IR 05000245/1986017
| ML20210V621 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 10/01/1986 |
| From: | Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20210V568 | List: |
| References | |
| 50-245-86-17, 50-336-86-19, IEB-85-001, IEB-85-002, IEB-85-003, IEB-85-1, IEB-85-2, IEB-85-3, NUDOCS 8610100748 | |
| Download: ML20210V621 (9) | |
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U.S. NUCLEAR REGULATORY COP 91ISSION
REGION I
Report Nos:
50-245/86-17; 50-336/86-19 Docket Nos:
50-245/50-336 License Nos: DPR-21; DPR-65 Licensee:
Northeast Nuclear Energy Company Facility:
Millstone Nuclear Power Station, Waterford, Connecticut Inspection at: Millstone Units 1 & 2 Dates:
August 18, 1986 through September 29, 1986 Inspectors:
Geoffrey E. Grant, Resident Inspector Eben L. Conner, Project Engineer
' Approved by:
b O
loIIISC
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.E. C. McCabe, Chief, Reactor Projects Section 3B Date
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l Summary:
50-245/86-17; 50-336/86-19 (August 18 to September 29, 1986)
Areas Inspected:
This inspection included routine NRC resident inspection (91.0
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hours) of plant operations, surveillance, maintenance, radiation protection, physical security, fire protection, IE Bulletins, and the Unit 1 Standby Gas Treatment System.
Results:
No unacceptable conditions were identified.
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TABLE OF CONTENTS Page 1.
Summary of Facility Activities.......................................
2.
Review of Plant Operations - Units 1 & 2.............................
3.
Reactor Trip Due to Low Steam Generator Water Level - Unit 2.........
4.
Commencement of Refueling Outage - Unit 2............................
5.
Observation of Surveillance Testing..................................
5.1 Low Pressure Coolant Injection System Operability...............
5.2 Core Spray System Operability...................................
5.3 Emergency Condensate Transfer Pump Operational Readiness Test...
6.
Standby Gas Treatment System (SGTS) Design Inspection................
7.
IE Bulletin 85-01, Steam Binding of Auxiliary Feedwater Pumps........
8.
IE Bulletin 85-02, Undervoltage Trip Attachments of Reactor Trip Breakers.............................................................
9.
IE Bulletin 85-03, Motor-0perated Valve Common Mode Failure Due to Improper Switch Settings.............................................
10.
On-Site Safety Review Committee......................................
11.
Review of Periodic and Special Reports...............................
12.
Review of Licensee Event Reports (LERs)..............................
13.
Management Meetings..................................................
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DElna S 1.
. Summary of Facility Activities Unit 1 generally operated at full power during this report period.
Several times during the' period power was reduced to 65% - 75% to support locating and repairing condenser tube leaks.
Other routine power reductions were re-quired for thermal backwashing of main circulating water bays and testing of Main Steam Isolation Valves, turbine stop valves, and turbine control and intercept valves.
Unit 2. generally operated at full power during this report period until the unit tripped from 100% power on September 3 (see Detail 3).
Upon return to 100% power on September 5, the unit entered a two week period of reactor coastdown preceding the planned refueling outage.
On September 20 the unit was shutdown to commence a refueling outage scheduled to end December 20, 1986.
2.
. Review of Plant Operations (Units 1 & 2)
The inspector observed plant operations during regular and back shift tours of the following areas:
Control Rooms Turbine Buildings Auxiliary Building Cable Vaults Enclosure Building Fence Line (Protected Area)
Reactor Building Intake Structure Diesel Generator Rooms Gas Turbine Building Vital Switchgear Rooms Control Room instruments were observed for correlation between channels, pro-per functioning, and conformance with Technical Specificatioas.
Alarm condi-tions in effect and alarms received in the control room wera reviewed and discussed with the operators.
Operator awareness and response to these con-ditions were reviewed.
Operators were found cognizant of board and plant conditions.
Control room and shift manning was compared with Technical Specification requirements.
Posting and control of radiation,. contamination, and high. radiation areas were inspected.
Use of and compliance with Radiation Work Permits and use of required personnel monitoring devices were checked.
