IR 05000334/1990007
| ML20042F353 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 04/24/1990 |
| From: | Young F NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20042F350 | List: |
| References | |
| 50-334-90-07, 50-334-90-7, 50-412-90-06, 50-412-90-6, NUDOCS 9005080220 | |
| Download: ML20042F353 (15) | |
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i-U. S. !4UCLEAR REGULATORY COMMISSION
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L REGION 1 Report Nos.
50-334/90-07 License: DPR-66 50-412/90-06 NPF-73 Licensee:
Duquesne Light Company Bne Oxford Center 301 Grant Street Pittsburgh, PA 15279-Facility Name: Beaver Valley Power Station, Unii.s 1 and 2 Location:
Shippingport, pennsylvania t
Dates:
February 17 - March 30, 1990 Inspector:
J. E. Beall, Senior Resident Inspector P. R. Wilson, Resident Inspector
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4/04]d Approved by:
h Francis Yo g Ay%ing Section Chief Date Reactor Pro ts M tion No. 4B Division of Reactor Projects Inspection Summary This inspection report documents routine and reactive inspections of station activities during day and backshift hours including:
plant operations; radiological protection; surveillance and maintenance; emergency preparedness; security; engineering and technical support; and safety assessment / quality verification.
Results Overall, the facility was operated safely. No violations were identified.
Improvements in Unit 1 general housekeeping were observed (Section 2.1).
A review of Unit 1 operation above 100*4 power was conducted. The licensee's corrective actions were reviewed and determined to be adequate (Section 2.3.1).
A preliminary review of Unit 1 trip resulting from water in instrument air lines was documented (Section 2.3.2).
Weakness was identified in the licensee's Quality Concern Resolution Program.
(Section 8.2).
Four previous l
open NRC items were reviewed and closed during this inspection (Section 9).
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An Executive Summary follows.
9005080220 900427
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PDR ADOCK 05000334
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EXECUTIVE SUMMARY
Plant Operations The inspectors reviewed a Unit 1 event concerning operation above 10D% power.
The inspectors found that the licensee's corrective actions were adequate to prevent recurrence. The inspector conducted a preliminary review of a Unit I reactor trip resulting from moisture in the instrument air
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lines to the Main Feedwater Regulating Valves.
Evaluation of this event was still in progress at the end of the inspection period.
Significant improve-ments in Unit 1 general housekeeping were observed, t. -
Radiological Protection Routine review of the area identified no noteworthy deficiencies.
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Surveillance and Maintenance Both surveillance testing and maintenance activities were observed.
No deficiencies were identified.
Emergency Preparedness Routine review of this area identified no deficiencies.
Security Routine review of this area identified no deficiencies.
Eng.ineering and Technical Suaport The inspectors reviewed previously identi-TTed NRC concerns. including Jnit 1 emergency diesel generator reliability, r
control of measuring and test equipment, root cause evaluation of a 1988 overcooling event, and the adequacy of the ventilation system in the Unit 1 Safeguards Building.
The inspectors found that the. concerns had been adequately resolved..
Safety Assessment / Quality Verification The inspector found the licensee's i
investigation, root-cause analysis, and corrective actions associated with a generic casting defect in some safety related motor operated valves to be noteworthy. An inspection of the licensee's Quality Concern Resolution Program was conducted. The inspectors found that the program was generally effective, however, weaknesses were identified in the timeliness and scope of some of the licensee's investigations.
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TABLE OF CONTENTS
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' Summary.of Facility Activities...............
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Plant Operations (IP 71707, 71710, 93702)
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2.1 Operational Safety Verification...........
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2.2-Engineered Safety Features System Walkdown
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2.3 Followup of Events Occurring _During the
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Radiological Controls (71707)...............
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Maintenancefand Surveillance (61726, 62703, 71707)....
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4.1 Maintenance Observation...............
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I-4.2 Surveillance Observation 5-
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Emergency Preparedness (71707)..............
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Security (71707).....................
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Engineering.and Technical Support (37700, 37828, 71707)..
