IR 05000423/1990007
| ML20043E944 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 06/06/1990 |
| From: | Haverkamp D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20043E942 | List: |
| References | |
| 50-423-90-07, 50-423-90-7, NUDOCS 9006140022 | |
| Download: ML20043E944 (19) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No.:
50-423/90-07 I
Docket No.:
50-423 License No.
NPF-49 Licensee:
Northeast Nuclear Energy Company P.O. Box 270 Hartford, Connecticut 06141-0270 i
Facility Name: Millstone Nuclear Power Station, Unit 3 j
Inspection at: Waterford, Connecticut f
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Inspection Conducted:
March 20 - May 2, 1990
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Reporting Inspector:
Kenneth S. Kolaczyk, Resident Inspector, Millstone 3 Inspectors:
William J.- Raymond, Millstone Senior Resident Inspector Kenneth S. Kolaczyk, Resident Inspector, Millstone 3 Approved by:
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A (///fo bnald R. Haverkamp/061ef
'Date ReactorProjectsgection4A Division of Reactor Projects
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Inspection Summary:
Inspection on March 20 - May 2, 1990 (Inspection Report
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No. 50-423/90-07)
Areas Inspected:
-Routine onsite Inspection at Millstone 3 during normal and backshift
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work periods of plant operations; maintenance and surveillance; security; engineering and technical support; and safety assessment and quality verification.
Results:
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General Conclusions on Safety Perspective, Strengths or Weaknesses in the Licensee's Programs The licensee needs to focus resources on intake system weaknesses to assure timely resolution of problems.
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Violations
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Within the scope of this inspection, no violations were observed.
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900614002'2 900606 PUR ADOCK 05000423 O
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. Unresolved Items-
- One open item (89-14-02)'was closed which involved the return. to' service
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Executive Summary Plant Operations (Modules 71707,93702) Two reactor trips during this report period highlighted the continuing need to address intake system debris removal deficiencies.
Radiological Controls (Module 71707) The licensee has placed the steam generator blowdown monitor into continuous service for the first tine since the start of commercial operation in 1986.
Maintenance / Surveillance (Modules 61726,62703)
Improper installation of a linkage on a pressurizer siray valve caused the plant to begin a slow depressurization.
Security (71707) Routine review in this area identified no noteworthy findings.
Safety Assessment / Quality Verification (Modules 71707,90712,72702) Routine review in this area identified no noteworthy findings.
Engineering / Technical Support (Modules 37701,93702,62705) Additional actions to increase the reliability of the intake structure are needed in the area of galvanic protection for the structure.
Additionally, the licensee has concluded that an indepth study of the feedwater system will be performed to determine the cause of the feedwater line pressure surges.
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e'1 TABLE OF CONTENTS
Page 1.0 Persons Contacted..............................................
I 2.0 S umma ry o f Fa c i l i ty Ac t i vi t i e s.................................
3.0 PlantOperations(IP 71707/93.02)*.............................
3.1 Control Room Observations.................................
3.2 Plant Tours...............................................
3.3 Review of Plant Incident Reports (PIRs)...................
3.4 Reactor Trips Caused by Intake System Fouling.............
3.4.1 March 30 Reactor Tr1p..............................
3.4.2 April 16 Reactor Trip..............................
4.0 Radiological Controls (IP 71707)...............................
4.1 Concern Regarding Not Receiving a Whole Body Count........
4.2 (Closed) 50-423/89-14-02, Steam Generator Blovdown Radiation Monitor Restored................................
5.0 Maintenance / Surveillance (IP 62703/61726)......................
5.1 Observation of Maintenance Activities.....................
5.1.1 Pressurizer Spray Valve Failure on April 8, 1990...
5.2 Observation of Surveillance Activities....................
6.0 Engineering / Technical Support (IP 37701/93702/62705)...........
6.1 Plant Design Modifications................................
6.1.1 Intake System Fouling..............................
6.2 Waiver of Compliance Issued...............................
6.3 Feedwater Heater Relief Piping Damaged....................
6.4 NRC Regional Initiatives..................................
6.4.1 (Closed) Temporary Instruction 86-02, Inspection of General Electric AK-F-2-25 Breakers............
6.4.2 (Closed) Temporary Instruction 87-06, Diesel Generator Air Start Motor Lubrication Followup.....
