IR 05000280/1997006
| ML18153A161 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 08/07/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18153A160 | List: |
| References | |
| 50-280-97-06, 50-280-97-6, NUDOCS 9708220016 | |
| Download: ML18153A161 (19) | |
Text
Docket Nos:
License Nos:
Report No:
Licensee:
Facility:
Location:
Dates:
Inspectors:
Approved by:
U.S. NUCLEAR REGULATORY COMMISSION
REGION II
50-280, 50-281 DPR-32, DPR-37 50-280/97-06, 50-281/97-06 Virginia Electric and Power Company (VEPCO)
Surry Power Station, Units 1 & 2 5850 Hog Island Road Surry, VA 23883 June 1 - July 12, 1997 R. Musser, Senior Resident Inspector K. Poertner, Resident Inspector P. Byron, Resident Inspector J. York, Reactor Inspector (Sections El.l, El.3, and E2.1)
P. Fillion, Reactor Inspector (Sections El.2, and E8.1)
R. Franovich, Resident Inspector, Catawba, (Sections 08.3, 08.4, and M8.2)
G. Belisle, Chief, Reactor Projects Branch 5 Division of Reactor Projects Enclosure 9708220016 970807 PDR ADOCK 05000280 G
EXECUTIVE SUMMARY Surry Power Station, Units 1 & 2 NRC Inspection Report Nos. 50-280/97-06, 50-281/97-06 This integrated inspection included aspects of licensee operations, engineering, maintenance, and plant support. The report covers a 6-week period of resident inspection: in addition, it includes the results of announced inspections by regional reactor inspector Operations
NRC Form 3s were posted where required (Section 01.2).
- Operator actions to identify and isolate a leak in the letdown line downstream of the orifice block valves were conducted in a methodical manne The decision to remove the C letdown orifice from service unless required to support plant operation was an appropriate decision by licensee management (Section 01.3).
- Activities associated with a Unit 2 power reduction were conducted in accordance with approved procedure Command and control during the power reduction and return to full power were excellent (Section 01.4).
Maintenance
The assembly and installation of the lA circulating water pump was performed in a professional and cautious manne The mechanics were knowledgeable and appeared to anticipate issues (Section Ml.1).
- Operators performing emergency diesel generator testing were knowledgeable and followed procedures. The obtained data met procedural requirements (Section Ml.2).
Engineering
A review of the engineering backlog revealed that the number of items appeared reasonable and consistent (Section El.l).
- Review of an operational event, that had occurred at another utility, revealed that the licensee was properly addressing the event (Section El.2).
- The licensee properly concluded in their root cause evaluation that the major contributors to continuing letdown line weld failures were degraded orifices combined with a decreased operating back pressure that produced cavitation and a resulting vibration which caused fatigue failure of the socket weld However, in the root cause determination for one of the failed welds it was stated erroneously that poor welding was a major contributor to this weld's failure (Section El.3).
- The 'inspectors concluded from a review of the Design Basis Document (DBD) program that documentation and resolution of problems identified
- during development of the DBDs were in accordance with 10 CFR 50, Appendix B, Criterion XVI, Corrective Action (Section E2.1).
- The licensee had acceptable fuse control and software control procedures (Section E2.1).
- At the time of the review, *the number of Temporary Modifications (TMs)
in effect remains relatively low with three per unit. The inspectors reviewed the safety evaluations for the TMs in place and concluded that they adequately addressed the safety issues (Section E2.2).
Plant Support
Health physics practices observed were conducted properly (Section Rl).
- Security and material condition of the protected area perimeter barrier were acceptable (Section Sl).
- Report Details Summary of Plant Status Unit 1 operated at power the entire reporting perio Unit 2 operated at power the entire reporting perio On July 2, 1997, a power reduction was initiated due to an inadvertent closure of a turbine intercept valve (See Section 01.4). Power was reduced to 72 percent, the valve was reopened and the unit returned to 100 percent power the same da I. Operations
Conduct of Operations 01.1 General Comments (71707. 40500)
The inspectors conducted frequent control room tours to verify proper staffing, operator attentiveness, and adherence to approved procedure The inspectors attended daily plant status meetings to maintain awareness of overall facility operations and reviewed operator logs to verify operational safety and compliance with Technical Specifications (TSs).