Plant housekeeping controls were observed including control of flammable and other hazardous materials.
During plant tours, logs and records were reviewed
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to ensure compliance with station procedures, to determine if entries were correctly made, and to verify correct communication of equipment status.
These records included various operating logs, turnover sheets, tagout and jumper logs, process computer printouts and Plant Information Reports.
The inspector observed selected actions concerning site security including per-l sonnel monitoring, access control, placement of physical barriers, alarm station operations, and compensatory measures.
No unacceptable conditions
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were identified.
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2 3.
Reactor Trip Due to Low Steam Generator Water Level - Unit 2 At 1110 on September 3, 1986 the reactor tripped from 100% power.
No major plant evolutions were being conducted at the time.
Due to a computer mal-function, the normal post-trip data analysis printout was not available to assist in the dete-mination of the trip cause and sequence of events.
How-ever, other computer printouts and control room indications available to the operators indicated the trip was caused by low levels in the Steam Generators (S/Gs) resulting from a loss of both operating Steam Generator Feed Pumps (SGFPs).
The SGFPs were lost on low pump suction pressure resulting from the loss of Feedwater Heater Drain flow caused by a failure of the Heater Drain pump combined discharge header flow control valve.
That flow control valve normally responds to level fluctuations in the Heater Drain Tank (HDT) and maintains level by varying the flow output of the Heater Drain pumps.
The valve operator failed when its control air line separated causing the valve to rapidly close to its minimum position. The cause of the 1/4" copper tubing air line failure has not been determined, but the tubing is undergoing failure analysis.
At the time of the tubing failure a great deal of heavy industrial activity was taking place in the immediate vicinity of the valve and its con-trol air line.
This activity was related to preparations for the Unit 2 re-fueling outage.
The inspector observed control room operator post-trip actions and followed the sequence of events to a stable shutdown condition.
The operators' actions, coordination, use of procedures, communications and level of knowledge were all excellent and resulted in a smooth and efficient transition to stable plant shutdown conditions.
The inspector reviewed trip related logs and records and attended the licensee's post-trip review and restart board.
No unacceptable conditions were identified.
4.
Commencement of Refueling Outage - Unit 2 On September 20, 1986 Unit 2 shut down to commence a 92-day refueling and maintenance outage.
The outage includes extensive work in the following areas:
l Reactor core shuffle and refueling;
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Main condenser tube, tubesheet and waterbox replacement;
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Steam Generator tube plugging and sleeving;
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Process computer replacement;
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"C" Reactor Coolant Pump motor replacement;
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Main Turbine overhaul; i
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Feedwater heater replacements; Appendix R Fire Protection including fire shutdown panel, emergency
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lighting, fire dampers, and fire suppression systems.
To date the majority of outage work has involved removal of interferences and project preparations.
Twice daily outage status / management meetings are con-ducted with a wide variety of project representatives.
Active discussions and information interchange appear to enhance management effectiveness and identify potential conflicts or coordination problems.
The inspector observed these activities and identified no deficiencies.
5.
Observation of Surveillance Testing The resident inspector observed several surveillance tests for performance in accordance with approved procedures and Limiting Conditions for Operations, correct removal and restoration of equipment, adequate deficiency review and resolution, proper calibration and use of test instrumentation, personnel qualification, and adequacy of procedure details.
5.1 Low Pressure Coolant Injection (LPCI) System Operability On September 11, 1936 the inspector observed the conduct of SP 622.7, LFCI System Operability from the control room and from equipment loca-tions.
LPCI and Core Spray (CS) valve operability tests were observed from the control room, motor control cabinets, and the valve locations.
This surveillance operates pumps, determines system flow rates, and ex-ercises critical motor-operated valves.
Special inspector attention was given to review of the surveillance procedure for technical specification conformance, adherence to administrative controls, observation of the test and review of the completed test documentation.
Discussions with licensee personnel conducting the surveillance found adequate preparation and a high level of knowledge pertinent to the surveillance.
No defi-ciencies were identified.
