IR 05000423/1987005
| ML20205P011 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 03/23/1987 |
| From: | Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20205N986 | List: |
| References | |
| 50-423-87-05, 50-423-87-5, IEIN-87-008, IEIN-87-8, NUDOCS 8704030156 | |
| Download: ML20205P011 (11) | |
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U.S. NUCLEAR REGULATORY COP 9tISSION
' REGION I Report No.
50-423/87-05 Docket No.
50-423 License No.
NPF-49 Licensee:
Northeast Nuclear Energy Company P.O.- Box 270 Hartford, CT 06101-0270 Facility Name: Millstone Nuclear Power Station, Unit 3 Inspection At: Waterford, Connecticut i
Inspection Conducted:
February 18, 1987 - March 16,-1987
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Inspectors:
J. T. Shedlosky, Senior Resident Inspector F. A. Casella, Resident Inspector E. L. Conner, Project Engineer M. C. Kray, Reactor Engineer M. Dev, Reactor Engineer Approved by:
h 3lidF E. C. McCabe, Chief, Reactor Projects Section 3B Date Inspection Summary:
Areas Inspected:
Routine on-site resident inspection (110 hours0.00127 days <br />0.0306 hours <br />1.818783e-4 weeks <br />4.1855e-5 months <br />) of shutdown planning, plant operations, radiation protection, physical: security, fire protec-
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tion, surveillance, and maintenance.
Results: This inspection identified satisfactory performance in all areas.
e704030156 870326 DR ADOCK o 4y3
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TABLE OF CONTENTS PAGE 1.
Summary of Facility Activities.......................................
2.
Review of Specific Activities........................................
a.
Reactor Scram - March 7, 1987...................................
b.
Power Reduction due to Quadrant Tilt Measurement Problem........
c.
Motor Driven Feed Pump Trip Test................................
d.
Normal Switchgear C02 Fire Suppression Test.....................
3.
Licensee Event Reports...............................................
4.
Followup of Actions on SALEM ATWS Events (Generic Letter 83-28)......
5.
Fitness for Duty Program.............................................
6.
Allegation RI-86-A-077, Class IE Battery Masonry Walls...............
7.
Licensee Actions on Previously Identified Items......................
a.
Management Control of PDCRs.....................................
b.
PDCR Documentation.............................................
c.
AFW Pump Suction Pressure Oscillations..........................
8.
Reporting of Events..................................................
IE Information Notice 87-08, " Degraded Mctor Leads in Limitor 9.
Motor Operators".............................................que DC
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10.
Observation of Haintenance...........................................
11. Observation of Surveillance Testing..................................
12.
Correction of Chlorine Monitor Problems..............................
13. Management Meetings..................................................
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DETAILS 1.
Summary of Facility Activities The plant operated at full power from the beginning of the inspection period until 2227, March 2 when a forced reduction in power was made in response to a Motor-Driven Feedwater Pump shaft seal failure.
Because the "A" Turbine-Driven Feedwater Pump was out of service for similar maintenance, power was reduced to 50%.
Feedwater demand at that level was within the capacity of one pump.
Repairs were completed to the "A" Turbine-Driven Pump at 1845, March 3 and the reactor was returned to full power at 1404, March 4.
A reactor scram occurred at 1743, March 7 because of low level in the No. 4 Steam Generator.
Feedwater flow to that steam generator was stopped when the Feedwater Isolation Valve shut in the associated feedwater line.
The valve was re-opened after the replacement of a failed solenoid valve.
The reactor was made critical at 0737, March 8, and was returned to full power at.2200, March 10.
The plant was shutdown on March 14 for a planned 25 day mid-cycle maintenance outage. An inservice test to monitor the effect of a feedwater pump trip was performed incident to shutting down.
With the plant at 50% power, the. running feedwater pump was stopped.
Test instrumentation measured feedwater pump discharge pipe displacement and strain.
A manual reactor scram was made about 15 seconds after stopping feedwater flow.
A reactor cooldown was then made
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to Mode 5 for the outage.
2.
Review of Specific Activities Although the perfornance of plant operators and equipment generally was as expected, several problems occurred and are summarized below:
a.
Reactor Scram - March 7 The "D" Feedwater Line Isolation (FWLI) valve (3FWS*CTV41D) closed at 1743, March 7 while at full power.
The subsequent reactor scram on low steam generator level occurred 21.7 seconds after the valve started to shut.