7.0 Security (IP 717107).........................................
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8.0 Safety Assessment / Quality Verification (IP71707/90712/92702).........................................
8.1 Committee Activities......................................
8.2 Licensee Event Reports ( LERS).............................
8.3 Periodic Reports..........................................
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8.4 Management Meetings.......................................
- The NRC inspection manual inspection procedure (IP) or temporary instructions (TI) that was used as inspection guidance is listed for each applicable report section.
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Northeast Nuclear Energy Company
- Enclosure: NRC Region I Inspectioh Report No. 50-423/90-07 cc w/ enc 1:
W. D. Romberg, Vice President, Nuclear Operations R. M. Kacich, Manager, Generation Facilities Licensing D. O. Nordquist, Director of Quality Services S. E. Scace, Nuclear Station Director C. H. Clement, Nuclear Unit Director, Millstone Unit 3 Gerald Garfield, Esquire Public Document Room (PDR)
local Public Document Room (LPDR)
Nuclear Safety Information Center (NSIC)
NRC Senior Resident Inspector State of Connecticut
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bec w/ enc 1:
Region I Docket Room (with concurrences)
Management Assistant, DRMA (w/o enclosures)
DRP Section Chief J. Shediosky, SRI, Haddam Neck D. Jaffe, PM, NRR W. Rayna nd, SRI, Millstone bec w/ Executive Summary only:
J. Wiggins, DRP W. Hodges, DRS W. Johnston, DRS J. Durr, DRS R. Gallo, DRS
M. Knapp, DRSS R. Cooper, DRSS R. Bellamy, DRSS J. Stolz, NRR/PDI-4 j
RI:DRP RI:DRP RI:DRP RI:DRP Kolaczyk Raymond Haverkamp Wenzinger 05/ /90 05/ /90 05/ /90 05/ /90
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0FFICIAL RECORD COPY IR MILL 3 90-07 - 0002.0.0 05/30/90 l
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DETAILS 1.0 Persons Contacted Interviews and discussions were conducted with. licensee staff and management during the report period to obtain information pertinent to the
areas inspected.
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Inspection findings were discussed periodically with the supervisory and management personnel identified below.
S. Scace, Nuclear Station Director, Millstone Station C. Clement, Nuclear Unit Director, Millstone Unit 3
- M. Gentry, Operations Manager, Millstone Unit 3
- R. Rothgeb, Maintenance Manager, Millstone Unit 3
- J. Harris, Engineering Manager, Millstone Unit 3 S. Sudigala, Senior Engineer D. McDaniel, Engineering Supervisor, Millstone Unit 3 R. Sachatello, Radiation Protection Supervisor, Millstone Unit 3 M. Pearson, Operations Assistut B. Enoch, Instrument and Controls Manager, Millstone Unit 3 R. Joshi, Licensing, Millstone Unit 3
- B. Beckman, Ir.strument and Controls Supervisor, Millstone Unit 3
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- Attendee at post-inspection exit meeting on May 24, 1990.
2.0 Summary of Facility Activities
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Balance of piant problems continued to reduce plant reliability during this report period. On March 30, the reactor was manually tripped from 80% power when seaweed buildup on the traveling screens in the intake structure caused two circulating water pumps to automatically trip.
The plant was maintained in mode 3 for two weeks while repairs / modifications were performed on the intake stru::ture. During the shutdown period, on April 8, a primary coolant system pressure transient occurred, when the
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air supply line to the pressurizer spray valve PCV-455B failed causing the valve to open, depressurizing the plant. A containment entry was per-formed to close the valve (manually) and primary coolant system pressure was stabilized at 432 psig, four hours after the transient began.
On April 15, the plant was taken critical and power ascension began. On April 26, while conducting trash raking operations, the plant was manually tripped from 51% power when the B circulating water pump automatically tripped due to high differential pressure across it's traveling screens.
Additional modifications were subsequently performed on the intake structure and the plant was restarted on April 20.
For the next six days, power was held at 30% while intake system operation was monitored._ Power ascension began on April 27, with full power being reached on April 28.
The plant remained at 100% of rated thermal power for the rest of the report perio j
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The inspection activities during the report period included 124 total
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hours of inspe Is iuring normal working hours.
In addition, the review of plant operat ac9 as routinely conducted during the periods of backshift (even1ng' shifts) and deep backshifts (weekends and midnight shifts).