Instrumentation and safety system lineups were periodically reviewed from control room indications to assess operability. Frequent plant tours were conducted to observe equipment status and housekeepin Deviation Reports (DRs) were reviewed to assure that potential safety concerns were properly reported and resolved. The inspectors found that daily operations were generally conducted in accordance with regulatory requirements and plant procedure,
01.2 Posting of Code of Federal Regulations (CFR) Required Notices a. Inspection Scope (71707)
The inspectors reviewed designated areas to ensure that the notices to workers required by 10 CFR 19.11 were poste b. Observations and Findings 10 CFR 19.11 specifies documents that are required to be posted so that workers are informed. Section 6.5.1 of VPAP-2802, "Licensing,"
Revision 6, incorporates the requirements of 10 CFR 19.11. The postings at Surry Power Station are located at the following locations:
Administration Building Clean Change Room Machine Shop Radwaste Facility Secondary Access Area Service Building The inspectors verified that the specified areas contained the required postings with the exception of the Secondary Access Area which is close *
c. Conclusions The inspectors verified that NRC Form 3s were posted where require.3 Unit 2 Letdown Line Weld Failure Inspection Scope (71707)
The inspectors reviewed licensee actions associated with a Unit 2 throughwall weld leak on the letdown line piping inside containmen b. Observations and Findings At 7:00 p.m., on June 24, the Unit 2 containment particulate radiation monitor alert annunciator alarm was received in the control roo Review of the chart recorder determined that particulate activity had increased by approximately_ 1600 counts per minute since the previous shift. Based on the increased activity, a containment air sample and a Reactor Coolant System (RCS) leak rate calculation was initiated by the operating crew. The containment air_ sample did not indicate above normal: however, the leakage calculation indicated that total leakage and unidentified leakage had increased from the previous steady state values. Total RCS leakage increased from 0.276 gpm to 0.600 gpm and unidentified leakage increased from 0.102 gpm to 0.438 gp A containment entry was made at 12:18 a.m., on June ~5. to try to identify the source of the increased RCS leakage. The major emphasis of the entry team was the letdown line piping downstream of the orifice block valves due to a history of weld failures at this location. The containment entry identified a throughwall leak immediately downstream of valve 2-CH-HCV-2200C, the C letdown orifice block valve. Excess letdown was placed in service and normal letdown was secured to isolate the lea Subsequent to isolating the letdown line and removing the lagging for repair, the licensee determined that the leak initiated at the socket weld at the outlet of the block valve. The licensee plans to replace the block valves *at the next refueling outage with butt welded valve The licensee also plans to replace the letdown orifices during the refueling outage to reduce vibration levels in the letdown line pipin The weld was repaired and the letdown line was returned to service on June 27, at 10:30 p.m. Subsequent vibration monitoring determined that operation with the A and C letdown orifices in service was acceptable:
however, vibration levels with the C orifice in service was significantly higher than with just the A orifice in service. Based on the significantly higher vibration levels with the A and C orifices in service, the licensee decided to operate with the A orifice in service and the C orifice isolated. The C orifice would be placed in service if
- require Letdown line weld failures are discussed further in Section E c. Conclusions Operator actions to identify and isolate a leak in the letdown line downstream of the orifice block valves were conducted in a methodical manne The decision to remove the C orifice from service unless required to support plant operation was an appropriate decision by licensee managemen.4 Unit 2 Power Reduction Inspection Scope (71707)
The inspectors monitored activities in the control room and at the main turbine during a Unit 2 power reduction to 72 percent power after the C intercept valve isolated while the unit was operating at 100 percent powe b. Observations and Findings On July 2, at 9:43 a.m., Unit 2 experienced a 10 MWe load decrease and the control room operator noted that the C intercept valve was drifting closed. At 9:49 a.m., a unit ramp was commenced due to the closure of the intercept valve. During the power reduction, maintenance personnel were mobilized to investigate the cause of the valve closur Investigation determined that the associated Electro Hydraulic Control (EHC) dump valve was leaking past the seat resulting in a loss of EHC control pressure to the valve. The power reduction was stopped at 72 percent power and the intercept valve was stroked using the test circuit. The intercept valve opened as designed when the test button was released. At 12:26 p.m., a power increase was commenced to return the unit to 100 percent powe The unit achieved full power at 6:00 p.m. that same da The inspectors observed activities in the control room and at the main turbine. Control room activities were accomplished in accordance with approved procedures and the operating crew discussed expected plant response throughout the power reduction and attempt to reopen the intercept valve.* Command and control during the power reduction and return to full power were excellen c. Conclusions Activities associated with the Unit 2 power reduction were conducted in accordance with approved procedure Command and control during the power reduction and return to full power were excellent.