7.1 Twview'of Emergency Diesel Generator Reliability
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. Concerns
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7.3' 1988 Unit 1 Overcooling Event.
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7.4 Unit 1 Safeguards Building Ventilation........
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-Safety: Assessment.and Quality Verification (40500, 71707, p
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90712, 92700)................'........
8.11 Review of Written Reports..............
8.2, Quality Concerns Resolution Program.........
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Followup of Previous Inspection Find'ngs (IP 71707, 92702, 92701).................
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Exit Meetings (30703)..................
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DETAILS 1.
Summary of Facility Activities At the beginning of the period, Unit I was operating at 100% power and Unit 2 at 90% power as part of a core life extension schedule. Unit 1 operated at full power until March 30, when the reactor tripped due to a main feedwater regulating valve partially stroking shut (see Detail 2.3.2).
At the end of the period, Unit I was in Hot Standby (Mode 3). Unit 2 operated at approximately 85% power for the majority of the period, reducing power to approximately 47% on weekends as a fuel saving measure. On March 30, Unit 2 power was raised to 98% following the Unit 1 trip and remained at that level at the end of the period.
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 procedures and regulatory requirements.
Regular tours were conducted on the following plant areas:
-- Control Room
-- Safeguard Areas
-- Auxiliary Buildings
-- Service Buildings
-- Switchgear Areas _
-- Diesel Generator Buildings
-- Access Control Points
-- Intake S'tr:ture
-- Protected Area Fence Line -- Yard Are<>
-- Spent Fuel Building
-- Containmt:4 Penetration
-- Turbine Buildings Areas During the course of the inspection, discussions were conducted with operators concerning knowledge of recent changes to procedures,
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facility configuration and plant conditions.
The inspector verified adherence to approved procedures for ongoing activities observed.
Shift turnovers were witnessed and staffing requirerents confirmed.
The inspectors found that control room access was properly controlled and a professional atmosphere was maintained. Inspector comments or questions resulting from these reviews were resolved by licensee personnel.
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 onsite and offsite power
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sources were verified. The inspectors observed various alarm conditions and confirmed that operator response was in accordance with plant operating procedures. Compliance with TS and implementation of appropriate action statements for equipment out of service was inspected.
Logs and records were rev1ewed to determine if entries were accurate and identified equipment status or deficiencies. These records included operating logs, turnover sheets, system safety tags, and the jumper and lifted lead book.
The inspector also examined the condition of various fire protection, meteorological, and seismic monitoring systems.
Plant housekeeping controls were monitored, including control and a
storage of flammable material and other potential safety hazards.
The inspector conducted detailed walkdowns of accessible areas, including normally locked high radiation areas, of both Unit 1 and Unit 2.
Significant improvements in Unit I housekeeping were observed. 'The amount of loose dirt and debris in the Unit 1 Safe-guards and Auxiliary Buildings was greatly reduced from the last inspection period.
Ladders and equipment carts were found to be adequately restrained. General housekeeping in Unit 2 was good.
2.2 Engineered Safety Features System Walkdown The operability of selected engineered safety feature systems was verified by performing detailed walkdowns of the accessible portions of the systems.
The inspectors confirmed that system components were in the' required alignments, instrumentation was valved in with ap-propriate calibration dates, as-built prints reflected the as-installed systems and the overall conditions observed were satisfactory. The systems inspected during this period include the Emergency Diesel Generator, Safety Injection Auxiliary Feed and Recirculation Spray systems.
No concerns were identified.
2.3 Followup of Events Occurring During the Inspection Period During the inspection period, the inspectors provided onsite coverage and followup of. unplanned events.
Plant parameters, performance of safety systems, and licensee actions were reviewed.
The inspectors confirmed that the required notifications were made to NRC.
The following events were reviewed:
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- 2. 3.'1 Unit 1 Operation Above Ful' Power On-February 27, 1990, operators performing a daily thermal heat balance surveillance, determined that the Unit I reactor was operating at 100.4% power.
The operators
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immediately lowered power to below 100%. Chemistry analysis revealed no chan0e in coolant activity as a result
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of operation above 100% power.
