IR 05000213/1990004
| ML20034B669 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 04/18/1990 |
| From: | Haverkamp D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20034B665 | List: |
| References | |
| 50-213-90-04, 50-213-90-4, IEIN-86-001, IEIN-86-1, NUDOCS 9004300161 | |
| Download: ML20034B669 (12) | |
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U.S. NUCLEAR REGULATORY COMMIS$10N
REGION I
Report No.
50-213/90-04
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License No.
DPR-61 Licensee:
Connecticut Yankee Atomic Power Company P. O. Box 270 Hartford, CT 06141-0270 Facility:
Haddam Neck Plant location:
Haddam Neck Connecticut
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Inspection dates:
February 16, 1990 to March 27, 1990 Reporting Inspector:
John T. Shediosky, Senior Resident Inspector Inspectors:
Andra A. Asars, Resident Inspector John T. Shediosky, Senior Resident Inspector Approved by:
b oM w({a-4[P/AD onal T R. Haverkamp, Chief /
Date Reactor Projects Section 4A Division of Reactor Projects Inspection Summary:
Inspection on February 16, 1990 to March 27, 1990
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(Inspection Report No. 50-213/90-04)
i Areas Inspected:
Routine safety inspection by the resident inspectors. Areas reviewed included plant operations, radiological controls during thermal shield cutting and. cleaning of the refueling water storage tank, reactor vessel thermal shield cutting and shipping, plant operations review committee activi-ties, and-written reports.
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Results: This was the fifth routine resident inspection during the 1989/1990 Refueling Outage.
Significant progress was made in the design change to the reactor interals; the thermal shield was cut into small sections and some of those were shipped off site (Section 4.1.1).
The refueling water storage tank was drained and cleaned (Section 4.1.2).
The potential for degradation of auxiliary feedwater flow was discovered and reported (Sections 2.2.1 and 6.1).
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TABLE OF CONTENTS Page
1.
Summary of Facility Activities (71707)*.............
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Plant Operations (71707, 71710, and 93702)
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2.1 Operational Safety Verification -..............
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2.2 Follow-up of Events Occurring During the Inspection
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Period...........................
2.2.1 Discovery of Unanalyzed Condition Affecting
Auxiliary Feedwater..............,.
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Radiological Controls (71707)..................
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Maintenance and Surveillance (61726. 62703, and 71707)
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i 4.1 Maintenance Observation
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4.1.1 Reactor Vessel Thermal Shield Disposal
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4.1.2 Refueling Water Storage Tank Cleaning...._,.... 6 i
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Security (71707)
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Engineering and Technical Support (37700, 37828, and 71707)...
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6.1 Potential Degradation of Auxiliary Feedwater Flow.....
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Safety Assessment and Quality Verification (40500, 71707,
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90712, and 92700)..................-......
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j 7.1 Plant Operations Review Committee
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7.2 Review of Written Reports.................
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8.
ExitInterview(92703)
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- The NRC Inspection Manual inspection procedure or temporary instruction
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that was used as inspection guidance is listed for each applicable report
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section.
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I DETAILS 1.
Summary of Faci,11ty Activities The Fifteenth Refueling Outage continued during this inspection period.
The major work activity was the cutting and shipping of the reactor vessel thermal shield sections. All reactor fuel remained in the spent fuel pool.
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2.
Plant Operations 2.1 Operational Safety Verification The inspectors observed plant operation and verified that the plant was operated safely and in accordance with licensee p-ocedures and regulatory requirements.
Regular tours were conducted of the follow-
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ing plant areas:
control room security access point
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primary auxiliary building protected area fence
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radiological control point intake structure
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electrical switchgear rooms diesel generator rooms
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auxiliary feedwater pump room turbine building
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Control room instruments and plant computer indications were observed for correlation between channels and for conformance with technical
specification (TS) requirements. Operability of engineered safety features, other safety related systems and on-site and off-site power sources were verified.
The inspectors observed various alarm condi-tions and confirmed that operator response was in accordance with plant operating procedures.
Routine operations surveillance testing was also observed. Compliance with TS and implementation of appro-priate action statements for equipment out of service was inspected.
Plant radiation monitoring system indications and plant stack traces were reviewed for unexpected changes.
Logs and records were reviewed l
to determine if entries were accurate and identified equipment status
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or deficiencies. These records included operating logs, turnover sheets, system safety tags, and the jumper and lifted lead book.