Inspection coverage was provided for 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br /> of backshifts and 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> of deep backshifts.
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i From April 10 to April 25, a team consisting of regional, resident and contractor personnel examined the Millstone Unit 3 emergency operating i
procedures. Additionally, a mid-SALP inspection of licensee activities
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was conducted on,ite by resident, regional and headquarters personnel on April 17-18.
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3.0 Plant Operations 3.1 Control Room Observations The inspector reviewed plant operations from the control room and reviewed the operational status of plant safety systems to verify safe operation of the plart in accordance with the requirements of technical specifications aid plant operating procedures. Actions taken to meet technical specification requirements when equipment was inoperable were reviewed to verify the limiting conditions for operations were met.
Plar t logs and control room indicators were reviewed to identify chan ges in plant operational status since the last review and to verif3 the changes in the status of plant equipment was properly communicated in the logs and records.
Control room instruments were observed for correlation between channels, proper functioning, and conformance with technical specifications. Alarm conditions in effect were reviewed with control room operators to verify proper response to of f-normal conditions and to verify operators were knowledgeable of plant status.
Trainees who were manipulating reactor controls were under instruction by licensed operators.
Operators were found to be cognizant of control room indications and plant status during normal working hours and backshift observations.
Control room manning and shift staffing were reviewed and compared to technical specification requirements.
No inadequacies were identified.
3.2 Plant Tours The inspector observed plant operations during regular and backshift tours of the following areas:
Control Room Containment
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Vital Switchgear Rooms Diesel Generator Rooms
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Turbine Building Intake Structure Auxiliary Building ESF Building
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- During plant tours, logs and records were reviewed to ensure compliance with station procedures, to determine if entries were
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correctly made, and to verify correct communication and equipment status.
No inadequacies were noted.
3.3 Review of Plant Incident Reports (PIRs)
The plant incident reports (PIRs) listed below were reviewed during the inspection period to (i) determine the significance of the events; (ii) review the licensee's evaluation of the events; (iii)
verify the licensee's response and corrective actions were proper; and, (iv) verify that the licensee reported the events in accordance with applicable requirements, if required. The PIRs reviewed were:
number's 90-48 thru 90-68.
The following PIRs warranted inspector
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followup:
90-56, 90-60 and 90-62 and are discussed in sections
3.4.1, 3.4.2, and 5.1.1, respectively.
i 3.4 Reactor Trips Caused by Intake System Fouling On March 30 and April 16, two reactor trips were manually initiated
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by operators in anticipation of automatic turbine / reactor trips due to loss of condenser vacuum.
Both events are described in below.
3.4.1 March 30 Reactor Trip A manual reactor trip from 80% power was initiated on March 30 in anticipation of an automatic turbine / reactor trip due to low con-denser vacuum.
The loss of vacuum occurred when the "A" and "B" circulating pumps automatically tripped when 30" of differential pressure was reached across their respective traveling screens. The plant responded as designed except the pipe to the feedwater inlet thermal reliefs on the 1A and IC feedwater heaters severed causing steam discharge into the turbine structure. No personnel or equip-ment was damaged by the steam discharge; however, an intermittent ground appeared on bus 32-2P which is located in the turbine build-ing. The relief valve piping failure is discussed in further detail in section 6.3 of this inspection report. The inspector was notified of the manual reactor trip by the licensee duty officer and went to the control room to observe operator actions and evaluate plant response to the event.
Inspector review included interviews with operations, instrumentation and controls technicians and maintenance personnel.
Additionally, the inspector reviewed control room annunciator status and operator
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response to the plant conditions. The following sequence of events was noted by the inspector:
13:22:49 Traveling Screen Differential Pressure-High 13:23:22 Boric Acid to Charging Pump Suction-Partial
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13:24:14 Main Steam Stop Valve-Partial i
13:27:25 Circulating Pump A Trip 13:28:3 Traveling Screen Differential Pressure-Normal 13:28:7 Traveling Screen Differential Pressure-High 13:28:49 Circulating Pump B Trip 13:28:55 Manual Reactor Trip Prior to the trip, the licensee had removed the automatic screen wash system from service to allow replacement of a broken fiberglass elbow joint on the A train of the screen wash system.
