such a condition eventually failing and leaking. The duration of plant operation ranged from one month to several years. More accurate root cause analysis of #5 thermocouple seal weld failure will be determined by Westinghouse after the core exit thermocouples are removed during 4th refueling outage (DCR-2EC-2232) and a portion of this column is sent to Westinghouse. A report will be provided to Nuclear Regulatory Commission when this information becomes available next year. 10.0 BORIC ACID CORROSION INFORMATION SURVEY The principal effect of unremoved boric acid residue in a moist condition that can concentrate is the wastage (or general dissolution corrosion) of both steel and stainless *teel. The general corrosion rate (wastage) for 15% to 25% concentrated boric acid under aqueous, aerated conditions at 200 to 210 degrees F can be significant and is estimated at approximately 400 mils/month based on Westinghouse test results. The wastage under deaerated condition is somewhat lower (Up to 250 mils/month). The boric acid wastage, however, is significantly low when the acid exists in dry crystal form at higher temperatures (approximately 500 degrees F). This is estimated at as low as 10 to 20 mils/yr. Thus the extent of wetness or moisture at the location of interest plays significant role here. The corrosion rate seen on the reactor vessel head is consistent with the estimated 400 mil/month wastage rate based on the Westinghouse studies under aqueous conditions.

11.0 FUTURE ACTIONS An evaluation is in progress regarding the feasibility of installing viewing ports on the CRDM ventilation shroud for both Salem 1 and 2 which can be used to view any leakage in various penetrations on the top of closure head (S.E. by the next available refueling). An evaluation is in progress regarding the feasibility of installing additional in-strumentation to detect increase in moisture and/or boric acid in CRDM ventilation ducts (S.E. by 4/30/88). Westinghouse's owners group involvement is being sought. Receive, review and accept Code Class 1 stress analysis* report £or (Salem Unit 2 specific) split canopy welding on #5 thermocouple column (W&SE by 9/30/87). Actions to improve leak detection Salem is taking several additional steps to ensure that leaks in borated water systems are detected as soon as possible.

1.

The accessible reactor head area will be visually inspected for boron deposits and other evidence of primary system leakage whenever the plant is taken to Mode 3 if an inspection has not been performed in the last 30 days.

2.

A leak inspection procedure will be written to add more specific instructions for inspecting components in the containment which may be affected by a boric acid leak. A detailed root-cause-analysis will be performed on affected #5 T/C column by Westinghouse after it is removed.. (Spring of 1988).

13.0 EVENT CHRONOLOGY 5/28/87 8/7/87 8/8/87 8/9/87 8/10/87 8/11/87 8/12/87 8/13/87 8/14/87 8/15/87 8/16/87 8/17/87 Increase in the noble gas channel (2R12A) 4k to 9k indicating a possible new leak. (Ref. 6,2,18). Unit removed from service for main transformer repairs Cooldown to Mode 5. Containment entry by System Engineers: inspection for RCS leakage. Boric acid found on CRDM duct work on reactor head. Planner enters containment to scope out job. Request System Engineering to further investigate. System Engineer enters containment with Maintenance to remove shroud; inspect T/C column below conoseal. Discovered.3 pin hole leaks in the lower canopy seal on #5 T/C column. Westinghouse contacted for support. Will send welding engineer to site. Meeting with Westinghouse and Salem personnel to discuss repair. Begin removing shrouding and mirror insulation. Decided on split canopy type of repair. Shroud and insulation removed. Pile of boric acid found covering 5 studs. Clean up operation started. Project team formed. Welder qualifications begin. PT'd T/C columns 1,2, & 4. Clean up activities continue. Preparations to remove 24 stud. Clean up activities continue.

1. 24 stud pulled.

24 stud cleaned and MT'd. Westinghouse inspection of stud and stud hole cleanup. Vessel flange gap inspection began with fiberscope. Started setting up for split canopy weld repair.

8/18/87 Reinstalled #24 stud. Decision made to pull more studs.

1. 25 stud pulled and MT'd.

Stud hole cleaned. Weld of split canopy. 8/19/87

1. 23 stud pulled, cleaned, and MT'd.

Stud hole inspected.

1. 26 stud pulled, cleaned and MT'd.

8/20/87 Pulled studs 27, 28, 22, 54 and 29 for inspection, cleaning. 8/21/87 Pulled studs 22, 21 and 30 for inspection, cleaning. Megger and OHM check on 2 CRDM coil stacks-SAT. 8/22/87 . Pulled stud 31, 20 for inspection and cleaning. Removed #23 CRDM vent duct work. 8/23/87 Pulled stud #34, stud was accepted. Pulled stud

1. 17, stud was accepted.