Plant housekeeping controls were monitored, including control and storage of flammable material and other potential safety hazards.
The inspectors also examined the condition of various fire protec-
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tion, meteorological, and seismic monitoring systems. Control room
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and shift manning were compared to regulatory requirements and portions of shift turnovers were observed. Control-room access was properly controlled and a professional atmosphere maintained.
In addition to normal utility working hours, the review of plant operations was routinely conducted during backshifts (evening shifts)
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and deep backshifts (weekend and midnight shifts).
Extended coverage was provided for 63 hours7.291667e-4 days <br />0.0175 hours <br />1.041667e-4 weeks <br />2.39715e-5 months <br /> during backshifts and 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> during deep backshifts. Operators were alert and displayed no signs of inattention to duty or fatigu.
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2.2 Follow-up of Events Occurring During Inspection Period During the inspection period, the inspectors provided on-site cover-age and follow-up of unplanned events.
Plant conditions, alignment-of safety systems, and licensee actions were reviewed. The inspec-
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tors confirmed that required notifications were made to the NRC.
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During event follow-up, the inspector reviewed the corresponding plant information report (PIR) package, including the event details, root cause analysis, and corrective actions taken to prevent recur-rence. The following events were reviewed:
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2.2.1 Discovery of Unanalyzed Condition Affecting-Auxiliary Feedwater
The discovery of a previously unanalyzed condition affecting the performance of auxiliary feedwater (AFW) was reported via the NRC Emergency Notification System (ENS) on March 16, 1990 at 10:10 a.m.
This finding resulted from the engineering investiga-
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tion -into an in-service test (IST) surveillance deficiency of a
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December 7, 1989 test.
Reverse direction seat leakage exceeded the acceptance criteria during testing.
The point of concern is that four check valves are each relied upon to direct auxiliary feedwater flow through the feedwater regulating valves bypass lines to the steam generators.
Failure of any one of the four check valves may
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result in back-flow of auxiliary feedwater into the mein feedwater system. A reportability evaluation of this finding, which was initiated on March 3, 1990, had not been completed at the end of this inspection period. However, based on prelimi-nary information, the ENS report was made by the licensee to the requirements of 10 CFR Part 50.72(b)(2)(iii).
The ongoing analysis of this preexisting condition includes an examination of the acceptability of an alternate AFW flow path.
In the event that the plant design basis was met with the check
valve failure, the licensee will make a telephone update the NRC duty officer; otherwise, a licensee event treport will be submitted. Additional details are provided in section 6.1.
3, Radiological Controls During routine tours of the accessible pisnt areas, the inspectors ob-served the implementation of selected portions of the licensee's radiolog-ical controls program.
The utilization and compliance with radiation work permits (RWPs) were reviewed to ensure detailed descriptions of radiologi-cal conditions was provided and that personnel adhered to RWP require-ments. The inspectors observed controls of access to various radiologi-cally controlled areas and the use of personnel monitors and frisking methods upon exit from these areas.
Posting and control of radiation areas, contaminated areas and hot spots, and labelling and control of containers holding radioactive materials were verified to be in accordance
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During this inspection period, radiological
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controls for the following activities were observed.
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reactor vessel thermal shield cutting and shipment, and
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i refueling water storage tank cleaning (see report sections 4.1.1 and
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4.1.2)
Health physics technician control and monitoring of these activities were determined to be adequate.
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Maintenance and Surveillance 4.1 Maintenance Observation The inspectors observed various maintenance and problem investigation activities for compliance with procedures, plant technical snecifica-tions, and applicable codes and standards.
The inspectors also verified the appropriate quality services department (QSD) involve-ment, safety tags, equipment alignment and use of jumpers, radiologi-cal and fire preventico controls, personnel qualifications,
post-maintenance testing, and reportability.
Portions of activities that were reviewed included:
4.1.1 _ Reactor Vessel Thermal Shield Disposal
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The reactor vessel thermal shield had been cut into three
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segments and removed from the core support barrel prior to
this inspection period. At that time, special lifting devices were attached to each segment and those assemblies were set on
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i the floor of the reactor cavity.
l During this inspection period the first two segments were each cut into thirteen pieces to allow shipment within either the TN-RAM or CNS-3-55 transportation casks. The third segment
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was placed into the stand in preparation for cutting, i
The first five of fourteen shipments with material from these segments were dispatched from the site during the inspection period. Three of the five arrived and were accepted by the burial site at Barnwell, SC.