The cause of the valve closure was found to be a failed electric solenoid valve (S0V41D1).
The normally energized solenoid failed elec-trically, causing the isolation valve to shut.
The FWLI valve is an Anchor Darling (AD) self-contained, hydraulic actu-ator, non-return valve.
Each FWLI valve utilizes four AD solenoid valves for opening, closing and maintaining accumulator charge.
The "A" sole-noid valve, 50V4101, is the only one of these which is energized to keep the FWLI valve open during normal operation.
The following sequence summarizes the transient:
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17:43:12.52 Partial closure of "D" Feedwater Line Isolation Valve (3FWS*CTV410)
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"0" Feedwater Line Isolation. Valve. Closed-t 17:443:14.23 No. 3 Steam Generator Flow Mismatch-
17:43:20.82 No. 3 Steam ~ Generator Level. Deviation 17:43:32.92 No. 3. Steam Generator Low Low Level Trip.
i 17:43:34.232-Reactor Scram Breaker A open'
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17:43:34.236 Reactor Scram Breaker B open 17:43:34.33 Turbine Trip.
17:44:02.45 Generator Breaker _ Trip
The Senior Resident Inspector arrived in the Control Room shortly after-the reactor trip and verified that the licensee's~ personnel had adhered-to the post-trip procedures.
Review found that the licensee had properly
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identified the cause for the reactor scram,-that equipment had performed
satisfactorily, and that corrective actions were' identified and imple-
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mented.
i Inspections'were also made of the isolation. valve.following repairs.
It was noted that the cover of the electrical terminal box had only one of four cover fasteners engaged.
The inspector brought this'to1the at-tention of the Shift Supervisor on duty who had the cover properly j
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secured.
The licensee informed the inspector that this terminal-box did
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not contain environmentally qualified electrical equipment.
There were
no additional problems identified.
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Following the scram review and repair of the'"A" solenoid-valve, a.reac-l
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tor startup was commenced at 0449, March 8.
The reactor was made criti-
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cal at 0737; the turbine generator was placed on line at 1434, March 8;
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the plant reached full power at 2200, March 10.
q b.
Power Reduction due to Quadrant Tilt Measurement Problems
4 During the Quadrant Power Tilt Ratio (QPTR): surveillance performed on-i March 9, 1987, Nuclear Instrument (NI) Channel 43 was observed-reading _
4% higher tilt, to the bottom-of the core, than the'incore detectors.
- The channel was declared inoperable and placed in the tripped condition
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per Technical Specification Table 3.3-1, Action 2.a.
Action 2.c of this
Specification requires that either reactor power be less than175% within.
four hours or the QPTR be monitored at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. When
the second QPTR was due, 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after NI-43 became inoperable, the-1-
incore monitoring system became inoperable due to problems.with drive j
motor "C".
Since the QPTR could not be determined,' reactor power was 1-reduced below 75% at 0950, March 10. -The'incore detector neutron flux
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mapping system was returned to service and power ascension commenced'
j about 1245, March 10.
The inspector witnessed part-of the power reduc-T tion and power ascension evolutions; no problems were identified.
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I c.
Motor-Driven Feed Pump -Trip' Test
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.The inspectors observed performance of,In-Service Test (IST) 3-87-006;
" Motor. Driven Feed Pump - 50% Trip", PORC' approved on March.12, 1987.
The test was. performed to measure and record feed water. pump discharge-
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pipe displacement and vibration.between the motor-driven. feed pump.and the.first point feed heaters, includingLthe feed pump. recirculation.line,'
i during the high loading conditions presented by a pump trip from 50%
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power while supplying all-feed flow.
Computer models'have. indicated-additional lateral support on the subject piping was necessary-and _ tem-porary supports have been added.. Data from the test will be used to
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refine the model's evaluation and to determine.a final. support design. _
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The test was originally _ scheduled during the power ascension test period but was. postponed due to the licensee's inability to comply with speci-fied plant conditions.
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Test performance was deliberate and well_ controlled.' Operators _were fully aware of required actions and the expected response.
The reactor
was manually scrammed approximately 15' seconds.after the feed pump trip.
There were no anomalous plant conditions experienced.
Pipe displacement-
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was not visible to the unassisted eye..The inspectors had no questions.
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d.
Normal Switchgear Carbon Dioxide' Fire Suppression Test
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i Although the normal switchgear room Carbon Dioxide _ system operation was found satisfactory for fire suppression during:the start-up test' program,
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j an excessive amount of carbon dioxide was discharged, causing the Health
Physics, Chemistry, and Instrumentation and Control shops to be unin-i habitable.