Re-entered Mode 5. Shroud and insulation reassembled. 8/25/87 1000 PSI hydro of split canopy weld and visual leak check of T/C column #3. 8/26/87 2300 PSI hydro of split canopy weld and visual leak check on T/C column #3. Unit synchronized.

0 LO .-i

14.0 REFERENCES

Boric acid cleanup criteria, Westinghouse Memo MED-PCE-5131 dated. Plant operation with boric acid deposition on the RPV studs, Westinghouse Memo, Memo MED-PCE-5118 dated August 19, 1987. Salem Unit-2 reactor vessel stress evaluation with wastage, Westinghouse Memo: MED-PCE-5142. Engineering and Plant Betterment Safety Evaluation No. Closure head pit evaluation. Reactor vessel closure studs, E&PB engineering evaluation No. S-C-R200-MEE-137 Rev. 0 dated November 7, 1986. NRC Regulatory Guide 1.65 "Materials and Inspections for Reactor Vessel Closure Studs." Appendix G to 10CRF50. Deficiency Report DR No. SSP-87-298 for weld repair on

1. 1 thermocouple cqlumn.

Deficiency Report DR No. SSP-87-308 for support steel of mirron insulation. Deficiency Report DR No. SMD-M-87-107 "Grinding, VT & MT Examination of Pits on Closure Head". 'Deficiency Report DR No. SSP-87-313 "RPV Stud Removal, Cleaning and Examination". Deficiency Report DR No. SSP-87-314 "PT on thermocouple column #3 not done." Various inspection reports for inspections performed by R. Tome from Westinghouse. Receiving Nonconformance Report RNR No. MC-87-0644. Maintenance Department Procedure No. T~209 "Inspection of Unit #2 reactor vessel studs for damage due to leaking boric acid. Preliminary evaluation of Salem Unit #2 Instrumentation on part repair design compliance.

SECL-87-396 Revision 1 Cllstamer Reference No ( s). Westirghouse Reference No ( s) * (Cl'lan:Je control or RFQ As Applicable) NS-RCSCLJ-CLir-87-486 Revision 1 WESTINGHOUSE NUCJ:FAR SAFETY E'VAWATION OiECK LIST

1)

NUCLEAR PIANT(S)_-=s=~==-UNIT==-"2=----------------

2)

OiECK UST APPLICABI.E 'IO: EVAIIJATION OF 'lHE REACTOR VESSEL FOUOO!NG 'lHE CANOPY WEI.D lEAK

3)

'Ihe safety evaluation of the revised proce::lure, design chan:Je or no:li.fication required by 10CFRS0.59 has been.prepared to the extent required an::l is attached. If a safety evaluation is not required or is i.ncarplete for arr:J reason, explain on Page 2. Parts A and B of this safety Evaluation Clleck List are to be ccnpleted only on the basis of the safety evaluation perfonned. OiECK UST - PAR!' A (3.1) Yes_ No_K._ A chan;Je to the plant as described in the FSAR? (3.2) Yes_ No_K._ A chan;Je to procedures as describeu in the FSAR? (3.3) Yes_ No_K._ A test or experilTlent not described in the FSAR? (3.4) Yes_ No_K._ A ~e to the plant technical specifications (~A to-the Operatin; License)?

4)

OiECK UST - PAR!' B (Justification for Part B answers must be included on page 2.) (4.1) (4.2) (4.3) (4.4) (4.5) (4.6) (4. 7)

!r.

.;t,. . )i Yes_ No_K._ Will the probability of an acx:ident previously evaluated in the FSAR be increased? Yes_ No_K._ Will the consequences of an accident previously evaluated in the FSAR be i.ncrea.sed? Yes_ No_K._ May the poss.ibility of an accident which is different than aey already evaluated in the FSAR be created? Yes_ No_K._* Will the prd::>ability of a malfunction of equipnent .inportant to safety previously evaluated in the FSAR be increased? Yes_ No_K._ Will the oonsequences of a malfunction of equip-ment.inportant to safety previously evaluated in the FSAR be increased? Yes_ No_K._ May the poss.ibility of a malfunction of equipnent .inportant to safety different than any already evaluated in the FSAR be created? Yes_ No_K._ Will the margin of safety as *defined in the. 'bases to arr:J technical specification be redticed?* PAGE 1 OF 6 j.. 1

SECL-87-396 Revision 1 CUstarner Reference No ( s)

• Westin;Jhouse Reference No(s).