In addition to station procedures, these activities were
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controlled by a series of vendor procedures:
- VP 526, MP 2.7.1 CYW-20 Thermal Shield Removal Phase III CYW, Revision 0, dated February 16, 1990.
- VP 527, TR-0P-013 Handling Procedures for Transport Cask l
No. 1-13G Certificate of Compliance No. 9216 Revision 0, dated February 2, 1990.
- VP 528, QA-TP-003 Air Leak Test of CNS 3-55 Transport Cask, Revision 0, dated February 23, 1990.
- VP 529, TR-OP-019, Handling Procedure for Chem-Nuclear Systems,Inc.(CNSI)TransportCaskCNS-3-55 Certificate of Compliance No. 5805, Revision 0, dated February 23, 1990.
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- VP 530, 0113s Procedure for Handling and Loading of the TN-RAM Shipping Cask (Haddam Neck), Revision 0, dated February 23, 1990.
Debris Control:
The control of debris generated during the thermal shield removal continued to be a well-defined process.
Plasma are was selected for the secondary cutting technique in which the three 120' segments were cut into smaller sections which were suitpble for shipment. Because the debris generated by this process is granular in consistency with some slag.and heavy bead size produced, the activity was isolated in a confinement enclosure within the reactor cavity.
The thermal shield segments were rotated to a horizontal posi,on and placed onto a stand, which provided support and remotely operated positioning for both the segment and the plasma arc device.
The entire assembly was within a stainless steel enclosure at the east end of the reactor refueling cavity.
The confinement enclosure contained the debris which ranged in size from 38 microns to heavy pieces of slag.
While isolated within the enclosure, heavier debris fell into a tray on the cutting stand while finer particles were fil-tered.
Recirculation filters operated on both the water within the enclosure and in the refueling cavity.
Hydrogen peroxide was added to the water to promote the oxidation of the soluble ferrous hydroxide into an insoluble ferric hydroxide.
Hydrogen peroxide, maintained between 8 to 12 ppm, therefore improved the effectiveness of the filtration systems.
A fresh debris barrier was also installed over the top of the reactor vessel prior to initiating the secondary cutting.
This polyester felt material was intended to provide a second level of protection to the reactor components.
Extensive cleanup evolutic,ns are planned for the reactor cavity following the shipment of the thermal shield sections.
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Reactor system piping will be cleaned through a series of flushes and inspections. With the exception of cleaning the refueling water storage tank, addressed below, these were in the planning stages at the end of the inspection period.
Radiological Protection:
In addition to controls normally employed during remote handling of reactor components, this process required that the area above the cutting enclosure be ventilated directly to the containment purge system.
This was because of the large amount of gas generated during the metal burning.
In addition
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e to radioactive particles which may have been brought out of
the pool 21eng with this gas, hydrogen is liberated by the
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plasma are process.
A tent-like enclosure over the cutting area captured the airborne debris.
Its exhaust system provided dilution air and
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direct discharge of exhaust particles and gases to the contain-r ment purge system for radiation monitoring and eventual exhaust.
Flammability / Explosion Hazards:
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In addition to the hydrogen gas generated by the high tempera-ture plasma a,c metal burning, a 30/70% mixture of hydrogen and argon gas was used as the supply to the plasma are pro-cess.
The pctential for accumulation of hydrogen in the
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containment deme was addressed through ventilation, explosive
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gas monitoringi gas supply safety features and procedural controls.
Direct exhaust was provided from the area above the cutting enclosure to the containment purge system.
This handled both airborne radioactive material as well as any hydrogen gas.
The containment air recirculation system provided for mixing in the containment dome. The exhaust system and containment i
ventilation alignment requirements were stated in the work procedure.
Four explosive gas detectors were employed, one monitoring the i
volume within the exhaust tent and three monitoring the containment dome. Of the three dome monitors, two drew a sample to their detectors using a pump from near the top of the dome, the third had its detector above the highest igni-tion source, the polar crane power rail. The detectors were of different types, two were battery operated and two operated on 120 volt ac.
Their alarm settings were at 20% of the lower explosive level.
There was no accumulation of hydrogen detected.
Gas supply to the containment was through a high pressure, i
armored hose.
Emergency shut off valves were prominently marked. An excess flow check valve was installed at the supply tank truck, which was parked near the plant hydrogen storage tanks at the south east corner of the protected area.
By procedure the gas supply was in service only during plasma arc operation. At other times the supply was isolated at the supply tank and the line vented.