- 06.1 Significant Personnel Changes at Surry (71707)
During this inspection period, a number of personnel changes were announced at Surry Nuclear Power Station and are as follows:
T. Sowers, Superintendent-Engineering, assumed the position of Superintendent-Nuclear Training. This change was effective July *
M. Adams, Supervisor-Engineering (North Anna Power Station),
assumed the position of Superintendent-Engineering. This change was effective July *
D. Llewellyn, Superintendent-Training, will be attending SRO license clas Miscellaneous Operations Issues (92700, 92901)
08.1 Turbine Driven Auxiliary Feedwater Pumps (TDAFWPs):
Region II sent information to the inspectors regarding problems with TDAFWPs at two nuclear plqnts. The problems related to the TDAFWP turbines tripping on overspeed due to entrapped water in the turbine casing. The steam traps in the steam line to the turbine were piped to the floor drains in the TDAFWP room The condensate from the steam traps flowed to the floor drains and for a variety of problems the floor drains were unable to accommodate the condensation. The floor drains backed up causing the steam traps to backup. This resulted in the steam line condensation flowing to the turbine rather than being diverted to the drain syste Water (condensation) in the turbine casing resulted in a water slug causing the turbines to trip on overspee The inspectors were requested to review the licensee's TDAFWPs to determine if a similar condition existed. The inspectors reviewed drawings for the TDAFWP piping and held discussions with the system engineer. A similar condition does not exist at Surry as the steam traps in the TDAFWP steam line are piped to the condenser. This configuration eliminates the possibility of the described issue from occurring at this sit.2 (Closed) Licensee Event Report CLER) 50-281/95007-00: reactor trip due to failed reactor coolant pump moto On November 7, 1995, Unit 2 was operating at 100 percent powe At 9:56 a.m., a reactor coolant low flow signal on one out of three loops caused. a reactor trip. The low flow signal was caused by the automatic opening of Reactor Coolant Pump (RCP) Breaker 25C3 due to a phase to ground fault in the C RCP moto Appropriate operator actions were taken and the unit was quickly brought to a stable no load condition.
The licensee determined during their post trip review that the RCP C motor stator windings were grounded. The following deficiencies also occurred during the transient:
The Moisture Separator Reheater (MSR) Control System did not reset as require *
The Condensate Polishing Building Bypass Valve (2-CP-AOV-222) did not close when attempting to place the Condensate Polishing Building in service in accordance with Emergency Operating Procedure 2-ES-0.1, "Reactor Trip Response."
- One of the main steam dump valves (2-MS-TCV-206A) did not automatically ope The licensee determined that a defective relay caused the MSR not to reset and failed solenoids caused the other two problems. The MSR relay was replaced under Work Order (WO) 329692-01. The solenoid in 2-CP-AOV-222 was replaced under WO 329649-01 and the 2-MS-TCV-206A solenoid was replaced under WO 00330040-03. The RCP C motor was replaced and the damaged motor was sent to the vendor to be repaire On November 22, 1995, at 6:37 a.m., Unit 2 was returned to the gri The inspectors reviewed Reactor Trip Report S2-ll-07-95, Root Cause Evaluation (RCE) 95-16, "Unit 2 Trip Due to an* RCS Low Flow Signal on One of Three Loops, " and the three WOs 1 i sted above;. The inspectors verified by WO review that the repairs had been completed. The licensee's investigation appeared to be thorough and corrective actions were adequat.3 (Closed) LER 50-280/95001-01: Unit 1 automatic reactor trip due,to coupling failure on main feed pump (MFWP).
On January 8, 1995, Unit 1 was operating at 100 percent reactor power when MFWP B tripped on low *
lube oil pressur Two minutes later, a Steam Generator (SG) low-low water level signal caused Unit 1 to trip from 74 percent reactor powe On February 4, 1995, the licensee submitted LER 50-280/95001-00 to inform the NRC of the reactor trip and provide associated required informatio On August 24, 1995, the licensee provided additional information to the NRC and committed to additional corrective actions in Supplement 1 to the LE NRC Inspection Report Nos. 50-280, 281/97004 documents the closure of the initial LE In Supplement 1 of LER 50-280/95001, the licensee identified actions to prevent recurrence in addition to those delineated in the initial LE Those actions included the following: (1) replace the MFWP lube oil system teflon coupling with a metal coupling connected to an insulated return line, (2) upgrade post maintenance testing requirements to verify TDAFWP governor stability following maintenance on the governor control system, (3) revise TDAFWP procedures to include more detailed information on the governor and governor linkage field setup, (4) modify the TDAFWP.governor linkage to ensure vertical alignment, (5) replace