The cause of the event was determined to be an aborted program in the unit process computer (P-250) which operators utilized to provide three inputs to the daily thermal heat balance calculation. The aborted program provided 10-minute averages for steam generator feed flow, pressure and steam flow. When the above program aborted on February 18, 1990, it retained the average values last calculated.
The operators, who were unaware that the
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program had aborted, continued to utilize the average values retained in the program memory.
On february 19 and 24, the power range nuclear instruments were adjusted downward (0.3% and 0.5% respectively) to match indicated with calculated reactor power.
Following each adjustment, reactor power was subsequently raised to 100% indicated power. This in turn resulted in actual
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condition was not identified until the P-250 computer was reactor power being raised slightly above 100%.
This rebooted on February 27 and the 10-minute average values calculated by the computer reflected actual plant values.
Review of data from an independent computer system indicated that actual reactor power never exceeded 100%
-except for approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> between February 26 ani February 27.
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Contributing to this event was the lack of error messages or other computer prompt informing the operator that the averaging program had aborted. As a corrective measure, the licensee changed the daily thermal heat balance surveillance procedure to require operators to verify the computer calculated averages vary in conjunction
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with control room indicators before using the values in the heat balance calculation.
In addition, a requirement was added to verify heat balance calculations using independent parameters prior to adjusting the power range nuclear instrument indication downward.
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.The inspector found that the licensee's corrective actions were adequate and had no further questions concerning this event.
2.3.2 Unit 1 Reactor Trip
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On March 30, 1990, Unit I reactor tripped from 100% power due to the IC Main Feedwater Regulating Valve (MFRV)
stroking partially shut. The control room operators took
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manual control of the valve and attempted to reopen it.
However, the valve failed to respond. The reactor subsequently tripped on steam flow / feed flow mismatch coincident with low water level in the IC steam generator.
All systems responded as designed. The control room operators quickly stabilized the plant in accordance with procedures.
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Subsequent investigation found a significant amount of water in the electro pneumatic (1/P) converters for the IC MFRV.
Water was also found in the I/P converters for the other MFRVs.
For approximately four days prior to the event, the Unit i instrument air dryers had apparently been inoperable and had only intermittent service since the recovery from the last refueling outage (about 94 days prior to the trip). Aside from the MFRVs, the licensee found no appreciable amounts of. water in the instrument air system.
The valve house where the MFRVs were located was subject to variations in >mperature as outside temperature varied. The :.censee believed that the variations in room temperature in conjunction with the inoperable instrument air dryer resulted in the condensation and build up of water in the.MFRVs.
At the end of the inspection period, the licensee was still evaluating the event and developing corrective actions.
Some corrective actions in progress included the disassembly.and inspection of all three MFRVs and the repair of the air dryer.
The inspectors independent review and evaluation of the event and the licensee's corrective actions will be discussed in future inspection reports.
3.
Radiological Controls Posting and control of radiation and high radiation areas were inspected.
Radiation Work Permit compliance and use of personnel monitoring devices were checked.
Conditions of step-off pads,
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disposal of protective clothing, radiation control job coverage, area monitor operability and calibration (portable and permanent)
and personnel frisking were observed on a sampling basis.
No deficiencies were identified.
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Maintenance and Surveillance 4.1~. Maintenance Observation
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The-inspector reviewed selected maintenance activities to
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assure that:
the activity did not violate Technical Specification
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Limiting Conditions for Operation and that redundant-i components were operable; required approvals and releases had been obtained. prior to -
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commencing work;
procedures used for the task were adequate and work was within
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the skills of the trade; activities were accomplished by qualified personnel;
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where necessary, radiological and fire preventive
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controls were adequate and implemented; QC hold points were established where required and observed;
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equipment was properly tested and returned to service.
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Maintenance activities reviewed included:
MWR 896461 Replacement of Unit 2 Auxiliary Feedwater Pump Turbine Emergency Overspeed Trip Tappet Assembly.