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4 Inspection Activities:
The inspectors reviewed the thermal shield activities through-
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out the inspection period. Observations were made frequently at the work locations in the reactor containment. Several of
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the radiological surveys were observed prior to shipment of
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the container with thermal shield sections.
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The inspectors attended most of the PORC meetings in which reviews were made of the engineering analysis supporting the work and the work procedures.
Discussions were held frequently with plant management, supervisors involved with the thermal shield project, engineering and licensing person-nel, as well as those responsible for performing the work activities.
Of particular interest to the inspectors were measures to
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address the flammable gas hazards and radiation protection measures, along with any other activity which may have affect-
ed worker safety.. This included safe rigging practices of heavy loads, worker interface with radiation protection and station procedure requirements and monitoring the effective-ness of management oversight.
The inspectors observed a high regard for procedural compliance with the work procedures and radiation protection procedures. Work practices were professional with appropriate levels of supervision.
Proper interface was observed between the workers and the health physics technicians. The work progress was modulated to conform with these requirements.
The licensee maintained work practices designed to control
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potential " hot particles" from within the reactor cavity.
Good controls were observed in place over the work activities.
There were no unacceptable conditions identified.
l 4.1.2 Refueling Water Storage Tank Cleaning
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During this inspection period, the licensee cleaned the l
Refueling Water Storage Tank (RWST) in preparation for the primary system flushes. The RWST is to be used as the clean-
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water source for these flushes.
In support of the entries, scaffolding and a work platform
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were installed on the tank to permit entrance at the top manway.
The initial entry for tank survey was made on February 21 by a health physics technician. Difficulties were encountered with radiation monitoring equipment and the inadequate space on the work platform.
Health physics supervisors stopped work l
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A complete survey of the tank bottom was completed on February 22.
Radiation levels ranged from 40 millirem / hour to 3000 millirem / hour.
Smears indicated contamination levels up to 50,000 dpm/100 cm2 alpha.
The health physics technician also observed debris and several small metallic chips similar to those removed during cleaning and inspection of the reactor fuel.
The RWST was dewatered and the debris was collected and inspected.
Examination of the sludge removed from the bottom of the tank revealed that a small number of metallic chips were found.
Radiation levels of debris did not allow extended examination of_the sludge, but less than a dozen chips of the same type found to have damaged the fuel clad were identified.
This material was retained to allow laboratory analysis if determined necessary.
Tank cleaning continued until March 16. Quality services personnel conducted the final inspection and accepted the tank when two swipes from each quadrant contained no particles larger than one-sixteenth by one-sixteenth inch.
The inspectors observed the initial tank entries and verified that the entries were made in accordance with ADM 1.1-72, Enclosed Volume and Hazardous Atmosphere Work Practices.
Pre-entry briefings, rescue equipment readiness, and health physics coverage were observed. All activities were in accordance with station procedures and contamination controls were effective.
A second cleaning of the RWST was performed after the conclu-sion of this inspection period because the tank cleanliness acceptance criterion was changed to a partial size of no greater than one-sixteenth by one-thirtysecond inch.
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Security During routine inspection tours, the inspectors observed implementation of portions of the security plan. Areas observed included access point search equipment operation, condition of physical barriers, site access control, security force staffing, and response to system alarms and degraded conditions.
These areas of program implementation were deter-mined to be adequate.
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Engineering r ' Technical Support The inspector reviewed selected design changes and modifications made to the facility which the licensee determined were not unreviewed~ safety questions and did not require prior NRC approval as described by 10 CFR 50.59.
Particular attention was given to safety evaluations, Plant Operations Review Committee approval, procedural controls,
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post-modification testing, procedure changes resulting from these modifica-tions, operator training, and UFSAR and drawing revisions. The following activities were reviewed:
l 6.1 Potential Degradation of Auxiliary Feedwater Flow An analysis of the effect of check valve seat leakage on auxiliary
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I feedwater (AFW) performance was started after the test failure of valves in the feedwater regulating bypass lines.
Three of four valves failed to meet seat leakage acceptance criteria during a December 7, 1989 inservice test. The valves direct AFV flow to the
steam generators by preventing reverse direction flow back into the main feedwater system.
Leakage through any of the four valves will reduce the total AFW flow.
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There is one check valve (FW-CV-135-1, 2, 3, & 4) in each of four regulating valve bypass lines. A regulating valve and a bypass flow regulating valve are associated with each steam generator.