More important, carbon dioxide concentrations were measurable
in the control room.
To correct this, the initial release time from the
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3-inch supply line has been reduced from.5 minutes to 2 minutes and 25
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seconds.
System operability was retested at 0400, March 14, 1987.
The acceptance criteria of: (1) attaining a-30% carbon dioxide concentration
in 2 minutes was satisfied in 1 minute, and (2) attaining a 50% concen-l
tration in 7 minutes was satisfied in 2 minutes.
The final concentration-
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was about 70% ca'rbon dioxide.
The amount of' carbon dioxide released was approximately 9 tons, compared to 17 tons released during the start-up
test.
There was no detectable carbon dioxide in the Control Room or in
the Instrument and Controls shop.
The Chemistry and Health Physics shops
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received slight concentrations of carbon dioxide but were fully habitable.-
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The inspector had no further comments.
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3.
Licensee Event Reports
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LERs submitted during this report period were reviewed.
The inspector assessed LER accuracy, whether further information was required, if there were generic
j implications, adequacy of corrective actions, and compliance with the reporting j
requirements of 10 CFR 50.73 and Administrative Control Procedure ACP-QA-10.09.
Selected corrective actions were checked for thoroughness and implementation.
i The LERs reviewed were:
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87-003-00.
In two tests, the "B" Emergency Diesel Generator failed.to
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start in less than 10 seconds.
This machine normally starts in about-9.5 seconds.
In these tests, start times were 10.16 and 10.22 seconds.
The licensee considers these tests invalid.
Maintenance starting had not been accomplished following injector maintenance on January 14.
An inadequate air purge of the fuel system was the cause.
In the second test, occurring on February 5, low oil temperature resulted from insuf-ficient warming time following maintenance.
In both cases, after deter-mining the cause, a satisfactory follow-up test was completed.
The in-spector had no further questions.
87-004-00.
Trip of Motor-Driven Auxiliary Feedwater Pumps during testing.
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This addressed in report 50-423/87-02. The LER addressed whether the Auxiliary Feedwater (AFW) Pump suction line pressure oscillations might cause a premature pump trip during post-accident operation when Demineral-ized Water Storage Tank (DWST) level decreased.
The licensee considered the worst pressure decrease measured during steady flow testing, and assumed the minimum required initial DWST level.
The licensee's calcu-lations indicated that the "B" AFW pump would have tripped with about 15,000 gallons of the design basis volume remaining in the DWST.
The pressure oscillations measured for the
"A" pump were not significant; calculations indicate that the entire design basis volume in the DWST would have been available.
The low suction pressure pump trip was re-moved from the motor-driven AFW pumps.
This trip function was removed from the turbine-driven AFW pump during plant construction.
Unresolved item 50-423/87-02-01.is closed.
87-005-00.
Failed Control Room Pressure test due to a mispositioned
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throttle valve.
The cause was personnel error and procedure inadequacy.
The valve was repositioned.
A successful test which kept the Control Room Envelope at a pressure of +1/8 inch (water) or greater for at least 60 minutes followed.
The inspector had no further questions.
87-006-00.
Failure to surveil an area temperature every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
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was missed once for the Auxiliary, Fuel, and Engineered Safety Features Buildings Pipe Tunnel for 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> on February 1.
No abnormal tempera-tures were identified in the surveillances preceding or following the missed one.
This was a minor licensee-identified item on which accept-able corrective action and reporting were accomplished.
No previous violation corrective actions which should have prevented this occurrence were identified.
The inspector had no further questions.
87-007-00.
Failure to surveill the containment sump volume every 12
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hours.
This was missed once for a 17 hour1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br /> period on February 11.
There were no safety consequences.
This was a minor licensee-identified item on which acceptable corrective action and reporting were accomplished.
No previous violation corrective actions which should have prevented this occurrence were identified.
The inspector had no further questions.
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4.
Followup of Actions on SALEM ATWS Events (Generic Letter 83-28)
An NRC Region I inspector reviewed the licensee's letter dated November 8, 1983, " Generic Letter 83-28 120-day Response" for Millstone Unit No. 3.
This letter was initiated during construction of the facility.
In response to Action Items 3.1, 3.1.2, 3.2.1, and 3.2.2, the licensee committed to review vendor and engineering recommendations, incorporate them in the Technical Specifications, and develop test and maintenance procedures to assure post-maintenance operability testing of Reactor Trip System and all other safety-related components prior to putting them into service.
