IR 05000423/1986035
| ML20210A233 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 01/16/1987 |
| From: | Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20210A209 | List: |
| References | |
| 50-423-86-35, NUDOCS 8702060436 | |
| Download: ML20210A233 (12) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report N /86-35 Docket N License N NPF-49 Licensee: Northeast Nuclear Energy Company P.O. Box 270 Hartford, CT 06101-0270 Facility Name: Millstone Nuclear Power Station, Unit 3 Inspecticn At: Waterford, Connecticut Inspection Conducted: November 18, 1986 - January 5,1987 Inspectors: J. T. Shedlosky, Senior Resident Inspector, Millstone 3 T. A. Rebelowski, Senior Resident Inspector, Millstone 1 and 2 F. A. Casella, Resident Inspector E. L. Conner, Pro ect Engineer Approved by: N4 i fl6[f7 E. C. McCabe, Chief, Reactor Projects Section 3B Date Inspection Summary:
Areas Inspected: Routine on-site resident inspection (154 hours0.00178 days <br />0.0428 hours <br />2.546296e-4 weeks <br />5.8597e-5 months <br />) of shutdown planning, plant operations, radiation protection, physical security, fire protec-tion, surveillance and maintenanc Results: This inspection identified satisfactory performance in all areas.
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8702060436 870127 PDR ADOCK 05000423 ( 0 PDR L
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TABLE OF CONTENTS P. age Summary of Facility Activities....................................... 1 Review of Specific Activities........................................ 1 Steam Generator Feedwater Flow Transients....................... 1 Reactor Coolant System Loop Flow Rates.......................... 2 Reactor Coolant System Flow Indication Deficiency............... 3 RHS Heat Exchanger Seismic Support Deficiency................... 4 Modification to Containment Gaseous and Particulate Activity Monitor......................................................... 4 Radioactive Liquid and Gaseous Effluent Monitoring by Analog Channels........................................................ 5 Licensee Event Reports............................................... 5 Drug Abuse Allegation RI-86-A-139.................................... 6 Review of IE Bulletin 86-03, Potential Failure of Multiple ECCS Pumps Due to the Single Failure of an Air Operated Minimum Flow Valve...... 8 Review of IE Information Notice 85-45, Potential Seismic Interaction Involving the Movable In-Core Flux Mapping System Used in Westinghouse Designed P1 ants...................................................... 8 Observation of Maintenance........................................... 9 Observation of Surveillance Testing.................................. 9 Safety Committee Meetings............................................ 9 10. Management Meetings.................................................. 10 i
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DETAILS 1. Summary of Facility Activities The plant operated at full power except for brief periods for surveillance testing and/or preventive maintenance at reduced power. The unit has operated since December 21 with the "B" Steam Generator Feedwater Regulating Valve in manual. The "B" Feedwater Regulating Valve Bypass Valve is controlling i automatic. This action was taken to stop valve motion and reduce a leak through worn packing. Prior to this action, the "B" Regulating Valve had not been operating smoothly due to the packing proble . Review of Specific Activities Although the performance of plant operators and equipment was as expected during the inspection period, several problems occurred and are summarized below: Steam Generator Feedwater Flow Transients The licensee has continued to monitor the performance of the steam generator feedwater system and evaluate any abnormal conditions in an effort to reach stability under all flow conditions. During this in-spection, several transients were initiated by the automatic trip of a fourth point heater drain pump due to low heater shell leve The first transient occurred on November 30 at 1855 when an air line associated with the "2B" heater normal level control valve broke and resulted in the valve failing closed. The loss of water to the cascad-ing heater drains resulted in a trip of the heater drain pum The operators reduced power to 40 MW(e) generator output below full power and stabilized steam generator level Feedwater flow oscillations were experienced in the "C" and "D" feedwater lines at 2300, December 3 while at 90% power. These were of an approxi-mate magnitude of 0.6 E + 6 pounds per hour above and below the normal 3.5 E + 6 pounds per hour. The control room operators placed the two i steam generators in manual level control and placed the turbine driven feedwater pumps in manual speed control to stabilize these oscillations.
! The licensee suspects the cause to be a system interaction occurring when a Circulating Water Pump was started and changed condenser vacuum.