(Cl'larxJe Control or RFQ As Applicable) NS-RCSCirC/I.r-87-486 Revision 1 WESTINGHCUSE NUCI.FAR SAFEIY EVAil.IATION CliECK LISI'

1)

NUCI.EAR PIANI'(S)_~SAllM~~~UNIT~~2---------------

2)

CHECK LIST APPLICABLE TO: EVAIJJATION OF '!HE RFACIOR VESSEL FOLI.CMING '!HE CANOPY WEI.D I.EAK

3)

'!he safety evaluation of the revised procedure, design change or no:lification required by 10CFRS0.59 has been prepared to the extent required an::l is attached. If a safety evaluation is not required or is incarplete for any reason, explain on Page 2. Parts A an::l B of this Safety Evaluation Qieck List are to be completed only on the basis of the safety evaluation perfonned. CHECK LIST - PARI' A (3.1) Yes_ No_x.. A ~e to the plant as described in the FSAR? (3.2) Yes_ No_x.. A change to procedures as descr:ibeJ in the FSAR? (3.3) Yes_ No_x.. A test or eJq:>eriment not descrilJed.lll the FSAR? (3.4) Yes_ No_x.. A change to the plant tedmical specifications (Appen:lix A to the Operat:in;J License)?

4)

CHECK LIST - PARI' B (Justification for Part B answers must be included (4.1) (4.2) (4. 3) (4.4) (4.5) (4.6) (4.7) on page 2.) Yes_ No_x.. Will the probability of an accident previously evaluated in the FSAR be.increased? Yes_ No..JL Will the consequences of an accident previously evaluated in the FSAR be.increased? Yes_ No..JL May the possibility of an accident which is different than any already evaluated in the FSAR. be created? Yes_ No_lL Will the probability of a malfunction of equipnent in'lportant to safety previously evaluated in the FSAR be increased? Yes_ No..JL Will the con.sequences of a malfunction of equip-ment in'lportant to safety previously evaluated in the FSAR be :ir¥::reased? Yes_ No_x.. May the possibility of a malfUnction of equipnent

inp:>rtant to safety different than any already evaluated in the FSAR be created?

Yes_ No_K_ Will the margin of safety as *defined in the 'bases to any technical specification be redtJc:ed?* PAGE 1 OF 6 1

5/28/87 8/7/87 3/8/87 8/9/87 8/10/87 3/11/87 8/12/87 8/13/87 8/14/87 8/15/87 8/16/87 8/17/87 Increase in ~he noble ~as channel (2Rl2~) 4Y. tc 9~ indicating a ~ossible new leak. (~e:. 5 r 2' 13). UGi~ re2ov~d ~ro~ service fer ~ain transfor~er repai=s Cooldown to Mode 5. Ccntain~ent cnt~y by System Engineers: inspection for RCS leakage. Boric acid found on CRDH due~ work on re3ctor h2ad. Planner enters ccntain~ent to scope out job. Request Systa= Enginaering to further investigate. Systen Engineer enters contain~ent with Haintananca to remove shroud; inspect T/C column below conoseal. Discove~ad 3 pin hole leaks in the lower canopy seal on #5 T/C column. Westing~ouse contacted for support. Will send ~alding angineer to site. 11eeting with Westinghouse and Salem personnel to disc~ss re~air. Bagin removing shrouding and ~irror insulation. Deci1ed on split canopy type of repair. Shroud and insulation removed. Pile of boric acid found covering 5 studs. Clean up operation started. Project team formed. Welder qualifications begin. PT'd T/C columns 1,2, & 4. Clean up activities continue. rernc*..re 24 stud. Clean up activities continue.

1. 24 stud pulled.

Preparations to 24 stud cleaned and MT'd. Westinghouse inspection of stud and stud hole cleanup. Vessel flange gap inspection began with fiberscope. Started setting up for split canopy weld repair.

8/18/87 3/19/37 8/20/87 8/21/87 8/22/27 3/23/87 8/25/87 3/26/87 Reinstalled #24 st11d. Decision ~ade ~o pull ~ore studs.

1. 25 scud pulled and MT'd.

Stud hole clea~ed. W~ld c~ splic ca~opy.

1. 23 s~ud p~lled, clea~ed, ~nd MT'd.

Stud ho:e inspacted. ~26 stud pulled, =leaned and HT'd. Pulled studs 27, 23, 22, 54 and 29 ~or inspection, clea~ing. Pulled studs 22, ~~ and JO for inspection, cleaning. Mag~~r and OHM check on 2 CRDM coil stacks-S~T. Pulled st~d J:, :o for inspection and cl=aning. Ra~o7ed #23 CRDM vent duct work. Pulled st~d #34, stud was accepted. Pulled stud

1. 17, st~d was accepted.