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the TDAFWP governor connecting rod with a solid control rod, (6) permanently retire Governor 227, since the root cause of the diverging steam flow oscillations as the pump ramped to full flow could not be determined, and (7) provide an update to the results of the initial RCE distributed on the INPO Nuclear Networ The inspectors reviewed Trip Report Sl-01-08-95 and associated RCE 95-01; reviewed WO documentation to verify that MFWP lube oil p1p1ng and teflon couplings had been replaced with flexible hoses and metal connectors; inspected all MFWPs to verify that modifications to the lube oil system had been completed; reviewed RCE 95-02 associated with the TDAFWP divergent oscillations; reviewed Design Change Package 95-039-03 and the associated safety evaluation for modification of the TDAFWP turbine governor valve; reviewed procedure O-MCM-1403-01, "Terry Turbine Overhaul, 1-FW-T-2 and 2-FW-T-2," Revision 8, to verify that appropriate procedure changes had been made to reflect Unit 1 changes and provide for more aggressive post-maintenance testing; inspected Unit 1 TDAFWP turbine governor valve to verify that the governor linkage and connecting rod had been modified; verified that modification to the Unit 2 TDAFWP turbine governor valve is planned for the September 1997 refueling outage and was approved as WO 00310109-01; verified that Governor 227 was no longer installed in the plant and that measures were in place to prevent installation in the future; and reviewed an update to the initial INPO Nuclear Network notic (Closed) LER 50-280/95010: four inoperable component cooling heat exchangers due to macrofoulin On October 6, 1995, the licensee determined that Component Cooling Heat Exchanger (CCHX) C was inoperable while Unit 1 was at cold shutdown and Unit 2 was operating at 100 percent reactor powe The heat exchanger was cleaned, satisfactorily tested and returned to servic On October 7, the licensee determined that CCHXs A, Band D also were inoperable. With only 1 CCHX operable, Unit 2 entered TS Action Statement 3.13.B, which required restoration of a second CCHX within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or that the unit be placed in hot shutdown within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. A second CCHX was returned to service within the time allowed by T The licensee's RCE revealed that the CCHX Service Water (SW) tubesheet had been blocked by hydroids and other marine material from High Level Intake Structure (HLIS) Band the B-Train 96-inch Circulating Water (CW)
piping. The hydroids had been present on the walls of the HLIS. but apparently had become dormant when conditions exposed them to air or stagnated wate Four hours before indications of CCHX blockage were evident, the CCHX SW supply was swapped from HLIS D to HLIS B, its alternate source. Condenser B was placed in_ service when the swap from HLIS D to HLIS B was completed. The turbulence created as a result of changes in water velocity when Main Condenser B was returned to service caused the pre-existing hydroids to break off and migrate toward the condenser; some of the hydroids entered the SW supply line.
Based on the results of the licensee's RCE the procedures used to return Main Condenser Band D to service were modified to reduce the impact of
hydroids on the CCHXs following maintenance on the associated HLIS The inspectors reviewed Commitment Tracking System (CTS) Item 3214, which was used to track the commitment to change procedures, to verify that the CTS had been closed. The inspectors verified that changes to procedure 1-MOP-CW-004, "Return to Service of Waterbox B," were incorporated into Revision 4 and that changes to 1-MOP-CW-008, "Return to Service of Waterbox D," were incorporated into Revision 3. The changes provided instructions to require that, following the completion of maintenance at the HLIS with the potential to cause introduction of marine debris into the subject intake bays, actions be performed to (1)
flow each waterbox for a minimum of 10 minutes before opening the associated CCHX SW supply valve, and (2) isolate at least one CCHX before returning the waterbox to service, following the removal of the associate stop log.5 (Closed) LER 50-281/96004: turbine/reactor trip due to high level in the steam generator. This LER describes a Unit 2 reactor trip from 16 percent power during power escalation following completion of a scheduled refueling outage. The reactor trip resulted from SG level oscillations at low power levels. This event was discussed in detail in NRC Inspection Report Nos. 50-280, 281/96005. The inspectors determined that the post trip review evaluated the trip and prerequisites for restart in appropriate detail.