- Other instances of feedwater oscillations accompanied trips of a fourth
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point heater drain pum These have been associated with a malfunction-ing level control system associated with the "B" first point normal level I
control valve. The pump trips have not recurred since control setpoint
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Another feed system oscillation was induced by removing a second conden- l sate demineralizer from service on December 19. The remaining demineral-izers created a sufficient differential pressure to create instabilities i in the feedwater system. This was corrected by opening the demineralizer bypass valve enough to lower the differential pressur The final feedwater problem during this period occurred on December 15 i due to worn packing in the "B" Feedwater Regulating Valve. This caused l leakage problems along with valve shaft binding. The condition was !
stabilized on December 21 when the Feedwater Rcgulating Valve was placed in manual control and steam generator level was maintained using the "B" Feedwater Regulating Valve Bypass Valve in automatic. The plant operated in this manner through the end of this inspection period. Control Room i operators have normal remote manual position control of the valve and the automatic Feedwater Isolation function is maintained. Since valve motion has stopped while in the manual mode, the packing leakage can be minimized with adjustments. The licensee intends to investigate the normal flow control system and attempt to reduce the amount of valve motion or " hunting" which has been observed during steady state plant operatio Feedwater control stability will be re-examined incident to routine in-spectio '
b. Reactor Coolant System Loop Flow Rates On December 16, 1986, while formulating an 18 month surveillance proce-dure for RCS loop flow rate determinations, a licensee engineer dis-
, covered that calculations performed under Start-Up Test 3-INT-8000,
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Appendix 8015, " Loop Flow Measurements" had been based on erroneous dat Temperature values used in these heat balances had been obtained using resistance conversion tables for RTDs different than the RTDs used. Each RTD has a unique conversion curv Raw data, in the form of resistance values, taken from the installed spare loop RTD's, were converted to temperature values using the in-service RTD conversion curve With the four (4) calculated loop flows in question, the low flow reactor trip set point values were in question for all four (4) loops. The
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problem was discovered at 8:30AM. The licensee entered Limiting Condition
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for Operation (LCO) 3.0.3 at 0945 and began preparations for a plant shutdown. Simultaneously, heat balance calculations were completed using the correct, spare RTD resistance to temperature conversion curve These calculations showed that the in-use flow values for loops 1, 3 and 4 were less than the recalculated values. The Technical Specification l Limiting Safety System Setting (LSSS), 2.2.1, Table 2.2-1.12 low flow
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trip bistables (3 per loop) trip setpoint is greater than or equal to 90% of the design loop flow of 94,600 GP The LSSS Instrument Allowable Value is greater than or equal to 89.3% of design flow. With the flow
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transmitter spans based on the previously calculated (lower) flowrates, the loop 1, 3, and 4 trip bistables were set more conservatively than the 90% of design flow setpoin The new calculation result for loop 2 was 102,099 GPM (107.9% of design)
vice 103,384 GPM (109.7% of design). With the loop 2 flow trip bistables set at 90% of design flow based on transmitter spans adjusted from 0 to 109.7%, the 1.8% reduction in measured flow meant that the trip setpoint was less conservative than the required Trip Setpoint. The loop 2 low flow reactor trip bistables were reset to 92% of design flow to compen-sate for the 1.8% difference in calculated flows. Plant shutdown acti-vities were then stopped since the plant was out of LCO 3. The inspectors reviewed the core safety implications and concluded that there were none in this case, even though the Limiting Safety System Setting 2.2.1 Instrumentation Allowable Value of greater than or equal to 89.3% of design flow was not met. If a loss of flow were to have oc-curred, the trip from loop 2 would have initiated when 88.2% of design flow was reached in that loop. This setpoint is 1.2% greater than the 87% flow trip assumed in the plant safety analysi The resident inspectors were promptly notified and kept informed about this matter. The NRC was officially notified via the ENS at 10:32 A Licensee Event Report (LER) 86-058-00 is to be submitted by the licensee to complete the reporting requirement Since loop flow transmitters are located within high radiation areas during reactor operations, the licensee's interim corrective action was to reset the trip bistabl Span adjustments based on the new calcula-tion will be completed during the planned March 1987 mid-cycle outag Resident inspector review of this matter noted licensee identification, low safety significance, and acceptable interim corrective action. No previous violation corrective actions that should have prevented this event were identified. This item is unresc1ved (86-35-01) pending cor-rective action completion and assessment of the adequacy of measures to prevent recurrenc c. Reactor Coolant System Flow Indication Deficiency Reactor engineering and control room personnel nottd an absence of fluc-tuations in indicated flow for Channel Number 1, Loop Number 4 in control room indicators. Instrument failure to respond was confirmed by observ-ing the data trend on the plant computer. With this deficiency confirmed in the output of instrument channel 3RCS-FT444, bistable 3RCS-FS444A was tripped at 1330, November 25, 1986. The channel will remain tripped until a containment entry is made and the instrument transmitter repaire .