Re-entered Mode 5. Shroud and insulation reassembled. 1000 FS: ~ydro of split canopy weld and visual leak check of T/C column #3. 2200 PSI hydro of split canopy weld and visual leak check on T/C column #J. Unit synchronized.

14.0 RSFERElICES Boric ~cid cle3~up crite=ia, Westi~g~~~sa ~a~o

J-?C:::-::~2.:1._ C.3.1:;;:::_.

Plant oparation ~ich ~oric a=id deposition on the RFV studs, Westi:igl:cuse r~e_r,o, l*!err<o HED-FCE-5113 dated August 19, 193/. Salem Unit-2 re~cta~ Vessel stress evaluation with wastage, Westinghouse Me~o: MED-PCE-5142. Engineering and Plant Betterment Safety Evaluation No. Closure head pit evaluation. Reactor vessel closure studs, E&PB engineering evaluation No. S-C-R200-MEE-137 Rev. 0 dated November 7, 1986. NRC Re*gulac.cry Gi.lide 1.65 "Materials and Inspections

or Reactor Vessel Closure Studs."

Appendix G to lOCRFSO. Deficie:lcy Report DR No. SSP-87-298 fer weld repair on

1. 1 th2r~oco~ple column.

Deficiency Report DR No. SSP-87-308 for support steel of ~irron insulation. Deficiency Report DR No. SMD-M-87-107 "Grinding, VT & HT Exa:nination of Pits on Closure Head". Deficiency Report DR No. SSP-87-313 "RPV Stud Removal, Cleaning and Examination". Deficiency Report DR No. SSP-87-314 "PT on thermocouple column #3 not done." Various inspection reports for inspections performed by R. Tome from Wistinghouse. Receiving ;ronconformance Report RNR No. MC-87-0644. Maintenance Depart~ent Procedure No. T-209 "Inspection of Unit #2 reactor vessel studs for damage due to leaking coric acid. Preli:ninary evaluation of Salem Unit #2 Instrumentation on part repair design compliance.

0 Ul .-I

SPLIT JACK SCREWS

• Apply NEOLUBE to Th'ds & Contact Surfaces Ff'MALE FLANGE

/.. HARMON CLAMP

---

Apply NEOWBE to All Cla111p Contact Surf aces l

,.r

~ E-t < ~ ~ ~ ~ SALEM UNIT 2 LEAK RATES ATTACHMENT 2 1 - CONTAINMENT SUMP - HIGHEST AT.47 GPM ~ IDENTIFJm Lf.AKACE - HIGHESr AT.59 GPM

• UNIDENTlAED LEAKAGE -

HIGHESr AT.29 GPM .9 t--~------~---------~~~--i .5 .4 .3 G-- .2 . 1 1 2 3 4 5 6 7 3 s 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 JULY/AUGUST 1987

7/lL 1/8

• 1 I

VEMT PIPE SUPPORT ATTACHMENT 3 7/lL

LEAKING PENETRATION 15/~2 ATTACHf.iEiH 3 1/8 15/32 39" HEAD FLANGE

3/16 ATTACHMENT 3 7132

5/16 ATTACHMENT 3 I.D. OF VENT SHROUD SUPPORT RING O.D. OF Vun

• SHROUD SUPPORT RING

Sw. R.I. STRAIGHT BEAM lLAMINATION EXAMINATION RECORD 'RO.IECT Ne. SITE* DATE*CDAY - llON. -YR. ) Tllll:

• 114 - IUI. C LOC* I SHEET No.

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• I. -

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• IN %

CHANGE 100 So 100 >< ~ A . 10 . -ua JI TO 48 ~J*e:,

• O l\\S rJ A.

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• * ** -111 I*"***-**- ao 1

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• ~~~i'9":=~'*FY I WATER 0 t! A()S.;cA) NVL V ~

SEARCH UNITS REFERENCE BLK. No.:.fl W - C\\ o° CL. S. I 0° ( ATT.J ~ * ,,A, CALIBRATION VERIFICATION qAND SERIAL HUMBER JJ ~~~ 1-11-11 ~ WIZ-1 .!2. 8 <<t' ii.l A TIME \\ '2. '~ ~;:;; '.7C-INITIALS @) .I I\\ I

• ZE 1.0Ro.

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2.