MSP 36.70-02
' Replacement of No. 2 Emergency Diesel Lube Oil Soak.Back Pump Filters.
DCP 1185 Installation of Inline Air Filters for Diesel Air Start System.
No deficiencies were identified, 4.2 Surveillance Observation The inspectors witnessed / reviewed selected surveillance tests to determine whether properly approved procedures were in use, details were adequate, test instrumentation was properly calibrated and used,
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Technical Specifications were satisfied, testing was performed by qualified personnel and test results satisfied acceptance criteria or were properly dispositioned.
The following surveillance testing activities were reviewed:
OST 1.30.1B (2WR-P-98) Auxiliary River Water Pump Test.
OST 2.36.1 Emergency Diesel Generator (2EGS*EG2-1)
Monthly Test.
No deficiencies were identified.
5.
Emergency Preparedness The inspector toured the licensee's new Joint Public Information Center (JPIC) located outside the 10-mile evacuation zone near the Pittsburgh International Airport. The alternate Emergency Response Facility was also located in the same facility. 'The inspector found that the new JPIC was l
well planned and equipped, providing excellent facilities for the various l
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public information services.
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6.
Security Implementation of the Physical Security Plan was observed in various plant
areas with regard to the following:
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Protected Area and Vital Area barriers were well maintained and
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not compromised;
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Isolation zones were clear;
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Personnel and vehicles entering and packages being delivered to i
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the Protected Area were properly searched and access control was in I
accordance with approved licensee procedures; l
i Persons granted access to the site were badged to indicate
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whether they have unescorted access or escorted authorization; Security access controls to Vital Areas were being
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properly authorized;
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Security posts were adequately staffed and equipped,
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security personnel were alert and knowledgeable regarding position requirements, and that written procedures were available; and Adequate illumination was maintained.
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No deficiencies were identified.
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Engineering and Technical Support
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7.1 Review of Emergency Diesel Generator Reliability Concerns Several concerns regarding Unit 1 Emergency Diesel Generator (EDG)
reliability were identified in previous inspections-(ses Section 9.1).
During this period, the inspector reviewed the licensee actions to improve the reliability of the EDG air start system Solenoid Operated Valves (SOVs) and licensee actions to evaluate possible failure mechanisms associated with a-failed EDG Manual Start Relay (MSR).
The licensee had committed to incorporate the EDG vendor recommenda-tion of replacing each EDG air start SOV every three years.
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previous inspection, the inspector found that there was no formal mechanism in place to ensure that the above valves would be replaced.
The licensee has since formally included the requirements into the Preventive Maintenance program.
The inspector verified the
requirements had been added to the program and noted that half of the
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SOVs were to be replaced.every refueling outage.
Half-of the SOVs-l were replaced as required during the past refueling outage.
The inspector had no further questions, j
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relay to an offsite laboratory who found a large chloride and sulfide I
content on the contact surfaces.
There was a concern that the Fast
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Start Relays (FSR) (which provide safety functions), might also be
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subject to the buildup of chlorides and sulfides and therefore be subject to the same failure as the MSRs.
Further testing by independent laboratories indicated that the chloride / sulfide buildup on the MSR contacts did not inhibit circuit flow and would chip away
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upon contact closure.
Since the initial failure of the. MSR relay,
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there has been no reported failures of other MSR or FSR relays.
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inspector had no further questions.
. 7.2 Control of Measuring and Test Equipment
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The inspector found that the licensee had made substantial improvements in the control of Maintenance and Test Equipment (M&TE)
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including the consolidation of M&TE issue points, and an improved accountability and record keeping system (see~Section 9.2).
The inspector selected approximately 25 instruments from the M&TE calibration records. The location of each instrument was checked i
against the location described in the Daily Issue and Record Log.
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deficiencies were identified. All the selected instruments were d
correctly labelled and none were past due for calibration.
The licensee had consolidated the five M&TE issue areas to two locations (a main issue point and potentially contaminated' gage issue point). The inspector toured these locations and found that M&TE was
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properly marked and segregated as required. The licensee had also made provisions to lock up the main M&TE issue point when unattended.