The letter indicated that the licensee would develop appropriate procedures and schedules for in-dependent testing of the undervoltage and shunt trip features of the Reactor Trip Breaker Attachment, in response to Action Item 4.5.1, in a time frame consistent with the development of the Technical Specifications.
Facility license NPF-49, Item 2.c(14), dated January 31, 1986, required the licensee to submit responses and implement the requirements of GL 83-28, Salem ATWS Events on a schedule as mentioned in the above letter, and their subse-quent commitments.
The inspector determined that the licensee's responses to Action Items 3.1.1, 3.1.2, 3.2.1, 3.2.2, and 4.5.1 were incomplete because the licensee submittals did not show implementation of their commitments as outlined in their letters.
The licensee's letter dated May 13, 1986, Attachment II, indicates the status of the above Action Items as " Closed" based on their formal communication with the NRC staff.
However, the licensee could not provide documentary evidence to support resolution of the Action Items and their acceptance by the NRC staff.
The NRC Region I staff is performing a Safety Evaluation of the licensee's responses to GL 83-28 Action Items.
The importance of expeditious licensee action to provide the NRC with a completed response to the above items for the Safety Evaluation Report was discussed with the licensee.
The licensee is conducting an interdisciplinary document review and agreed to provide the NRC Region I staff with an updated response to the above items by March 21, 1987.
This issue relates to information submitted.
Licensee testing of the trip
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breakers separately tests the shunt and undervoltage trips.
Previous NRC resident inspection of that testing identified no inadequacies.
5.
Fitness for Duty Program The licensee has implemented a fitness for duty program applicable to em-ployees with unescorted access to nuclear facilities.
An expansion of North-east Utilities Policy and Procedure (NUP) 90, Alcohol; and NUP 91, Drug Abuse; the program requires annual drug testing and became effective February 10, 1987.
Drug testing will be scheduled with annual physicals beginning March 10, 1987.
For those employees not requiring annual physicals, drug testing
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will also be scheduled on an annual basis.
All contractors and suppliers with unescorted access to nuclear facilities are scheduled to be subject to similar test requirements by June 1, 1987.
Mandatory drug testing has been required for all newly hired personnel since July 1, 1986 and for all transfers to a nuclear facility since August 18, 1986.
Testing for cause, that is, for employees who appear to be unfit for duty, has been used since January 1, 1984 and will continue.
A refusal to submit to testing has been, and will continue to be, considered equivalent to posi-tive test results.
Testing will be performed by a qualified laboratory.
Sample elimination will not be witnessed but will be subject to strict control.
Analysis will be sensitive to amphetamines, barbituates, valium, cocaine, opiates, darvon, THC, PCP, codeine, heroin and morphine.
Positive results from initial screening will result in confirmatory testing using gas chromotography and mass spec-trometry techniques.
A quality assurance and audit program will be instituted utilizing independent evaluators.
Positive confirmatory testing will result in revocation of unescorted access to nuclear facilities based on failure of physical examination requirements.
The new fitness for duty program is currently in the implementation process.
The underlying philosophy of the policy is that the employee is ultimately responsible for a safe, productive work environment.
That responsibility in-cludes both self fitness and fitness of others.
The inspector had no further questions on this matter.
6.
Allegation RI-86-A-077, Class IE Battery Space Masonry Walls An allegation that masonry walls enclosing the Class IE Battery spaces had not been properly designed and constructed resurfaced during this report period.
The alleger was concerned that there were no records of slump tests performed on the mud (mortar) and no records of calculations made to determine load withstanding capabilities.
The inspector found that slump tests are not required for mortar, only for high density concrete.
The inspector also obtained battery wall design cal-culations along with specifications and references and forwarded these to NRC civil / structural engineers in the Regional Office for review.
The allegation remains open pending completion of that review.
7.
Licensee Actions on Previously Identified Items a.
(Closed) UNR Item 50-423/84-07-01, Management Control of PDCRs This unresolved item relates to design change control process flaws identified by construction errors and inadequate operational procedures for the Post Accident Sampling Systems' at Haddam Neck and Millstone 1.
The modifications were completed under the Plant Design Change Request
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(PDCR) system.
On December 13, 1984, an order modifying the Haddam Neck License was issued as a result of circumstances associated with the failure of the reactor cavity seal on August 21, 1984. At issue was management control of the PDCR process.