BASIC CALIBRATION CAIL'E TYPE ELAY; 0'\\Ll-1 BLOCK Na. IL I 111174 ~ IATL. CAL: tqs dB'" LENGTH 3..k_IN. I ST ECHO I OF AMPLITUDE 1All8E 10 11112 a 2ND ECHO L~ OF MIPLITUDI

• AMPING:

N. &a.\\ ~__:_.. ~ OTHER C 2ND ECH ST ECHO) IEP RATE: 3\\< BASIC CALIBRATION IASIC CALIBRATION tlDEO: l\\\\hD M. BLOCK Ne. . k\\ BLOCK No. \\ ILTER: HI I ST ECHO A LlllEI Of' AMPLITUDE llT ECHO

• 4111~UNE~F AMPLITUDE ACK USED:

R I ND ECHn _:_;;;>*a ::A11eEI OF AllPLIT~* 11-1 Nb ECHO ~B LINE F AMPLITUDE ~ ( 2 ND ECHO

• I ST ECHO I

. ~ IOOE OF TRAI~ NOfff\\AI.. f2 ND ECHO - I ST ECHO I ~ tEVIEWED B~o~~,J I INT ~EVEL:.JJ;- DATE: 17 /)U.6 f:: r

ATTACHMENT Sw. R. I. MAGNETIC P~RTICLE EXAMINATION RECORD ftllO.llCT Ne.: 1111: DAT.I: IDAY-llO -Yll t TllE: 124... CLOCK t SHEET Ne.: I 1-1553 SALEM (i,E14, sm. &1.141T1f 11o A. \\Jc;, 1r1 HAM ITAllTED: I l 0£1 l~C.OOI EXAll HIDED : I \\ !.D UAMINATIOll AllEA: LINE / IU*AllEllBLY: 'Hir'MYS!b~ L

• LOCATION:

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• AND: l>>l.Al'* ~Af 911AND:6'-At(9 t:l4tlj)

TUii: 'la5"tl 11.)J APftLICATION: FLOODING IB

• lllAL Ne.:

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MATERIAL I l I APPLICATION: MAND: N "

• RAND:

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• Hiil.i
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SECir-87-396 Revision 1 If the answers to any of the above questions are unknown, irxticate urrler 5) REMARKS ani explain below. If the answer to any of the above questions in 4) cannot be answered in the negative, based on written safety evaluation, the change cannot be approved without an application for license amen:hnent submitted to the NRC p.lrSUallt to 10CFRS0.59.

5)

REMARKS:

• NONE

'Ihe followirg summarizes the justification upon the written safety evaluation, (*) for answers given in Part B of the safety Evaluation Oleck List: SEE MTACliED SAFEI"i EVAIIJATION REVISIONS 'IO '!HIS SAFETY EVAIIJATION CliECKLIST ARE IlIDICATED BY A VERTICAL BAR Ill 'IHE RIGHI' MAR:;IN (*) Reference to document(s) containirq written safety evaluation: roR FSAR UPDATE Section: ____ Pages: ____ Tables: _____ Figures: ____ _ Reason for / Description of Chan:Je: PAGE 2 OF 6

SECir87-396 Revision 1 NS-RCSCL-C/I.r-87-486 Rev 1 Page 3 of 6 SALEM UNIT 2 - SAFEIY EVAI.IJATION EVAIIJATION OF 'IHE RFACIDR VESSEL FOLL.CMING 'IHE CANOPY WEI.D LEAK INI'ROIXJCTION While the Salem 2 Nuclear Power Plant was da.m due to transfonrer problems, an inspection of the reactor vessel area was perfonned by the utility (Public Service Electric an:i Gas) personnel on August 7, 1987. 'nlis inspection was prompted by high airbom reaclin;Js in contairanent which the utility has reported as starting to cccur July 17, 1987. 'lhe utility also reported that they had an unidentifie:i leak rate of.2 to.3 gpm during th.is time. CUring the inspection, the utility founi boric acid crystals on the reactor vessel closure head. 'lhe source of the boric acid was identified as three pin holes in a canopy weld of a thenocx::ouple column ('lhenrocouple Col\\.mm #5). After an initial assessment, the utility an:l Westinghouse perfonned the evaluation an:i recovery. 'Ihe pw:pose of this safety evaluation is to (1) provide backgrcun:i infonnation related to the canopy weld leakage, (2) describe the effect of the boric acid on the reactor vessel head, (3) to evaluate the corrective action taken, an:i (4) allow the plant to start up an:i continu~ safe plant operation. LICENSING APPROACli AND SCDPE Olapter 10 of the Code of Federal Regulations, Section 50.59 (10 CFR 50.59) allows the holder of a license authorizin; c:p3ration of a nuclear power facility the capacity to initiate certain d'larr3'es, tests, an:i experiments not described in the Final safety Analysis Report (FSAR). Prior Nuclear Regulatory camnission (NRC) approval is not necessary to in'iJ!ement the proposed charge, test, or experiment if it does not involve an unreviewed safety question or change in the technical specifications incorportated in the license. It is however, the obligation of the licensee to maintain a record of these c:h.an;es, tests an:i experiments to the facility to that extent that such d'lange ilrpact the FSAR. 10 CFR 50.59 further sti?Uates that these records shall include a written safety evaluation which provides the bases for the detennination that the change, test, or experiment does not involve an wireviewed safety question. '!his evaluation addresses the integrity of the reactor vessel head for the plant to start up an:i continue safe plant ~tion as a result of the boric acid leak. It should be note:i that the Salem Unit 2 FSAR an:i Technical Specifications are not ilrpacted.