5. 2 Core Spray (CS) System Operability On September 11, 1986 the inspector observed the conduct of SP 621.10, CS System Operability Test, from both the control room and equipment locations. This surveillance operates pumps, determines the system flow i
rate, and exercises critical system motor-operated valves.
The inspec-I tor's review found no deficiencies.
5.3 Emergency Condensate Transfer (ECT) Pump Operational Readiness Test On September 11, 1986 the inspector observed SP 625.4, ECT Pump Opera-tional Readiness Test.
This surveillance tests pump operability, system flow rate, and motor-operated valve timing.
No deficiencies were noted.
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6.
Standby Gas Treatment System (SGTS) Design Inspection The Standby Gas Treatment system is designed to prevent the ground level escape of airborne radioactivity during accident or secondary isolation con-ditions.
This is accomplished by:
1) treatment and processing of the Reactor Building atmosphere through charcoal adsorbers,llEPA filters, and discharge through the 375-ft. stack, 2) ensuring only in-leakage to the Reactor Building by maintaining a small negative pressure within the Reactor Building and, 3) venting or purging the drywell or suppression chamber as necessary whenever primary containment is required.
The SGTS also provides a method for con-ducting Reactor Building leak rate testing.
The standby gas treatment system has two redundant trains of charcoal beds, fiiter beds and fans.
The system is sized to provide one air change per day in the Reactor Building.
Both trains are automatically started under con-tainment isolation conditions.
Both trains receive power from the emergency power system. The design is for one train to be operational with the other as standby.
Automatic and remotely controlled valves are provided for opera-tion of fans, filters and adsorbers.
The system is designed to filter and process Reactor Building atmosphere dur-ing Reactor Building isolation conditions, monitor the radioactivity for com-pliance with specified limits, and exhaust the secondary containment atmos-phere to the plant stack.
The system is also used whenever the secondary containment is providing containment, i.e. during refueling.
Automatic initiation of the SGTS is initiated by high drywell pressure, low reactor water level, high Reactor Building ventilation radiation, high steam tunnel ventilation radiation, and high refueling floor radiation.
Besides starting both trains of the SGTS, an automatic initiation signal will isolate the Reactor Building and steam tunnel ventilation systems.
Recently, another Boiling Water Reactor licensee's engineering review of the SGTS identified a design deficiency which could render the SGTS incapable of performing its design function in the event of a particular single failure.
The deficiency was the use of a fail-open damper in the cross-tie line on the outlet of the filter trains.
The single failure was one of the temperature detectors on the fire suppression system for the filter charcoal bed.
High temperature initiates the deluge system and renders the charcoal bed incapable of removing radioiodines.
This could result in an unfiltered release of con-tainment atmosphere.
The resident inspector reviewed and analyzed the Unit 1 SGTS design to deter-mine its susceptibility to this design deficiency and possible single failures.
The Millstone 1 SGTS does not have a fire suppression water deluge system associated with the filter trains.
Thus, this method of charcoal bed failure cannot occur.
The air-operated cross-tie damper on the outlet of the filter trains is of a normally closed / fail closed design which prevents inadvertent cross-connection of the separate filter trains upon loss of instrument air.
Additionally, the instrument air compressor is a vital load that sequences i
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onto the Emergency Diesel Generator approximately 30 seconds after loss of normal power, thus minimizing the potential loss of instrument air pressure.
No potential single mode failures were identified during the inspector's SGTS design review.
7.
IE Bulletin 85-01, Steam Binding of Auxiliary Feedwater Pumps By letter dated February 28, 1986, NNEC0 provided their corrective actions in response to the subject bulletin for Millstone 2.
Those actions were re-visions to Plant Equipment Operator Logs (OPS Form 2669A-1) and the Auxiliary Feedwater System operating procedure (OP 2322).
The inspector found the lic-ensee's response timely, in that corrective actions were taken within the specified 90 days and formal response to the bulletin was submitted within the specified 120 days.
Revisions to the log and procedure were in conform-ance with the licensee's commitments and the bulletin.
The inspector found that the revisions had been approved by the PORC on January 16, 1986, up-to-date copies were being used by the plant staff, and shift staff were aware of the bulletin and its implementation at Millstone 2.