To provide the requisite seaweed screen wash removal capability, a backup wash
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system was established by plant personnel through use of four 1.5 inch fire hoses. According to plant personnel, this method of screen cleaning was effective, however, the seaweed was discharged into the fish trough which empties into the west end of the intake structure
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rather than through the normal seaweed chute, which discharges into the east end of the structure.
Apparently the seaweed accumulated in the low flow pocket area that exists on the west end of the intake and was sucked back into the A water box.
This greater than normal seaweed flow clogged the lower ends of the traveling screens before they rotated upward to be cleaned by the boses.
This caused the differe'ntial pressure across the screens to rapidly increase to the circulating water pump automatic trip setpoint of 30".
Further complicating the seaweed removal process was the fact that the A rotating screen stopped rotating for a period of time possibly because of mechanical binding or tripping on electrical overload due
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to frequent start / stopping while in manual operation.
Although it was restarted prior to the trip, differential pressure across the screen had already begun to rapidly increase to the trip setpoint.
In response to the increase in differential pressure across the screens, operators began to rapidly reduce power at the maximum design rate of 5%/ minute.
However, when the B circulating water pump tripped, an automatic trip was inevitable since reactor power was still at 80% of full power, or 30% above the 50% turbine / trip / reactor trip interlock.
Post-trip review of the main control board annunciators and sequence of events printout revealed the plant responded as designed.
Operator response to the event was determined to be good.
Inspector review of the events preceding the trip revealed that removal of the screen wash system from service was a well planned evolution.
The inspector observed a preshif t briefing of the evolution and found it to be acceptable.- The evolution was performed during favorable sea conditions - high tide with an easterly wind -
both conditions should have directed seaweed away from the intate structure. Also, there were ample personnel in the screen house with supervision.
The inspector had no further questions on the
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preplanning that was performed prior to removing the screen house system from service or on the reactor trip.
No inadequacies were noted.
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3.4.2 April 16 Reactor Trip The reactor was manually tripped from 51% reactor power on April 16 when the B circulating water pump automatically tripped due to
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high differential pressure across it's travelling screen.
Prior to the trip, trash raking operations were in progress on the B bay. The
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A intake bay was tagged out of service to. facilitate repairs and
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operators were in the process of increasing reactor power at 3% per hour in recovery from the March 30 trip.
The licensee concluded that the trash rake cleaning operations which consist of dragging a trash rake (a fork-like object) over the trash grates, must have pushed debris through the two-inch wide space that exists between the trash
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grates and onto the B bay traveling screens causing them to overload.
When the B pump tripped, operators at first began to reduce power and did clear the P-9 turbine trip / reactor trip setpoint.
Therefore, the possibility existed that operators could have tripped the turbine without undergoing a corresponding reactor trip.
However, condenser vacuum in the A water box began to rapidly decrease due to the loss of circ'ulating water flow. Since a further loss of condenser vacuum would have locked out the steam dumps, the operators tripped the plant rather than risk the possibility of an automatic trip on high pressurizer pressure or level.
Inspector review of the event included review of the sequence of events, interviews with operators and attending c1 'mi post-trip licensec critique of the trip.
The following sevnce of events was noted by the inspector:
12:0:16 Traveling screen differential pressure high 12:0:43 Circulating water pump B trip s
12:0:50 Traveling screen differential pressure normal 12:1:14 P-9 turbine _ trip / reactor trip interlock clear 12:1:15 Reactor trip.
The plant responded to the event as designed and operators stabilized the plant in mode 3 - hot standby.
To prevent recurrence of the event, the licensee performed the L
following actions:
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Stationed a dedicated individual at the intake structure who
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- Improved trash raking efficiency by lengthening the rake teeth.
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Modified the traveling screen method of operation whereby the
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screens can manually be placed in fast speed while raking.
It was theorized by the licensee that by placing the screens in fast speed before raking, any debris that is pushed through the grates by cleaning will be quickly removed by the faster
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rotating screens prior to excessive loading.
Prior to this modification, the screens would automatically switch from slow -
5 feet per second, to fast - 20 feet per second only after they begin to foul.
The inspector noted no inadequacies regarding the modifications to the intake structure, or in the operator and plant responses to the
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trip.
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4.0 Radiological Controls 4.1 Concern Regarding Not Receiving a Whole Body Count The inspector was contacted by an individual who expressed a concern that 5e did not receive a whole body count when he left the Millstone site in November.