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4 RHS Heat Exchanger Seismic Support Deficiency Two (2) bolts associated with a residual heat removal (RHS) heat exchan-ger seismic support were found broken during an engineering inspection of unrelated items nearby. This support consists of shim blocks and four lugs mounted at the "A" RHS heat exchanger mid plane. The shim blocks are captured by the support plate and also bolted in place with 3/8 inch cap screws. The licensee has analyzed this failure and concluded that a cap screw failure occurred when the heat exchanger lug moved vertically approximately 1/4 inch due to thermal expansion. The original clearance of the RHS heat exchanger was extremely small, allowing binding to occur between the shim block and the lug. There is a 0.0002 to 0.125 inch gap specified to allow free vertical movement of the heat exchanger. In this case, one of the eight lug to shim block clearances was insufficient, allowing sufficient bir. ding to lift the block and shear the 3/8 inch cap screws. The licensee's engineering analysis concluded that the broken cap screws do not compromise the seismic integrity of the heat exchanger as long as the shim block remains in place. Additionally, seismic forces in the vertical direction are not sufficient to overcome the weight of the shim blocks. The licensee intends to re-manufacture the shim blocks to provide more clearance for heat exchanger thermal expansion. In the interim, this heat exchanger will not be used for routine cooldown No unacceptable conditions were identified by the inspector's revie Modification to Containment Gaseous and Particulate Activity Monitor Plant Technical Specification 3.4.6 " Reactor Coolant System Leakage; Leakage Detection Systems" requires the operability of 3 leakage detec-tion systems: 1) The Containment Atmosphere Gaseous Radioactivity Moni-toring System, 2) The Containment Drain Sump Level or Pumped Capacity Monitoring System, and 3) The Containment Atmosphere Particulate Radio-activity Monitoring System. If any 2 of these systems are inoperable, the plant must be brought to Hot Standby within 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> A single skid mounted process monitor in the radiation monitoring system (RMS) performs the above gaseous and particulate leakage detection func-tions in addition to sampling and measuring iodine activity. The licen-see discovered that a failure or maintenance action on one of the three subsystems necessitated declaring the entire skid inoperable, thereby placing the plant in a 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> action statemen This inflexibility was due to a solenoid-operated valve intended to separate the particulate and gaseous samplers. The valve would not shut when the skid was placed in the bypass mod A Plant Modification Request, PMR 3-86-582, was initiated to correct this apparent deficiency. During the time required to resolve this PMR, a temporary modification has been instituted under Jumper-Lifted-Lead-By-pass (bypass jumper) 3-86-114, which alters the control circuits to the solenoid valve to cause it to shut in the bypass mode.
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The inspector reviewed the technical and safety assessments and safety evaluation worksheet mandated by Administrative Control Procedures (ACP's) QA-2.06B and QA03.08, as well as the levels of review the package received. He concluded that the modification fulfilled the requirements of 10 CFR 50.59. The bypass jumper was installed, verified, and tested on December 2, 1986. The inspector had no further question Radioactive Liquid and Gaseous Effluent Monitoring by Analog Channels On 12/18/86, the resident inspectpr was notified by the licensee that TS Table 4.3-8 and 4.3-9 Notations (1)/(2) b. and c were not required by the SP 3450 and SP 3449 series surveillance procedures and had not been performe These notes requi"ed an analog channel operational test to demonstrate that control room annunciation occurs on circuit failure or instrument downscale failure. On the same date, the PORC approved the revised surveillance procedures. The inspector reviewed the docu-mentation and discussed this issue with plant personnel, and asked if other I&C surveillance tests also may not have been performed. This issue is under review by the licensee and will be evaluated by the in-spector under the normal inspection progra i Failure to perform the circuit failure and downscale failure alarm operational tests on the analog channels is a violation of NRC require-ments. In accordance with 10CFR2-Appendix C, a notice of violation will not be issued because the condition was identified by the licensee, was of low severity level, was reported to the NRC, was quickly corrected, and was not a violation which should have been prevented by the correc-tive action on a previous violatio . Licensee Event Reports LERs submitted during this report period were reviewed. The inspector as-sessed LER accuracy, whether further information was required, if there were generic implications, adequacy of corrective actions, and compliance with the reporting requirements of 10 CFR 50.73 and Administrative Control Procedure ACP-QA-10.09. Selected corrective actions were checked for thoroughness and implementation as documented elsewhere in this report. The LERs reviewed were:
86-054-00 Failure to Establish a Fire Watch Within the Reactor Containment Within One (1) Hour A portion of the station fire water system was out of service for one hour and fourty-eight minutes on October 30, 1986 to allow repairs to an under-ground fire main isolation valve.