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valves are opened by flow when using the bypass regulating valves; they shut when AFW flow is applied between the check valve and the bypass regulating valve. Automatic initiation of AFW causes the l
bypass regulating valves to fully open.
Flow through the bypass lines is indicated on the main control board.
An engineering analysis of the impact of the leaking check valve on system design basis was ongoing at the end of the inspection period.
The quantity of lost flow due to seat leakage and the availability of an alternate flow path enter into these considerations.
Three of the four valves failed the 0.5 gpm seat leakage testing acceptance criteria in a test performed on December 7, 1989 per surveillance procedure SUR 5.7-129, Leak Testing of Feed Regulator Valves Bypass Line Check Valve FW-CV-135-1, 2, 3, and 4.
Seat leakage of these three valves was not quantified because of the inability to pressurize the valve to the required 1000 psi with a 4600 psi, 5 gpm pump.
Each valve was successfully retested after lapping the seat to the disc.
The maintenance history for each valve documents several instances of seat damage since inspections were first made in 1986, in response to NRC Information Notice 86-01, Failures of Main Feedwater Check Valves Causes Loss of Feedwater System Integrity and Water Hammer Damage.
Leakage testing of the valves began ia 1987.
Following licensee review of the maintenance history, serious doubt was expressed whether the valves could perform for more than one operating cycle.
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AFV flow can be directed into a separate three-inch line supplying each steam generator or directly through individual one and one-half inch lines.
This alternate flow path connection is made down stream of the main feedwater check valves.
Transfer of flow is accomplished remotely from the main control room by repositioning two motor operated valves. The operator action of opening the alternate supply header isolation (FW-MOV-35) and closing the AFW pump discharge
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header cross-tie valve (FW-MOV-160) is directed by emergency operat-ing functional recovery procedure FR-H.1, Response to Loss of Secon-l dary Heat Sink, Revision 6, dated July 7, 1989.
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l This action, however, allows flow from only one, (the "A") of the two steam turbine driven AFW pumps.
Use of the other turbine driven AFW pump would require manual valve operation. The electric motor driven pump is not safety related. Additionally, there is no flow indica-tion of the auxiliary path.
There were no unacceptable conditions identified with the licensee's ongoing evaluation.
Reportability_was discussed in section 2.2.1.
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Safety Assessment and Quality Verification 7.1 Plant Operations Review Committee The inspectors attended several plant operations review committee (PORC) meetings. Technical specification 6.5 requirements for required member attendance were verified.
The meeting agendas included procedural changes, proposed changes to the technical specifications, plant design change records, and minutes from previ-ous meetings.
PORC meetings were characterized by frank discussions and questioning of the proposed changes.
In particular, considera-tion was given to assure clarity and consistency among procedures.
Items for which adequate review time was not available were postponed to allow committee members time for further review and comment.
Dissenting opinions were encouraged and resolved to the satisfaction of the committee prior to approval. The inspectors observed that the
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l PORC adequately monitored and evaluated plant performance and j
conducted a thorough self-assessment of plant activities and programs.
7.2 Review of Written Reports Periodic and special reports, licensee event reports (LERs), and safeguards event reports (SERs) wore reviewed for clarity, validity, accuracy of the root cause and safety significance description, and adequacy of corrective action. The inspectors also verified that the reporting requirements of 10 CFR 50,73, 10 CFR 73.71, Station Administrative and Operating, and Security Procedures, and Technical
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Specification 6.9 had been met.
The following reports were reviewed:
LER 90-01 Design Deficiency Identified in Service Water Filters LER 90-02 Fire Barrier with Temporary Seal Determined Inoperable Haddam Neck Plant Semiannual Radioactive Effluent Report, July 1 to December 31, 1989
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Haddam Neck Plant Annual Radioactive Effluent Dose" Report, January 1
to December 31, 1989 Haddam Neck Plant Monthly Operating Report 90-02, February 1, to February 28, 1990 Haddam Neck Plant and Millstone Station Annual Report, January 1 to December 31, 1989 Haddam Neck Plant Bimonthly Progress Report No. 21, New Switchgear.
Building Construction No unacceptable conditions were identified.
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Exit Interview During this inspection, periodic meetings were held with station manage-
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ment to discuss inspection observations and findings. At the close of the inspection period,-an exit meeting was held to summarize the conclusions of the inspection. No written material was given to the licensee and no proprietary information related to this inspection was identified.
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