Bll.CKGRCUND SECL-87-396 Revision 1 NS-RC3CL-C/L-87-486 Rev 1 Page 4 of 6 'Ihe proposed hypothesis of how the pits and craters were formed on the enclosure head o. D. just inside arx:i urrler the vent shroud support ring is as follows: 'Ihe instrument port leak spray was cooled by the CRI:M cooling air an:i ran dcM1 the I. o. of the vent shroud support ring. 'Ihe liquid dri~ onto the closure head, evaporated, leaving boric acid crystals and was rewetted by the next drop of liquid off of the shroud ring causing the boric acid to.become 100re concentrated and attack the head in the pitted area. 'Ihis explains why the pits became so deep in such a short time. 'Ihe boric acid then ran down the head causing the erosion grooves to develop before it evaporated. After a while, the boric acid built up on the side of the head where it dried out with the continuing boric acid fran the leak drying on top of the initial boric acid stream. 'Ihis explains the minimal attack on the lower portion of the head done arx:i the top of the closure stud assemblies. INSPECTION RESULTS - CDRRECI'IVE ACrIONS PERFOR-IBD Follc:Ming are the results of the inspection and the corrective actions performed at salem Unit 2:

1.

A boric acid buildup was foun:i on the head dane. 'Ih.e boric acid had flowed down the head dane onto the flange arx:i.impi.nged on the tops of 4 closure studs, nuts and washers. '!his area was cleaned arx:i inspected. Several pits and grooves have been fourxi on the closure head in the area of the boric acid buildup. one of these grooves initiated at the camterbore of the leaking.instrument port. 'Ihe other pits an:i grooves started umer the cooling shroud ~rt rin;J arx:i exterv:i dC7ot1n toward the closure head flange. '!be pits are approx:iltately 1 to 3 inches in diameter and the grooves are several inches in lergth. 'Ih.e maximJm depth of the indications is 15/32 11

• Figures 1-A through 1-D are a set of fUl1 scale tracirgs of the pits in the closure head.

'lhe tracin;s follow the contour of the head and the paper should be placed edge to edge with no overlap. 'Ihe numbers represent the de.epest portion of each pit and the dark solid area represents the shape of the deepest portion. '!be rest of the depression is shallower and blen::ls out to the original head contour. Figure 2 sha.NS the camtertx:>re wastage at the base of the leaking .instrument port

• SECL-87-396 Revision 1 NS-RCSCL-C/L-87-486 Rev 1 PAGE 5 OF 6 An ultrasonic examination (ur) of the head thickness arrl a Magnetic Particle Test (MI') of the pitted area was performed at salem 2.

'Ihe ur sha.vs the head was 7.6" thick in the SOlll"rl rootal area arourrl the pits. 'Ihe base of the leakirg penetration is 7.4" arrl probably represents the effect of the erosion wash observed at the base of that penetration. For conservatism, Westin;house calculations were based on the head minimum drawing thickness of 7.011 * 'Ihe MI' shCMed no recordable in:lications in the area of the pitting corxiltion on the reactor vessel head. 'Ihe pits arrl grooves have been cleaned an::i dressed to reIIYJVe sharp edges.

2.

It was found that ooric acid had flCMed down the side of the closure head flange. A fiberscope inspection of the gap between the closure head arrl vessel flange revealed the presence of ooric acid crystals from studs 18 to 30. 'Ihe loose ooric acid in the gap between the flanges aroun::l the studs was reIIYJVed to the extent possible by mechanical cleaning. Studs 18 to 31 had ooric acid de,EX>Sits on the shank at the flange gap an:i in the threads at the vessel stud hole counterbore. Studs 17 through 31 an:i stud 34 were rem::ived one at a time. 'Ihese studs were cleaned, inspected, an:i reinstalled. No signs of damage was foun:i.