08.6 (Closed) LER 50-281/96005: manual reactor trip due to loss of electro hydraulic control pressure. This LER describes a Unit 2 manual reactor trip from 88 percent power due to spurious closure of the turbine governor valve The turbine governor valves closed as a result of a failed mechanical fitting joint in the EHC system. This event was discussed in detail in NRC Inspection Report Nos. 50-280, 281/9600 The inspectors determined that the post trip review evaluated the trip and prerequisites for restart in appropriate detai II. Maintenance Ml Conduct of Maintenance Ml.l Installation of Unit 1 Circulating Water (CW) Pump (1-CW-P-lA) Inspection Scope (62707)
The inspectors observed the assembly of the Unit 1 CW lA pump with its bowl and the lowering of the pump assembly into the pump pi b. Observations and Findings On June 13, 1997, the inspectors observed the assembly of the Unit 1 CW lA pump with its bow The pump was placed in position with the assistance of a cran WO 00356700-01 was the controlling document for the removal of the pump, replacement of the mechanical seal, and the replacement of the pump assembly in its pump pit. The inspectors observed t~e installation of RTV on the pump deck plate to aid in the
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sealing of the pump flange to the deck plate. The lift and transport to the pit went smoothl The licensee experienced a little difficulty aligning the pump with its attachment studs. The mechanics were able to align the pump with the assistance of two come-along The mechanics were cautious and methodical. Crane operations were controlled by hand signals. The inspectors noted that the mechanics did not use radios to control crane operations but used relayed signals when the pump was placed in the pit. The work was performed in a professional manner and all cautions were taken. Procedures were not evident as the work was performed by skill of the craf c. Conclusions The inspectors considered that the assembly and installation of the lA CW pump was performed in a professional and cautious manne The mechanics were knowledgeable and appeared to anticipate issue Ml.2 Emergency Diesel Generator (EDG) No. 1 Monthly Start Test Inspection Scope (61726)
The inspectors observed portions of 1-0PT-EG-001, "No. 1 EDG Monthly Start Exercise Test," Revision 8, and other related tests. The completed test procedures were reviewed.
b. Observation and Findings On June 9, 1997, the inspectors observed portions of the "No. 1 EDG Monthly Start Exercise Test" which was performed in accordance with Periodic Test (PT) 1-0PT-EG-001, Revision 8. The licensee also concurrently performed the monthly fuel oil transfer pumps test in accordance with Section 6.2 of l-OPT-EG-005, "No. 1 EDG Fuel Oil System Tests," Revision 4, and also sampled the lube oil and cooling water in accordance with Sections 6.2 and 6.3 of l-OSP-EG-6.1, "No. 1 EDG Lube Oil and Cooling Water Sampling," Revision 3. The inspectors noted that procedures were at the job site and were in use by the operator The inspectors reviewed the data in the above three procedures and in addition reviewed the data in Surveillance Procedure (SP) 1-0SP-EG-6.2,
"No. 1 EDG Starting Air System Quarterly Test," Revision 3. All of the data was within procedural requirement c. Conclusions The inspectors concluded that the operators were knowledgeable and followed procedures. The obtained data met procedural requirement M8 Miscellaneous Maintenance Issues (92902)
M8.l (Closed) Violation (VIO) 50-280/95003-02: failure to use approved detailed procedure On December 24 and 25, 1994, and January 10 and 11. 1995, maintenance was performed on the Unit 1 TDAFWP
governor. The inspectors determined that the work was performed by vendor representatives and licensee mechanics without approved detailed procedures. The licensee stated that procedures were not required as the work was performed by the vendor's representative who was knowledgeable of the equipment and licensee mechanics who utilized skill of the craft. VPAP-0801. "Maintenance Program," Revision 4, requires that maintenance activities performed by a vendor be accomplished in accordance with approved procedures and maintenance activities be reviewed to determine if it required approved procedures or work could be performed by skill of the craft. The licensee determined that similar maintenance activities had been performed several times and the work could be performed by skill of the craft. However, problems arose during the maintenance on the governor and the work became more complex than p 1 anne The mechanics continued to work by sk_i 11 of the craft when in fact procedures were neede The licensee initiated training for both management and the craft on the need for procedures as corrective actions. The inspectors reviewed the training outlines and considered that they addressed the issues. The training was given during the required continuing training which gives assurance that all maintenance personnel received the trainin *