After extensive review, the NRC closed this issue for Haddam Neck as documented in T.E. Murley's letter-of April 28, 1986 to J. F. 0peka.
Since the PDCR system is corporate in nature, the extensive review for_Haddam Neck applies to this issue at Millstone 3.
Also, design change control is addressed during routine NRC inspection of plant changes.
The inspector had no further questions.
b.
(Closed) UNR Item 50-423/86-26-01, PDCR Documentation This unresolved item relates to an inadequate PDCR pagination and iden-tification of attachments prior to PDCR transfer to the nuclear record facility (NRF).
The licensee's PDCR Task Force, after reviewing this-item, initiated changes to the PDCR procedure and form to incorporate the page and PDCR number on every page of the package and identify re-lated attachments.
The inspector reviewed Revision 6 to ACP-QA-3.10 confirming that the PDCR
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number and the Revision number are to be placed on each page of the PDCR.
This revision, dated November 21, 1986, to be fully implemented within 90 days, resolves this item.
PDCR documentation will be routinely re-viewed during normal inspection.
c.
(Closed) UNR Item 50-423/87-02-01, Pump Suction Pressure Oscillations This item (LER 50-423/87-004-00) is addressed in Report Detail 3.
8.
Reporting of Events In Inspection Report 50-423/86-21, a problem with the Millstone event reports to the NRC Incident Response Center was addressed.
The problem involved the use of the term " General Interest Item," a term not related to NRC Emergency Planning terms such as Unusual Event, Alert, etc.
Recently, the NRC Incident Response Center staff notified Region I that the problem still exists.
The inspector discussed this with Millstone station management on March 12, 1987.
The licensee plans to make short term and long term corrections. In the short term, they will bring the issue to the plant staff's attention by meeting with each unit's management.
The long term corrective actions :ould include:
1) revision of EPIP Form 4112-1 to replace " general interest event" with "10 CFR 50.72 (X)(Y)(Z)"; and 2) provide shift personnel with any necessary training.
This issue will continue to receive inspector attention to inprove the informa: ion transfer.
9.
IE Information Notice 87-08 " Degraded Motor Leads in Limitorque DC Motor Operators" This Information Notice was hand delivered to the licensee.
To date, in the results for Millstone Unit 3, no Limitorque Valve Operators with DC Motors have been identified.
This item will be re-examined during routine inspectio.
10.
Observation of Maintenance The inspector observed and reviewed preventive and corrective maintenance to verify compliance with regulations, use of administrative and maintenance procedures, compliance with codes and standards, proper QA/QC involvement, use of bypass jumpers and safety tags, personnel protection, and equipment alignment and retest.
The following activities were included:
Repair of "B" Service Water Pump Self-cleaning Strainer, February 25.
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Replacement of the Outboard Shaft Seal, "A" Turbine-Driven Feedwater Pump
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on March 3.
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Replacement of the Inboard Shaft Seal on the Motor-Driven Feedwater Pump on March 4.
-- Repair of "D" Feedwater Line Isolation Valves (3FWS*CTV41D).
No inadequacies were identified.
11. 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 restora-tion of equipment, and deficiency review and resolution.
The following tests were reviewed:
RPS Protection Set Cabinet No. 3 Operations Test, SP3443-C21, on February
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Inservice Test of "A" Motor-Driven Auxiliary Feedwater Pump on February
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Operational Test of "B" Emergency Diesel Generator on March 5.
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No inadequacies were identified.
12.
Correction of Chlorine Monitor Problems NRC Inspection Reports 86-21 and 86-28, along with four licensee LERS, dis-cussed recurring spurious Control Building Isolations (CBIs) caused by chlorine monitor actuations.
After several unsuccessful attempts to correct the problem, the licensee reoriented the detector probe from its horizontal position to a vertical one.
LER 86-40 projected the completion date of this modification to be December 31, 1986.
The inspector confirmed that the modi-fication was completed on December 23, 1986.
No design change control prob-lems were identified.
The inspector also verified, through discussions with Operations and I&C personnel, and review of the Shift Supervisor and Plant Incident Report logs, that no subsequent CBIs due to chlorine monitor actu-ations nave occurred.
The reorientation of the probe was suggested by the i
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detector vendor but was not indicated on the installation instructions.
The vendor has since revised its instructions to include a diagram of proper (vertical) probe orientation.
The inspector had no further questions on this matter.
13.
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.