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That valve was damaged during excavatio Technical Specification 3.7.12.2.k Action Statement 3.7.12.2.a requires a continuous fire watch with back-up fire suppression equipment to be stationed at the electrical cable penetration area which is located within the sub-
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atmospheric containment. This event was addressed in Section 3 of Inspection l Report 50-423/86-33.
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86-055-00 Emergency Diesel Generator "B" Failed to Start Within Ten (10)
Seconds During surveillance testing conducted on November 6, 1986 at 1715, the "B" Emergency Diesel Generator started in 10.59 seconds. The licensee's analysis concluded that the cause of the slow start time was low lubricating oil and jacket water temperature This was verified during subsequent testing; at that time the unit started in 9.32 seconds. Although the keep warm systems are normally in service, they had been secured during preventive maintenance prior to this event. The inspector verified that the licensee considers the Emergency Diesel Generator operable only if its oil temperature is greater than stated minimum standby values. These are verified locally during Plant Equipment Operator (PE0) rounds. Also, low temperature is annunciated as a trouble alar "B" Safety Injection Pump Cooler The "B" Safety Injection Pump was found with service water secured to its cooler sub-system on November 30, 1986 at 010 The normally open manual cooler outlet valve was shut for approximately 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br />. That exceeded Tech-nical Specification 3.5.2 which mandated a reactor shutdown after 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> On-site followup of this LER is documented in Special Inspection Report 50-423/86-3 . Drug Abuse Allegation RI-86-A-139 This allegation stated that drug usage was prevalent at Millstone 3 during construction. It was a general allegation. NRC Region I letter File RI-A-139 of January 6, 1987 was forwarded to the licensee asking for a description of the applicable licensee programs and findings. Also, the resident inspectors reviewed their experience and developed the following information on this matte Construction Senior Resident Inspector (CSRI)
The CSRI at Millstone 3 from September 1983 until March 1986 is now the operations senior resident inspector (0 SRI) for Millstone 1 and 2. His Millstone 3 activities included observing welding, concrete placements, hanger installation, electrical component and other equipment installa-tion, and other activities by pipefitters, welders, electricians, operat-ing engineers, sheet metal workers, and painters. Testing activities performed by professional engineers and technicians were also observed, as were evolutions and tests performed by the operating staff. No use of unauthorized drugs or alcohol was observed. No physically incapaci-tated persons on the job were observed. The training and supervision of workers and separate checks of work by non-destructive testing, com-ponent tests, system tests, integrated system tests, and startup and operational checks provided an in-depth cross check of activities. If
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widespread use of unauthorized substances had occurred, it should have been evident in the questionable behavior of personnel and in the test results. That was not the cas One past concern was identified in this regard. Busesbringir!qconstruc-tion tradesmen from Rhode Island were noted to have beer coolers. In-spector concern about the potential for beer drinking during luncheon breaks was identified to the licensee, who had the buses removed from the site except while discharging or picking up passengers. The exist-ence of these coolers was a potential problem corrected acceptably by
, the licensee; no unauthorized use and no abuse of alcohol was found to have occurre Overall, at Millstone 3, the CSRI found well trained tradesmen, tech-nically competent engineers, and a knowledgeable operating staff that conscientiously and capably performed their jobs. In addition, no abuse or unauthorized use of drugs or alcohol has been observed on the Mill-stone site since the CSRI became the OSRI for Millstone 1 and 2. A pro-fessional and competent staff, with appropriate cross checking and supervision of activities, has been observed at Millstone 1 and 2 als Operations Senior Resident Inspector (OSRI)
Drug and alcohol abuse has not been a problem area while the present OSRI has been assigned to Millstone 3. This includes the initial period of reactor operation for Unit During this period the Unit 3 facility and personnel have been integrated into the Millstone Site Physical Security Syste The licensee's unauthorized substance usage checks have generally consisted of observation of personnel by the security force, and normal management observation and supervision. This is presently
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under revision. Overall, facility activities have been found to be safely and conscientiously performed. Performance that has been suspect as a possible result of use of unauthorized substances has not been identified. Staff professionalism and competence has been continually eviden Summary Use of unauthorized substances has not been identified as a problem are Facility activities have been performed with professionalism and compe-tence generally evident. Cross-checks of activities provide a defense against personnel error and ma1 performance. Supervisory and security checks for unauthorized substance usage are identified elements of the licensee's program. Individuals observed have been considered fit for duty. What the inspectors have observed is that unauthorized substance use is not evident from NRC sampling checks of licensee activitie The licensee's drug and alcohol program, as described and observed, is performed incident to normal security and supervisory functions. An active and aggressive licensee program for preventing the use of un-l
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authorized substances through other means (e.g., ongoing education and training, etc.) has not, however, been clearly evident. The in progress revision of the licensee's program will be assessed incident to routine inspection and evaluation of performanc . Review of IE Bulletin 86-03, Potential Failure of Multiple ECCS Pumps Due to the Single Failure of an Air Operated Minimum Flow Valve As a result of IE Bulletin 86-01 and a 10 CFR 21 report from a Westinghouse PWR, the it pector previously reviewed the Millstone 3 Emergency Core Cooling System (ECCS) for potential single failure problems with common air operated minimum flow recirculation valves. A review of the licensee's actions was also documented in Detail 8 of NRC Inspection Report 50-423/86-1 The licensee responded to IEB 86-03 in letters' A06150, Mr. John F. Opeka to Dr. Thomas E. Murley, dated November 20, and December 17, 1986. The inspector reviewed that response for Millstone 3 and agreed that the problem does not exist at that unit. The recirculation valves are all motor-operated, not air
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operate In a previous assessment the inspector concluded that there was no concern that a loss of power to these common recirculation line motor-operated valves will disable the Safety Injection (SI) or Charging Pumps.
