05000336/LER-2002-001
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| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 3362002001R00 - NRC Website | |
This LER reports two events related to Reactor Coolant System (RCS) pressure boundary leakage. The two events were discovered during a single activity, namely inspections conducted during refueling outage 14 (2R14).
Therefore, the two events are reported in a single LER.
First Event:
1. Event Description On February 19, 2002, with the plant in Mode 5 (Cold Shutdown), an in-service visual inspection of Millstone Unit No. 2 pressurizer [AB] heater penetrations and pressurizer instrument nozzle penetrations was being performed. Two heater sleeve penetrations were found to show indications of minor leakage as evidenced by boron precipitation build up on the outside of the penetrations. This leakage was too small to have been detected via normal means (containment particulate radiation monitors or other leakage monitoring systems) during cycle 14 operation.
The heater penetration and instrumentation penetration nozzles in Combustion Engineering (CE) designed Nuclear Steam Supply Systems are fabricated from alloy 600 (Inconel 600) and are joined to the pressurizer using partial penetration J-groove welds. The weld metal is also alloy 600. Industry experience at other CE designed Nuclear Steam Supply Systems has shown that these welds and nozzles are susceptible to primary water stress corrosion cracking (PWSCC). The inspection was being performed at Millstone Unit No. 2 as a result of this industry experience.
This event is being reported pursuant to 10CFR50.73(a)(2)(i)(B) as a condition prohibited by the plant's Technical Specifications. Technical Specification 3.4.6.2 states that in Modes 1 through 4 "Reactor Coolant leakage shall be limited to no Pressure Boundary Leakage." From the amount of boric acid build up on the outside of the pressurizer it is conservatively assumed that the leakage could have existed in Modes 1 through 4.
2. Cause The cause of this event was a through wall crack in two of the pressurizer heater sleeves as a result of PWSCC.
This allowed primary coolant into the annulus between the pressurizer lower head and the heater sleeve and therefore was a breach of the reactor coolant pressure boundary. A plant specific analysis of these cracks has not been done but, based upon industry experience, cracks like these are due to primary water stress corrosion cracking of the Alloy 600 heater sleeve. The thickness of the pressurizer lower head adjacent to the two leaking nozzles and other non-leaking nozzles was measured and no differences were noted.
3. Assessment of Safety Consequences The design function of the pressurizer is to maintain reactor coolant system [AB] pressure. This is done with a combination of heaters to raise the temperature in the pressurizer and spray valves to lower the temperature in the pressurizer. The heater sleeves are part of the pressure boundary of the pressurizer. A crack in a heater sleeve creates a leak in the reactor coolant pressure boundary which is not permitted by Technical Specifications.
PWSCC of a number of Alloy 600 penetrations has been observed in CE designed reactor coolant systems.
Cracking has been observed in pressurizer heater sleeves, pressurizer instrumentation nozzles and resistance temperature detector (RTD) nozzles on the reactor coolant piping. All of these penetrations are made the same way with a partial penetration J-groove weld of the Alloy 600 to vessel clad or pipe. This cracking has been observed at virtually all CE designed plants.
The actual safety significance of the cracking found in these heater sleeves is low. The cracks from PWSCC are very tight and the leakage rate was well below the allowable rate of 1 gpm for unidentified leakage. The situation at Millstone Unit No. 2 is similar to the cracking found at all other CE designed Nuclear Steam Supply Systems.
The potential worst case situation of a circumferential crack that could result in the complete loss of the penetration would be well below the limiting hole size, less than 3 inch, for a Small Break Loss of Coolant Accident, and is therefore bounded by the current analysis basis.
4. Corrective Action The leaking heater sleeves have been repaired by the use of Mechanical Nozzle Seal Assembly (MNSA) clamps.
The use of the clamps was approved by the Nuclear Regulatory Commission's (NRC) Safety Evaluation Report (SER), "Safety Evaluation of Relief Request RR-89-35, Temporary Installation of Mechanical Nozzle Seal Assemblies on Pressurizer Heater Nozzles, Millstone Nuclear Power Station, Unit No. 2," dated March 22, 2002 (TAC No. MB4039).
5. Previous Occurrences The PWSCC of the Alloy 600 heater sleeves at Millstone Unit No. 2 is similar to the PWSCC that has been found in the Control Element/Rod Drive Mechanisms (CEDM/CRDMs) at various plants (including Millstone Unit No. 2) within the past year and a half. The materials of construction and the design with partial penetration J-groove welds for the pressurizer heater sleeves are the same as for the CEDM/CRDM and Incore Instrumentation (ICI) nozzles. Millstone Unit No. 2 has recently completed an inspection of 100% of the Reactor Vessel Head Penetrations (RVHPs), 69 CEDMs, 8 ICIs and the vent line, via Ultra Sonic to comply with NRC Bulletin 2001-01.
Three penetrations with shallow cracking on the outer diameter of three CEDMs below the J-groove weld were repaired.