MB.2 (Closed) LER 50-280/95007-01: operation with non-isolable leak in pressurizer instrumentation nozzle On September 12, 1995, Unit 1 was at cold shutdown for a scheduled refueling outage. Visual inspection of the pressurizer revealed a leak in two of the upper pressurizer instrument line nozzle One of the corrective actions was to conduct a metallurgical examination of one of the nozzles and report the results of the examination in a supplement to the original LE On January 10, 1996, the licensee received Nuclear Engineering and Services Materials Engineering Laboratory Failure Analysis Report. The report documented the root cause of the metallurgical examination, which attributed the nozzle cracks to transgranular stress corrosion cracking and intergranular stress corrosion crackin The licensee committed to perform additional visual inspections of the pressurizer instrument nozzles during each unit's subsequent refueling outage Upon completion of these inspections, the licensee would determine an appropriate frequency for the performance of future inspection The licensee generated CTS Item 3364 to ensure that the inspection would be performed in accordance with the committed action to prevent recurrence. The inspectors reviewed CTS 3364, which documented the inspection results of the Unit 1 pressurizer instrument taps. The inspection was performed during the Unit 1 refueling outage using a modified VT-2 with the pressurizer depressurize No evidence of leakage was detected. Similar inspections of the Unit 2 pressurizer instrument taps, completed May 20, 1996, have also revealed no leakag Since other methods of detecting RCS leakage (daily leakage rate calculations, radiation monitors inside containment, and periodic
- reactor coolant pressure boundary testing) were available and being performed, the licensee decided that additional visual inspections of the pressurizer instrument taps would be conducted during future refueling outages in accordance with the Inservice Inspection and Non-destructive Examination (ISI/NDE) Augmented Inspection Manua The inspectors concluded that this inspection *frequency was adequat III. Engineering El Conduct of Engineering El.1 Engineering Backlog (37550)
The inspectors discussed the various areas of engineering backlog with engineering management. Charts showing various levels of backlog for individual months were discussed for the following:
Requests for Engineering Assistance (REAs), Design Change Packages under development, Design Change Package Inventory, Drawing Inventory, WOs awaiting engineering disposition, DR Status, and Commitment Tracking histor The number of open items appeared to be maintained at a consistent level. The several anomalies that were observed to occur at various times were explained by the license El.2 Operating Experience Review (37550)
The inspectors discussed a problem that had occurred at another nuclear plant with the cognizant reactor engineer. While reducing reactor power for turbine valve testing; it was noted that the plot of the Over Power-Delta Temperature set points were slightly increasing instead of remaining at a constant value. A discussion was held with the
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reactor/system engineer that was reviewing this operating experience for Surry Nuclear Plant. The engineer explained how these instrument/set points were calibrated using procedure 1-IPT-CC-RC-T-432, "Delta T and Tavg Protection Set III Loop T-432 Channel Calibration," Revision 1 Reviewing diagrams and instrument data sheets revealed that the summator TM-1-432M was checked for a constant output of 1.000 volts DC between a power of Oto 100 percent with a Tavg between 547 and 573 degrees. With this value there should be no change in the set point of 106.9 power percent if Tavg decreases. Conservatively the engineer stated that the item would not be closed until the exact details were obtained from the other facilit El.3 Letdown Line Weld Failures Inspection Scope (37550)
Seven failures have occurred on Unit 2 letdown line socket welds since 1987, with six of these failures occurring recently (December 13, 1995 through June 25, 1997). The inspectors reviewed the engineering *
analyses for resolving these problems, e.g., Justification for Continued Operation (JCO), root cause analyses, et ~ b. Observations and Findings The inspectors reviewed the following licensee documents for analyzing the methods used for resolving the weld failures on the letdown line:
Engineering Transmittal No. S 97-0262, "Review of Letdown System Vibration Data Surry Power Station, Unit 2."
- Materials Engineering Laboratory Report NESML-0-282 concerning the failure of a weld on the Unit 2 letdown lin *
JCO No. S2-97-001, "Unit 2 Letdown Operation" (concerns using one*
letdown orifice instead of the normal two).
In the root cause document it was noted that the failures were the result of vibration due to degraded orifices or orifices and isolation valves, extended operation at low orifice back pressure (which causes high cycle vibration/fatigue when combined with the degraded orifices),
and possibly low cycle transient operational events (leakage of the downstream relief valve, thermal fatigue, etc.).
The category 1 RCE was performed after the four recent weld failures and was the only failure where a weld sample was removed for examination in the metallurgical laboratory. The inspectors' review of th metallurgical report (Report No. NESML-0-282) revealed that the licensee attached too much importance to the lack of fusion at the base of the socket wel The report stated that there was a significant amount of lack of fusion at the root of the wel Examination of the photomicrographs in the report showed an area of 0.012 inches that could be called lack of fusio The inspectors considered this to be a normal weld (consultations with another welding/metallurgical regional specialist resulted in the same conclusion), The inspectors agreed that the notch produced by the lack of fusion was the crack starter for the-fatigue failure when the vibration was present. Design information recognizes that a socket weld is not a desirable weld when the component/system is subject to vibration during operatio Additionally, the RCE erroneously concluded that one of the major contributors to the September 11, 1996, socket weld failure was poor weld quality. This erroneous conclusion on the failed weld concerning excessive lack of fusion was identified as a negative observation. The root cause also concluded that piping configuration and vibration were major factor Currently the licensee is operating Unit 2 (until a refueling outage later this year) with only one 45 GPM orifice (A orifice) in service but can use the C 60 GPM orifice if needed. Vibrational measurements on the lines lead the licensee to this decision. The B 60 GPM orifice operation leads to much higher vibration An X-ray of the B orifice by the licensee revealed extensive cavitation damage at the exit of this orifice and this would cause a vibration proble The inspectors