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(The valves will fall as is.) Further, all of the SI and Charging pump re-circulation valves have position indication and off-normal annunciators on Main Boards 2 and 3, making their closure readily apparent. The most probable cause of closure of these valves would be failure to fully recover from a maintenance lineup. Since these are ECCS valves, they require dual verifica-tion of position during system restoration from a tagout. It was concluded that there are valid equipment and administrative measures to assure proper lineup.
l In the Millstone 3 Probablistic Risk Assessment, disabling the charging and safety injection pumps due to closure of common mininmum flow recirculation valves did not surface as being significan The inspector had no further questions on this matte . Review of IE Information Notice 85-45, Potential Seismic Interation Involving the Movable In-Core Flux Mapping System Used in Westinghouse Designed Plants
! The subject notice identifies concerns involving interactions between the non-
! safety-related portions of the movable flux mapping system and the tubing / seal table during a seismic event. The issue was identified by Westinghouse. The I,
potential interactions exist because portions of the flux mapping system that
! have not been seismically analyzed are located directly above the in-core
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instrumentation tubing / seal tabl The inspector reviewed drawings and pictures and discussed the system arrange-ment with the licensee staff. This item was also reviewed during Inspection 50-423/86-33 and is documented in Section 4 of that report. At Millstone 3, the thimble guide tubes are connected to the seal table a few feet above the l
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concrete floo Concrete walls enclose the area and loosely encase the tubes as they make the bend to the bottom of the core. The remaining mechanism (drive motors, isolation valves, interconnecting devices, etc..) is located directly above the seal table, similar to the plant discussed in the notic By letter dated September 18, 1985, the licensee documented their review of this issue to the NRC in accordance with 10 CFR 50.55(e). The resultant seis-mic analysis indicated that the equipment anchorage was not sufficient to limit the displacement of the tubing connecting the flux mapping equipment to the seal table. The letter documented their intent to add supports to correct this situation. The pictures reviewed during this inspection showed that the
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modifications had been made. The inspector had no further question . Observation of Maintenance The inspector observed and reviewed preventive and corrective maintenance to
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verify compliance with regulations, use of administrative and maintenance procedures, compliance with codes and standards, proper QA/QC involvement,
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use of bypass jumpers and safety tags, personnel protection, and equipment alignment and retest. The following activities were included:
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"A" Emergency Diesel Generator Service Water Cooler Disassembly and Repai "C" Steam Generator Feedwater Regulating Valve Bypass Valve operator replacemen No unacceptable conditions were identifie . Observation of Surveillance Testing The inspector observed parts of tests to assess performance in accordance with approved procedures and Limiting Conditions of Operation, removal and restcra-tion of equipment, and deficiency review and resolution. The following tests were reviewed:
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"B" Emergency Diesel Generator operational Surveillance test.
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Safety Injection Accumulator vent valve stroke timing and loss of power tes No unacceptable conditions were identifie . Safety Committee Meetings The inspector attended Plant Operations Review Committee (PORC) meetings on December 19 and 20. Technical Specification requirements for attendance were met. The meetings were characterized by frank discussions and questioning of causes and corrective actions. Individual members were encouraged to pro-
vide their opinion No deficiencies in Committee performance were observed.
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1 Management Meetings During this inspection, periodic meetings were held with senior plant manage-ment to discuss the inspection scope and findings. No proprietary information was identified as being in the inspection coverage. No written material re-lating to inspection findings was provided to the licensee by the inspector.
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