3.

An inspection of the annulus urxier the cavity seal ledge between the reactor vessel an:i cavity revealed no evidt:nce of ooric acid.

4.

Evidence of ooric acid has been foun:i on the CRrM head adaptors. 'Ihis area was cleaned. am inspected an:1 no damage was c::bseived.

5.

'Ihe air duct in the vicinity of stud 19 was inspected an:i ooric acid crystals were fourxi to exten:i approxllna.tely 3 feet inside. 'Ihis duct was cleaned. 'Ihe remainin;J ducts showed no evidence of ooric acid.

6.

Evidence of sane damage was foon:i on a vent pipe 51.JRX>rt. 'Ibis was cleaned by the utility; no repair was required.

7.

'Ihe remavable insulation panels over the closure studs that were not renx::wed revealed no evidence of ooric acid. '!he rem:wed panels were inspected am shcMed no evidence of ooric acid deposition.

a.

'Ihe i.nconel CRIM head adaptors an:i the stainless steel reflective insulation in the CRIM pattem showed surface rust probably cxmri.n:J f:ran the ductile iron coil stacks. '!he anomt of :ruSt present has minimal effect on the stainless steel insulation or i.nCX>nel. CRI:M adaptors. - 9. Red coil checks were made on the two coil stacks where rust was m::st evident on the CEJwfs an::i insulation. 'Ihese tests showed acceptable results.

EVAI.IJATION OF CORRECI'IVE ACTION TAKEN: SECir87-396 Revision 1 NS-RCSCirC/Ii-87-486 Rev 1 Page 6 of 6 EValuation of the effects of the Boric Acid attack on the Reactor Vessel: 'Ihe effect of the boric acid corrlensation on the outside surface of the reactor vessel on stresses ard fatigue life of the structure has been evaluated. It has been foun:i that the stresses ar.d fatigue usage factors for a reactor vessel head ard head adaptor are still below the limits of ASME Section III of the B&P'V Code. 'Ille :reduction in thickness around the closure head penetrations also have been evaluated. '!he analysis shc:MS that the reduced head thickness arrl enlarged crevice around the penetration are acceptable. CONCIDSION Based. on the above evaluation of the reactor vessel head in:lications ar.d .other con::litions existing following t.he canopy weld leak at the Salem 2 nuclear pc:Mer station, it has been determined no unreviewed safety question exists as defined by 10CFRS0.59. 'lherefore, following satisfactory repair of the canopy seal weld leak by the utility, the plant can start up arrl continue safe plant c:peration.

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Wo. ~. L e°'-\\c:...j'.., \\1JC "*5 ATTACHMENT 5

Gap a-,.' 'rnE.RMOCOUPLE COLUMN ATTACHf-lENT 6 I I I I Ll i I I ~-CROM

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2!Na/Ol2117: 1 Q CALCULATIONS AND E~ALUATIONS FOR S.A.LEM UN IT #2 REACTOR VESSEL FOLLOWING RECOVERY FROM INSTRUMENT PORTlt5 ~ANOPY LEAK R. M. BLAUSHILD S. L. ABBOTT August 1987 Approved~ F. a.-.1-= D. E. Boyle, M~ager Primary Component Engineering ~.....

Prepared by: R. M. Blaushild, Engineer Primary Component Engrg. = : ~ ~ : r : :,; T : : N Checked by: /' S. L. Abbott, E~gineer Primary Component Engrg. Approved by: 'D.E.Y= D. E. Bo~anager Primary Component Engrg. It is hereby certified that the components described in this report have been reconciled with the requirements of the 1965 Edition thru Winter 1966 addenda or later Editions of Section III of the ASME Boiler and Pressure Vessel Code as required by Section XI of the ASME Boiler and Pressure Vessel Code l.* 7 S. L. Abbott, P.E. 2sH11oa2aa1:' o 8-3( - 87 Date

• ----~.... -

-j~ I."<.-.\\,,., I Evaluations of the stresses and fatigue behavior of the Salem #2 reactor vessel head and head adapter due to local cooling resulting from the instrument port canopy leakage, calculations of the minimum required head thickness in the area of the head pitting condition, and the determination of the available area of reinforcement in the head for the leaking instrument port penetration versus the required value were made. The reactor vessel closure studs were evaluated for the effects of the maximum expected material loss due to corrosion for the remaining 6 months of operation until the refueling outage. The calculations show that the structural integrity of the vessel and closure stud would not be affected as a result of the material loss. 25i111082817: 10 i i