The individual worked as an electrician in the production test department at Millstone and Connecticut Yankee and was issued dosimetry; however, he did not wear a respirator.
The inspector informed the individual that NRC requirements do not specifically require the licensee to give whole body counts to individuals who do not wear respiratory equipment.
However, as a t
matter of routine policy, the licensee normally requires a whole body count upon an individual's entrance to a site to establish a baseline level and when an individual leaves the site as a second check to ensure no uptake was received.
The inspector suggested that the individual contact the health physics office if he wanted a whole body count for his own benefit, Additionally, the inspector informed the individual that he would notify the health pnysics manager at Millstone of the conversation and inform the manager of the worker's I
desire to obtain a whole body count.
When the inspector notified the health physics manager of the conversation, the health physics office contacted the individual and l
administered a whole body count.
No activity uptake was identified.
During a' discussion with the inspector, the health physics manager reiterated the fact that the individual was not issued a respirator l
during his assignment at Millstone, therefore, per Millstone procedures, receiving a whole body count on exit from the facility is not required. The inspector has no further questions on this issue.
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4.2 (Closed) 50-423/89-14-02, Steam Generator Blowdown Radiation Monitor Restored
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Inspection Report 50-423/89-14 documented that the licensee has never placed the steam generator blowdown radiation monitor in service l
since the start of commercial operation in April of 1986.
Additionally, the semi-annual effluent reports that have been submitted to the NRC since December of 1986 have not reported an updated status of the monitor.
After the issue was discussed with licensee management, the licensee completed the required engineering reviews and modifications neces-sary to restore the monitor to service. Additionally, the next semi-annual effluent report that covered the July to December 1989 time-
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frame was updated to. inform the NRC of the correct status of the ironitor. On March 13, 1990, the monitor was restored to service and declared operable.
To ensure other effluent monitors are restored to service in a reasonable time, the Millstone Unit 3 chemistry super-visor informed the inspector that monthly audits of the effluent monitor's will be performed by the Berlin radiological assessment branch. The inspector noted the supervisor's comments and had no further questions at this time. lhe status of effluent monitors will be reviewed in future inspections.
5.0 Maintenance / Surveillance I
5.1 Observation of Maintenance Activities The inspector observed and reviewed selected portions of preventive and corrective maintenance to verify compliance with regulations, use of administrative and maintenance procedures, compliance with codes
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and standards, proper QA/QC involvement, use of bypass
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jumpers and safety tags, personnel protection, and equipment i
L alignment and retest. The following activities were included:
-- AWO M3-90-06819 'A' Diesel Generator Monthly Maintenance, May 2, 1990
-- AWO M3-90-01343, Test Tank 'B' Educator Stop Valve Replacement, May 1, 1990
-- AWO M3-90-06872, Pressurizer Spray Valve Failure, April 8,1990 No inadequacies were identified.
5.1.1 Pressurizer Spray Valve Failure on April 8,1990 With the reactor in hot standby conditions at normal operating temperature and pressure on April 8 during maintenance of the circulating water intake structure, the pressurizer spray valve PCV-455B failed open at about 4:30 p.m. and caused a slow depressurization of the reactor coolant system (RCS). After verifying that there were no indications of a loss of primary coolant and that the spray valve was the cause for the depressurization,
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plant operators decreased reactor coolant system temperature using the steam dump valves to maintain RCS temperature and pressure within technical specification limits. An emergency containment entry team was assembled and dispctched.
The cooldown was terminated at 9:00 p.m. when plant personnel manually secured the air supply to the
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spray valve positioner inside the containment.
Plant operators
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initiated actions to restore the RCS to normal operating conditions.
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The event was reported to the resident inspector at 9:30 p.m. on April 8.
The inspector reviewed the plant response after the event, along with
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the licensee investigation and corrective actions. The cooldown was completed using the normal operating procedure, OP 3208, and subcooling margins were maintained within procedure limits.
The total RCS cooldown was from 550 degrees F to 385 degrees F over a four-hour period and the cooldown rate was maintained at less than 50 degrees F/ hour.- No plant operating limits were exceeded and the reactor remained shutdown with adequate margin (all control rods inserted with a cold shutdown boron concentration) throughout the transient.
Inspector review of the plant response and the operator actions identified no inadequacies.