Second Event:
1. Event Description On February 22, 2002, with the plant in Mode 6 (Refueling) at zero percent Rated Thermal Power, helium leak tests were performed on the "A" and "C" Reactor Coolant Pump (RCP) [AB,P] seal coolers [AB,HX] and thermal barrier heat exchangers in an effort to identify the source(s) of the leakage from the Reactor Coolant System (RCS) [AB] into the Reactor Building Closed Cooling Water (RBCCW) [CC] system. This testing confirmed leakage on the "C" RCP. The RCP seal cooler is part of the RCS pressure boundary.
The RCP (Byron Jackson pump) is sealed from the atmosphere by mechanical seals mounted in a removable cartridge consisting of four seals in series. The reactor coolant entering the seal cavity is forced by the auxiliary impeller through an integral heat exchanger (seat cooler) cooled by RBCCW. By maintaining approximately 1 GPM of reactor coolant through the cartridge, the seals are kept at a temperature of approximately 80 to 90 degrees F during normal operation.
Millstone Unit No. 2 began to experience elevated levels of radionuclide activity in RBCCW on October 25, 1999.
The RBCCW samples showed the presence of xenon and iodine consistent with reactor coolant isotopes. After much trouble shooting, initial indications revealed the primary sample cooler as the suspected leak source. After repair of the primary sample cooler indications of the leak disappeared. On November 3, 2000, the leak reappeared and sampling of other systems verified the contamination was limited to the RBCCW system.
Operations personnel began to manipulate select components and determined the leak was limited to the "A" RBCCW header. Based on tritium and boron concentration rate increases in RBCCW, the initial leak rate was calculated to be approximately 10 to 12 ml/min and had dropped and remained between 1 and 4 ml/min over the next month. Detailed troubleshooting was accomplished on RBCCW system components interfacing with the RCS outside containment during plant operations in 2000. The leakage source remained categorized as unidentified since the active efforts to locate the degraded component were unsuccessful and the leakage rate subsequently decreased to levels which were only marginally detectable by December of 2000. At that time, the only major components remaining to be tested were the "A" and "C" RCPs. Once 2R14 commenced, helium testing was performed on both RCPs. Helium was used because the suspected leak point was believed to be so small that an air test or water hydro would not identify the leak. Once the RBCCW side of the RCP was pressurized with helium, a helium detector confirmed that the "C" RCP seal cooler did indeed have a very small leak. The test was repeated and the results were confirmed.
The RCP seal cooler is part of the RCS pressure boundary. The "C" RCP seal cooler leak did exist while the plant was operating in Modes 1 through 4. Therefore, this event is being reported pursuant to 10CFR50.73(a)(2)(i)(B) as a condition prohibited by the plant's Technical Specifications. Technical Specification 3.4.6.2 states that in Modes 1 through 4 "Reactor Coolant leakage shall be limited to no Pressure Boundary Leakage.
2. Cause The extent of the condition is limited to the "C" RCP. Testing verified the "A" RCP did not leak. No other indications or isotopes were found in the RBCCW headers that would indicate the "B" and "D" RCPs leak. This condition appears to be unique to Byron Jackson pumps and associated cooling equipment. Discussion with the vendor indicates their belief that the most likely area of the leak experienced on Millstone Unit No. 2 is in the drilled hole heat exchanger. This region has a thickness of approximately 0.250 inches between the pump shaft and the drilled hole of the heat exchanger. Along the pump shaft in this region there is approximately a 400 degree F temperature gradient. The thermally induced stresses would be maximum in this region. The tube-in-tube heat exchanger delta-T remains relatively constant from inlet to outlet with an average temperature 80 to 90 degrees F.
The stresses in this heat exchanger would be low compared to the drilled hole heat exchanger. Therefore, it is likely that the leak is in the drilled hole heat exchanger. The cause of the leak is uncertain.
3. Assessment of Safety Consequences The leakage was too small to be detected by normal volumetric means. The leak was detected by the RBCCW system rad monitors. The leakage was treated as unidentified leakage because the source was not positively known until testing was conducted during the refueling outage, 2R14. The nuclear safety aspects of a small leak (approximately 10 to 12 ml/min) from the RCS into the RBCCW were evaluated and it was concluded that leakage of this magnitude did not represent a significant safety concern. If a thermal barrier heat exchanger tube rupture occurs, resulting in an intersystem loss of coolant accident into RBCCW, four QA Category I relief valves (located in the RBCCW supply and return for the RBCCW service line to and from containment, relieving into containment) will ensure that the RBCCW motor-operated containment isolation valves can be closed during this event. The relief valves prevent overpressurization of the RBCCW System and prevent an unisolable release of radioactive fluid to outside the containment which would exceed the 10 CFR 100 limits. The setpoint of these valves is 165 psig.
4. Corrective Action The leaking "C" RCP cover and rotating element (which contains the RCP seal cooler) was replaced with a rebuilt spare. The rebuilt spare had been tested prior to the outage and was shown to have no leakage through the thermal barrier heat exchanger. Subsequent to Millstone Unit No. 2's return to power operation, there has been no detectable evidence of reactor coolant ingress to the RBCCW system. The system is continuously monitored for radioactivity via an in-line detector, and water samples are analyzed weekly for long term trending of radioactive contamination in accordance with IE Bulletin 80-10.
5. Previous Occurrences There have been no similar events of RCP seal cooler leaks at Millstone Unit No. 2.
Energy Industry Identification System (ENS) codes are identified in the text as [XX].