- discussed this operation with one less orifice than normal with the system engineer and it was concluded that the daily chemistry analysis would give an indication of a problem with the cleaning of the RC c. Conclusions The licensee properly concluded in their RCE that the major contributors to continuing letdown line weld failures were degraded orifices combined with a decreased operating back pressure that produced cavitation and a resulting vibration which caused fatigue failure of the socket weld However, in the root cause determination for one of the failed welds, it was stated erroneously that poor welding was a major contributor to this weld's failur E2 Engineering Support of Facilities and Equipment E2.1 Review of Design Basis Documents CDBDs) Inspection Scope (37550)
The inspectors reviewed the licensee's DBDs which were a feature of the Integrated Configuration Management Project. The review focused on understanding the scope of work completed and schedule of remaining work on the DBDs as well as problem identification* and resolution. The inspectors conducted discussions with the manager of the Integrated Configuration Management Program and engineers responsible for the DBD As examples of the DBD program, the inspectors reviewed in detail the Electrical System and Service Water System DBDs. Also, the inspectors reviewed the executive summary, or Information Assessment Report, for the Electrical System, Auxiliary Feedwater System and Safety Injection DBD The inspectors reviewed the Open Items List and Potential Problem Report Summary associated with the Electrical System DB A secondary objective of the inspection was to confirm certain statements made in*
the licensee's letter to the NRC dated February 7, 1997, on the subject of: "Response to Request for Information Pursuant to 10 CFR 50:54(f)
Regarding Adequacy and Availability of Design Bases Information," and to thereby determine the need for future inspections in this area. Other related topics inspected were:
Fuse control program and design controls on softwar *
b. Observations and Findings The licensee plans to have the total integrated review for all the safety-related systems completed by October 1998. A total integrated review means that a DBD for a system has been completed and that the DBD has been compared to all the relevant operations and maintenance procedures. The plant DBDs which assemble the DBDs related to selected design topics which cut across all systems such as single failure criterion would be completed after October 199 The inspectors observed that in general the.DBDs were carefully prepared. _The DBDs were sufficiently distributed throughout the
organization, and provide an excellent source document for understanding the design basis. Significant deficiencies identified in the process of creating the DBDs were resolved in a timely manne Open items generated by the DBD effort were of lesser significance and were being tracked to completion. The open items received regular reviews by management to ensure sufficient progress in their resolutio In the Electrical Power DBD, the inspectors identified two errors. One was that the minimum start and run voltages for 480 V loads given in Table 6.1-1 were reverse The second was that the text of Section 7.2.l, second paragraph refers to a change on Unit 1 "described above,"
however, the section does not contain the referenced information. While the errors are essentially typographical in nature, they could indicate a lack of thoroughness in reviewing the material. *
The inspectors reviewed the procedures for the fuse control program and the software control program. The fuse control program was discussed with the Supervisor of Electrical Engineering and the engineer responsible for much of the work done in the area of fuse control. The inspectors thought it significant that the program covers critical safety-related fuses and any fuse contained in a coordination stud Fuses on the list in procedures 1-DRP-002 and 015, "Instrument and Power Fuse Schedule," Revisions 6 and 1 respectively, had been verified by the licensee in a walkdown inspection. The inspectors observed by reviewing the controlling procedure, VPAP-0306, "Station Software Control,"
Revision 5, that software is controlled and validated using a detailed
. procedure. The inspectors concluded that the fuse control and software control procedures were credible program c. Conclusions The inspectors concluded from a review of the DBD program that documentation and resolution of problems identified during development of the DBDs were in accordance with 10 CFR 50, Appendix B, Criterion XVI, Corrective Actio The licensee had acceptable fuse control and software control procedure E2.2 Temporary Modifications (TMs) (37551)
On July 10, 1997, there were three open TMs for each unit. The first Unit 1 TM (Sl-97-07) reviewed combines two previously separate inputs into one annunciator panel window (1-RI-ANN-IF-H3).
TM Sl-97-09 added temporary discharge lines to two relief valves. This TM was scheduled to be closed out on July 14. The third TM, Sl-97-11, lifted two grounded wires from connections 446 and 447 in the turbine first stage pressure loops. The Unit 2 pressurizer safety valve (2-RC-SV-2551C)
lea~ed by which resulted in pressurizing the Pressurizer Relief Tank (PRT).