The spray valve controller is a Bailey Controls System Model AP4 positio'ner that was installed new during the 1989 outage as PDCR MP3-89-049.
Licensee investigation determined that the positioner failed when a linkage in the operating arm of the controller fell off, which allowed the spray valve to fail open.
Licensee review concluded the linkage was inadequately installed during the 1989 design change when a stud that connects the linkage was installed backward and inadequately secured to the valve stem clamp.
Licensee actions included re-installing a longer stud in the correct orienta-
tion with a nylon lock nut.
The stud arrangement was also checked
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and corrected on the second spray valve, PCV 4550. The licensee plans to also check the installation on 7 other valves in the plant using the same model controller.
The inspector reviewed the condition of two controllers installed on the reactor plant component cooling heat exchanger supply valves (3CCP-TV32BP and TV32CP) and found the linkage connections secure.
No inadequacies were identified in the licensee's investigation or planned corrective actions.
The event did not meet the reporting criteria of 10 CFR 50.72 or 50.73. The licensee submitted a Special Report by letter dated May 7, 1990,.for information due to interest in the event. The inspector had no further questions regarding this matter.
5.2 Observation of Surveillance Activities The inspector observed portions of and reviewed surveillance tests to assess performance in accordance with approved procedures and Limiting Conditions of Operation, removal and restoration of i.
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- equipment, and deficiency review and resolution.
The following tests were reviewed:
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-- SP 3646A.2 'B' Emergency Diesel Generator Operability Test, dated v
May 2, 1990
-- SP 3646B.1 Emergency Generator Fuel Oil Transfer Pump 1A Operational Readiness Test, dated May 2,1990
-- SP 3616A.1 Main Steam Valve Operability Test. dated April 26, 1990
-- SP 3443A21 Protection Set 1 Operational Test dated April 23,1990 No inadequacies were noted.
6.0 Engineering / Technical Support 6.1 Plant Design Modifications 6.1.1 Intake System Fouling The poor reliability of the intake system has caused several plant trips since the start of commercial operation of unit 3 in April of 1986.
Seaweed infusion that occurs during storms or unfavorable sea conditions collects on the traveling screens in the intake structure and can' rapidly exceed the screen wash cleaning capability. This eventually causes a circulating water pump to trip on high differential pressure.
In response to the earlier plant trips caused by intake system unreliability, the licensee (1) resequenced the performance of maintenance on the intake structure to times of low seaweed growth during the winter and summer months; (2) developed a call-in system whereby additional personnel would work at the intake structure during periods of high wind or seaweed loading conditions; (3)
performed modifications to the intake structure such as installation of a trap door on the trash chute to be used when the seaweed conveyor is overloaded. Additionally, a task force was formed to look at long-term improvements to the structure.
Several task force recommendations were accepted and developed into a draft project assignment.
Their recommendations included trash rake modifications, improved intake structure lighting, and increased rotating screen speed.
Review of the March 30 and April 16 trips reveals that the scope of the project assignment may be inadequate.
Specifically, prior to the March 30 reactor trip, the A traveling screen stopped rotating, al-though it quickly resumed operation, resulting in additional seaweed buildup on the screen causing differential pressure across the screen to increase to the A circulating water pump trip setpoint.
Post-trip operation of the screens on this bay caused mechanical screen damage to occur.
Subsequent drain down and inspection of the bay revealed l
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1 that the supporting framework had undergone severe galvanic corrosion causing structural steel members to detach from the stainless steel
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bolts.
It appears that this corrosion caused the structure to sag, i
which eventually caused the screen to bind in an area of reduced
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clearance.
The inspector noted that damage was also sustained to the 'B'
rotating screens when bolts which attach the screens to the rotating i
chain became dislodged. This allowed the screens to become damaged
on the intake bay supporting framework. Additional problems noted included plugged spray nozzles, misdirected spray flow and broken
screen wash nozzles. Although the licensee flushed out the spray nozzles, redirected fish spray flow and replaced broken nozzles, no method of improved maintenance procedures is currently being examined by the project assignment. Additionally, the inspector noted that the project assignment did not have any provision for addressing the galvanic corrosion concern.
The inspector concluded the structural integrity of the intake system needs to be assured to enhance intake system reliability. At the exit meeting the inspector discussed his observations with the li-censee Engineering Supervisor.