Title Page Summary of Calculations Performed 1 Evaluation of the Thermal and Fatigue Effect on The Head Adapter (RMB-870815-1) Evaluation of the Thermal and Fatigue Effect of the Reactor Vessel Head Local Cooling (RMB0-870815-2) Evaluation of Minimum Penetration Reinforcement (RMB-870815-3) Evaluation of Reduced Head Thickness Below Penetration (RMB-870815-4) Reactor Vessel Stud Reduced Diameter Evaluation (SLA-081887) Determination of Excess Thread Engagement (SLA-081987) References 25981/082887: 10

i i 3

9 18 23 25 28 . 30

1.

Head Adapter Local Cooling - Cale. #RMB-870815*-1 Maxim~m stress intensity on the outside surface is calculated using a simplified elastic-plastic analysis oT = 57.0 ksi, total stress intensity is 0 6 - or= 61.0 ksi. The cumulative usage factor of 0.002 has been determined assuming a very conservative approach.

== Conclusion:== The cooltng of the head adapter due to leakage is acceptable.

2.

RV Head Local Cooling - Cale. #RMB-870815-2 The maximum total stress on the outside surface of the dome aTeT = 78.4 ~80.1 ksi, the cumulative usage factor of 0.012 has been determined assuming very a conservative approach.

== Conclusion:== The local cooling of the closure head due to the canopy leakage is acceptable.

3.

Reduction in Thickness of RV Head - Cale. #RMB-870815-3 The minimum thickness requirement of the closure head has been determined to be tmin = 4.216 inch versus the design drawing specified minimum of 7.0 inch in the area of the pits. The area required for the reinforcement of the instrument port in the head is A = 23.06 in2 versus the area available AA= 26.24 in2 in the head with a 1/2" reduction in the head thickness to compensate for the presence of the pits.

== Conclusion:== with 15/32 = 0.5 in in reduction of RV head thickness, the minimum thick and instrument port penetration reinforcement requirements are still provided. 2S9ta/0112N7:10 1

4.- ~~~=*"*~ess a: =~e Lower s~~e of ~=~:st Penetratl:1 - ,:;RMB-870815-4 The ~equired area in the lower side of the instrument port penetration calls for a minimum length of Lmin = 5.6 inch for the reduced head thickness under the pits. The available Lunit = 6.35 in.

== Conclusion:== Area required at the lower side of the instrument port penetration is adequate even with a 0.5 inch reduction in thickness of the closure head.

5.

RV Stud Reduced Diameter Evaluation ~ Cale. #SLA-081887 Dry boric acid on carbon steel can result in corrosion rates up to 0.020 inches per hear. The maximum stud diameter reduction for the remaining six months until the refueling outage is therefore 0.020 inches. The stud shank diameter could be reduced by 0.032 inches in accordance with the Combustion Engineering sizing calculation or 0.117 inches in accordance with ASME Section III, Appendix E and still satisfy the allowable membra*ne stress.

== Conclusion:== A diameter reduction of 0.020 inches due to corrosion in the stud shank would not affect the structural integrity of the shank.

6.

Determination of Excess Thread Engagement - Cale. #SLA-081987 There is 1.19 inches more stud thread engagement in the vessel flange stud holes than is required during normal operation. Conclusio~: 1.0 inch of stud thread could be lost due to corrosion without affecting the structural integrity of the closure stud and reactor vessel. 2sga11oa2aa1: 10 2

RErERENC~S

1.

ASME Boiler and Pressure Vessel Code, Section III, Nuclear Power Plant Components, American Society of Mechanical Engineers.

2.

ASME Boiler and Pressure Vessel Code, Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, American Society of Mechanical Engineers.

3.

Analytical Report for Salem Unit No. 2 Reactor Vessel by Combustion Engineering, Inc., Report NO. CENC-1155.

4.

Definition of the Pitts and Grooves in Salem Unit 2 RV Head. Transmitted by R.* Tome, August 19, 1987.

5.

Definition of Salem Unit No. 2 R. V. Head Pitting and Leaking Penetration Counterbore Wastage. Transmitted by R. Tome, August 19, 1987.

6.

Transient Data from December 1986 through July 1987. Transmitted by T. Baird and R. Tome, August 21, 1987. 2511Sa/082Sl7: l 0 3

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