_TM S2-96-27 was issued to vent the PRT gas to the Overhead Gas System through the sample sin TM S2-96-30 was issued to install temporary ~pray shields around RCP 2-RC-P-lC. The problems with the
letdown lines which are described in more detail in Section E resulted in reduced charging flo TM S2-97-05, Revision 1, was issued to lower the "Charging Lo Flow" alarm set point which reduced annunciator alarms. The licensee plans to closeout the Unit 2 TMs during the October 1997 refueling outag The inspectors reviewed the safety evaluations for each of the TMs and concluded that they adequately addressed the safety issues. The inspectors also reviewed the WOs which are in place to closeout the TM TM S2-97-05 will be closed out upon the closure of DCP 96-40 which replaces the Unit 2 letdown orifices, piping and valve ES Miscellaneous Engineering Issues (92903)
E8.l (Closed) VIO 50-280/96009-03: Design change package failed to specify breaker set point. This VIO involved a situation where new motor control centers were installed as part of a plant modification to upgrade the Heating, Ventilation, and Air Conditioning CHVAC) in the electrical equipment room The new motor control centers had circuit breakers with adjustable instantaneous (magnetic) trip set points. The balance of the breakers in the plant did not have this adjustment. This difference caused the licensee to not specify a setting in the modification documentation. The actual set points for two or three breakers were wrong which resulted in undesired tripping during plant evolutions. The licensee's problem report process eventually identified the wrong set point When the NRC inspectors reviewed these problem reports, they concluded that the root cause, which was failure to specify set points in the modification documentation, had not been addressed by corrective actions stated in the problem reports. The licensee, in their response to the VIO, stated they would provide training to the appropriate individuals concerning this problem. The inspectors confirmed through review of documentation that adequate training was conducte In addition, the inspectors confirmed through review of completed work requests that the set points for the breakers in question had been corrected Cit was not possible to see the set point without removing the circuit breaker from the motor control center). In summary, the inspectors confirmed that the corrective actions stated in the response to the VIO had been complete *
IV. Plant Support Rl Radiological Protection (RP) and Chemistry Controls (71750)
On numerous occasions during the inspection period, the inspectors reviewed RP practices including radiation control area entry and exit, survey results, and radiological area material condition No discrepancies were noted, and the inspectors determined that RP practices were prope *
Sl
Conduct of Security and Safeguards Activities (71750)
On numerous occasions during the inspection period, the inspectors performed walkdowns of the protected area perimeter to assess security and general barrier condition No deficiencies were noted and the inspectors concluded that security posts were properly manned and that the perimeter barrier's material condition was properly maintaine V. Management Meetings Xl Exit Meeting SUD1Dary The inspectors presented the inspection results to members of licensee management at the conclusion of the inspection on July 21, 1997.. The licensee acknowledged the findings presente The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietar No proprietary information was identifie PARTIAL LIST OF PERSONS CONTACTED Licensee M. Adams, Superintendent, Engineering (effective 7/1/97)
R. Allen, Superintendent, Maintenance R. Blount, Assistant Station Manager, Nuclear Safety and Licensing D. Christian, Station Manager M. Crist, Superintendent, Operations J. *McCarthy, Acting Station Manager B. Shriver, Assistant Station Manager, Operations & Maintenance T.. Sowers, Superintendent, Engineering (Superintendent Training-effective 7 /1/97)
B. Stanley, Director, Nuclear Oversight W. Thorton, Superintendent, Radiological Protection NRC G. Belisle, Chief, Branch 5, Division of Reactor Projects, Region II G. Edison, Surry Project Manager, Office of Nuclear Reactor Regulation
INSPECTION PROCEDURES USED IP 37550:
Engineering IP 37551:
Onsite Engineering IP 40500:
Effectiveness of Licensee Controls in Identifying, Resolving, and Preventing Problems IP 61726:
Surveillance Observation IP 62707:
Maintenance Observation IP 71707:
Plant Operations IP 71750:
Plant Support Activities IP 92700:
Onsite Followup of Written Reports of Nonroutine Events at Power Reactor Facilities IP 92901:
Followup - Plant Operations IP 92902:
Followup - Maintenance IP 92903:
Followup - Engineering ITEMS CLOSED Closed 50-281/95007-00 50-280/95001-01 50-280/95010 50-281/96004 50-281/96005 50-280/95003-02 50-280/95007-01 50-280/96009-03 LER LER LER LER LER VIO LER VIO reactor trip due to failed reactor
. coolant pump motor (Section 08.2).
unit 1 automatic reactor trip due to coupling failure on main feed pump (Section 08.3).
four inoperable component cooling heat exchangers due to macrofouling (Section 08.4).
turbine/reactor trip due to high level in the steam generator (Section 08.5).
manual reactor trip due to loss of electro hydraulic control pressure (Section 08.6).
failure to use approved detailed procedures (Section M8.1).
operation with non-isolable leak in pressurizer instrumentation nozzles (Section M8.2).
design change package failed to specify breaker set point (Section E8.1).