The engineer noted the inspector's comments and stated that as a result of an inspection conducted on the A bay, the draft project assignment will be revised to address the galvanic corrosion problem.
Licensee actions to further develop and implement corrective actions in this area will be reviewed in future routine inspections.
6.2 Waiver of Compliance Issued On March 31, the inspector was informed by the licensee that a temporary waiver of compliance would be sought to allow an April 2 plant restart. Obtaining the waiver of compliance was necessiteted by the March 30, 1989 reactor trip which occurred while the plant was operating in the technical specification 3.7.12.1(c) fire suppression water system action statement. While operating in this statement, a mode change is not allowed.
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The licensee entered into this action statement when a crack in a cast iron " press fit" elbow joint was discovered in a section of the Millstone 3 fire main. When the crack was located on March 29, 1990, the licensee isolated the leak and established a backup fire suppression system within the required 24-hour time period. The compensatory actions consisted of (1) rigging fire main jumper hoses around the isolated section of pipe to supply fire main water to the affected engineered safety feature (ESF) and fuel buildings; l
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(2) staging additional sections of hose in hose station 4 that is located adjacent to the affected reserve station transformer (RSST);
and, (3) establishing a fire watch at the RSST.
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In an April 1, 1990 letter to the NRC, which requested a waiver of compliance, the licensee stated that the compensatory measures that were taken established an equivalent level of fire protection for the effected areas.
Therefore, the health and safety of the public would not be jeopardized by a plant startup.
Further, the licensee stated that the NRC had recognized in Generic Letter 87-09, dated February 21, that several technical specifications which currently
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allow operation in an action statement indefinitely should also state that technical specification 3.0.4 and 3.0.3 are not applicable.
Insertion of this statement would allow a licensee to change modes as long as an equivalent level of protection was taken in accordance with the technical specification action statement.
After review of the compensatory actions taken by the licensee in response to the disabled fire main header, the NRC staff granted a temporary waiver of compliance on April 2.
The waiver of compliance was to remain in effect until an emergency technical specification change submitted by the licensee on April 2, which would permanently modify the specification, was issued.
Becauselof difficulties encountered while repairing the intake
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structure, plant restart was delayed until April 15. Work on the fire main was initially slowed since replacement parts had to be ordered.
Consequently, the fire main was not declared operable
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until April 20.-
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Review of this event by the inspector revealed that the licensee had taken the proper compensatory actions when the operability of the fire main was questionable.
However, a temporary waiver of compliance would not have been required had the licensee conducted a timely review in response to Generic Letter 87-09.
Through discussion with licensing personnel, the inspector was informed that the response was delayed since a
page-by page review of the technical specifications was being conducted for all four Northeast Utilities plants, which would not only identify where the technical specification was too restrictive, (i.e., no insertion of the T.S. 3.03, 3.04 statement was made), but also where a mode change is currently allowed but should not be.
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The inspector noted that the licensee's review scope exceeded the actions requested by the generic letter, and would result in a
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further improvement to plant safety. The inspector acknowledged that a page-by page review of the technical specifications is time consuming, however, three years to conduct a review appears lengthy
for a single plant and is not responsive to NRC initiatives.
Timely response to NRC action items will continue to be reviewed in future resident inspections.
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6.3 Feedwater Heater Relief Piping Damaged Following the March 30 reactor trip, the inlet relief valve piping on the 1A and IC feedwater heaters severed causing steam discharge into the turbine building.
The turbine building was evacuated as a precaution and operators isolated the heater which thereby removed the majority of the steam source.
Inspection of the piping revealed the failure point was in the weld heat affected rone where the relief
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r piping was welded into the boss on the heat exchanger shell. This
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was identical.to the previous failure which occurred on March 9 when the 1A and IB relief valve piping failed.
Prior to ths March 30 reactor trip, both inlet pipes were magnetic particle tested and no inadequacies were identified.
Post-trip examination of the IB relief valve piping did not identify any indications.
i According to a licensee system engineer, the IC feedwater heater moved longitudinally 1/8" during the-transient. This is in comparison to the 1/4" and 1/2" movement observed on the 1A and IC feedwater heaters during the March 9 reactor trip. Magnetic particle examination of the inlet and outlet nozzles on the 1A feedwater heater which has the shortest run of pipe to the isolation valves did not identify any indications.