There were no dis-crepancies noted.
8.
IE Bulletin 85-02, Undervoltage Trip Attachments of Reactor Trip Breakers By letter dated January 24, 1986, NNEC0 provided a response to the bulletin regarding Undervoltage Trip Attachments of Westinghouse DB-50 Type Reactor Trip Breakers.
This response provided information why the requirements of the subject bulletin are not applicable to Millstone Units 1 and 2.
The in-spector reviewed the bulletin and confirmed that it is not applicable to BWRs (Unit 1) and PWRs manufactured by Combustion Engineering (Unit 2).
9.
IE Bulletin 85-03, Motor-Operated Valve Common Mode Failure Due to Improper Switch Settings By letter dated June 11, 1986, Northeast Utilities provided the response for their four plants.
The response describes implementation of a program to ensure that switch settings on the specified safety-related motor-operated valves are selected, set and maintained correctly to accommodate the maximum differential pressures expected during normal and abnormal events within the design basis.
Scheduled completion dates are provided.
The inspector found the licensee's response timely, in that the program for implementation was submitted within the extension period requested on May 14, 1986.
Discussions were held with corporate and site personnel regarding their selection of motor-operated valves covered by the bulletin, the three vendors being considered to supply test equipment, and the order of testing (Millstone
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I 1 in May 1987 with Haddam Neck, Millstone 3, and Millstone 2 following).
No problems with the licensee's program were identified.
However, the NU re-sponse to the concerns expressed in Bulletin 85-03 will be reviewed at a later i
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10. On-Site Safety Review Committee-The resident inspector attended meetings of the Unit 2 Plant Operations Review Committee (PORC) on September 17 and 25.
Technical Specification 6.5.1 re-quirements for committee composition were met.
The meeting agenda included reviews of the following:
Plant Design Change Requests (PDCRs) covering installation of main steam
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line radiation monitors; installation of control room emergency lighting; replacement of control board knife switches; use of alternate material in_ valve stem replacement; replacement of the process computer.
Bypass jumpers for placement of temporary lead shielding and removal of
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interference in the turbine building.
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Plant Incident Report (PIR) 86-13.
Various minor changes and revisions to operations maintenance and sur-
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veillance procedures.
The presentations elicited active questioning and discussions and indicated that adequate review and analysis of the items were conducted prior to the meeting.
Items having unresolved questions were tabled pending further an-
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alysis.
Committee members presented an informed and critical overview of plant design and operations.
No deficiencies in FORC performance were ob-served.
11.
Review of Periodic and Special Reports Upon receipt, periodic and special reports submitted pursuant to Technical Specifications were reviewed.
This review verified that the reported infor-mation was valid and included the NRC required data; that test results and supporting information were consistent with design predictions and performance specifications;~and that planned corrective actions were adequate for resolu-tion of the problem.
The inspector also ascertained whether any reported in-formation should be classified as an abnormal occurrence.
The following re-ports were reviewed:
Monthly Operating Report for Unit 2 operation from July 1-31, 1986.
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Monthly Operating Report for Unit 1 operation from July 1-31, 1986.
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Monthly Operating Report for Unit 2 operation from August 1-31, 1986.
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Monthly Operating Report for Unit 1 operation from August 1-31, 1986.
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No unacceptable conditions were identified.
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Review of Licensee Event Reports (LERs)
LERs recently submitted by the licensee were reviewed.
The inspector assessed LER accuracy, whether further information was required, if there were generic implications, adequacy of corrective actions, and compliance with the report-ing requirements of 10 CFR 50.73.
Selected corrective actions were checked-for thoroughness and implementation.
LERs reviewed included:
86-005, Reactor Trip due to Low Steam Generator Water Level (Unit 2)-- 86-020, Security Related
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86-021, Security Related
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No deficiencies were identified.
13.
Management Meeting:
During this inspection, periodic meetings were held with senior plant manage-ment to discuss inspection content and findings.
A summary of the findings was discussed with the licensee at the conclusion of the report period.
No proprietary information was identified as being in the inspection coverage.
No written material was provided to the licensee by the inspector.
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