The licensee has preliminarily concluded that the relief valve piping was damaged when a pressure surge was created during a feedwater isolation which normally occurs when Tave reaches 564 degrees F.
According to the licensee, the configuration of the relief valve piping, shaped like a letter C, results in pressure surges within the feedwater pipe which could not be quickly distri-buted. Therefore, pressure built up at the inlet pipe to inlet head transition point caused the failure. Accordingly, the licensee reshaped the inlet pipe to the relief valve to allow a smoother t
transition and, therefore, decrease the resulting flow restriction.
Additionally, three hangers located on the inlet side to the relief valves were removed to increase piping flexibility.
Bench testing conducted on the relief valves revealed that both the A and B reliefs were damaged due to excessive chattering.
Both valves were subsequently repaired and reinstalled.
l Through conversations with the system engineer, the inspector was l
informed that the Berlin engineering staff is evaluating the pipe to
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determine the failure mechanism. An initial reviaw cf the licensee modifications was also conducted by a consultant who deterr.iined it i
to be satisfactory.
The licensee is also considering the use of an l
architect engineer to model the feedwater system from the regulating
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valves to the 1A, 1B, and 1C heaters. This evaluation would be in addition to other inspections of the system that were planned as a result of an earlier water hammer event.
This evaluation would b-used to determine if the licensee's present fix is adequate; to recommend a permanent fix if necessary; and to evaluate the effect of the pressure surges on the remainder of the feedwater piping.
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The inspector noted that the licensee's immediate corrective actions appear to be appropriate; the effectiveness of the licensee's long j
term review of the feedwater system will be reviewed in future resident inspections.
6.4 NRC Regional Initiatives 6.4.1 (Closed) Tem >orary Instruction 86-02, Inspection of General
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Electric AK R-2-25 breakers As a result of failures of General Electric AK-F-2-25 breakers that had occurred at the Pilgrim Nuclear Power Station, the NRC regional office issued a temporary instruction (TI) that directed resident
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inspectors to review what type of preventive maintenance testing is performed on these breakers that are located in systems that are important to safety.
A review of breakers at Millstone Unit 3 determined that no GE AK-F-2-25 breakers are currently installed in
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important-to-safety applications.
Therefore, this TI is closed.
6.4.2 (Closed) Temporary Instruction 87-06, Diesel Generator Air Start Motor Lubrication Followup
This installation does not apply to Millstone Unit 3 since diesel start i's achieved through direct air injection into the cylinder head.
Therefore, this item is closed.
7.0 Security (IP 71707)
Selected aspects of site security were verified to be proper during inspection tours, including site access controls, personnel searches, personnel monitoring, placement of physical barriers, compensatory
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measures, guard force staffing, and response to alarms and degraded conditions.
8.0 Safety Assessment / Quality Verification
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8.1 Committee Activitics
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The inspector attended meetings of the Plant Operations Review Committee (PORC) and the Nuclear Review Board (NRB).
The inspector noted by observation that committee administrative requirements were met for the meetings, and that the committees discharged their functions in accordance with regulatory requirements.
The inspector observed a thorough discussion of matters before the PORC and NRB and a good regard for safety in the issues under consideration by the committees. No inadequacies were identified.
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8.2 Licensee Event Report Review l
Licensee Event Reports (LER) submitted during the report period were l
reviewed to assess LER compliance with 10CFR50.73 reporting
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. requirements, and to ' determine if there were generic implications or further information was required.
Selected corrective actions were-reviewed for implementation and thoroughness. The LERs reviewed were: 30-09-00 and 90-11-00.
No inadequacies were noted.
8.3 Periodic Reports s
Upon receipt, periodic reports submitted. pursuant to technical specifications were reviewed. This review verified'that the reported-information was valid and included the required NRC data. The
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. inspector also ascertained whether any reported information should be classified as an abnormal occurrence.
The following report was reviewed:
-- March Monthly Operating Report 8.4 Management Meetings Periodic meetings were held with station management' to discuss
. findings as a result of this and other inspections conducted during this period. On April 25, a formal exit was held with the licensee to discuss the findings during the E0P team inspection. A summary of findings was also discussed at the conclusion of the inspection.
No proprietary information was covered within the scope of this inspection. No written material was given to the licensee during-the inspection period.
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