To improve accountability and record keeping, the licensee. implemented a computerized tracking system._ The inspector found that the system was more than adequate. The system replaced the hand written Daily
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Use and Record Log and also had the capability of.providing reports of-M&TE which had not.been returned to issue point. Another improvement was the addition of a requirement on the Maintenance Work Request. form-that-requires the job foreman to ensure that all M&TE was returned prior to signing the form at the end of the maintenante activity.
Discussions with M&TE users indicated that there was an increased
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sensitivity to the importance for the proper control of M&TE.
No unattended and not in use M&TE was observed during tours of both Units 1 and 2,
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7.3 1988 Unit 1 Overcooling Event A Unit 1 reactor trip occurred from full power following the inadvertent trip of a reactor coolant pump (RCP). About 29 seconds after the trip an automatic safety injection was initiated on low pressurizer pressure due to rapid cooldown of the reactor coolant system (RCS).
The inspector reviewed the licensee's engineering evaluation.of this event (Section 9. 3).
The two main contributors to the RCS cooldown were the over-response of the steam dumps system and greater than previous injection of old feedwater.
.The steam dump over response was due to the way RCS temperature was sensed.
Specifically, the trip of the RCP. induced stagnant or backflow conditions in the RCS manifolds used for temperature sensing. These manifolds were rembved'in' the last outage and RCS temperature is now sensed in detectors mounted directly in the large diameter RCS piping.
This contributor to the overcocling event has, therefore, been eliminated.
Licensee actions to improve feedwater system reliability included addressing the high differential pressures across the feedwater regulating valves experienced during some operating conditions.
These actions included modifications to the main feedwater pumps impellers and the internals of the feedwater regulating valves.
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side effect of these modifications was to cause an increase in feedwater flow immediately following a. reactor trip above that which had previously been experienced.
Licensee analysis of other similar events before and after the feedwater modifications (but without the
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steam dump over-response) showed the contribution of the feedwater system response to be essentially negligible. The inspector reviewed the data associated with the various transients and the licensee's analysis and had no further questions.
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7.4 Unit 1 Safeguards Building Ventilation The inspector noted previously that loss of offsite power scenarios had the potential to result in elevated temperatures in two subcompart-s ments of the Unit 1-Safeguards Building.
In_ response to the inspector's concerns, the licensee conducted a review of the. adequacy of the HVAC for the subcompartments in question.
The inspector identified certain weaknesses in the calculations (see Inspection Report 50-334/89-18) a'nd-the licensee performed additional reanalysis. The results of the analyses led the licensee to change the subcompartments HVAC setpoints, increase related HVAC maintenance priorities, and confirm certain stoam-driven pump heat release assumptions.
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During the current period, the inspector verified that the HVAC setpoints had been changed and the maintenance priority to the-supporting coolers had been upgraded. The inspector also reviewed the i
test data from the room temperature effects of the operation of the steam driven pump. The data indicated that the heat calculations associated with pump operation were reasonable. This item is closed.
8.
Safety Assessment and Quality Verification 8.1 Review of Written Reports
I The inspector reviewed LERs and other written reports submitted to p
the' NRC Region I Office to verify that the details of the events were i
clearly reported, including accuracy of the description of cause and adequacy of corrective action. The inspector determined whether further information was required from the licensee, whether generic implications were indicated and whether the event warranted _onsite followup.
The following LERs were reviewed:
Unit 1:
LER 90-002-00 Plant Shutdown Due to Misinterpretation of Vital Bus Operability Requirements.
LER 90-004-00 Failure to Record Rod Positions.
LER 90-005-00 Casting Defect in Limit Stop Housing.
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LER 90-006-00 Plant Operation in Excess of Licensing Basis.
Unit 2:
i LER 90-002-00 Failure to Perform Time Response Testing.
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The above~LERs were reviewed with respect to the requirements of 10 CF.R 50.73 and the guidance provided-in NUREG 1022. Generally, the LERs were found to be of high quality with good documentation of event analyses, root cause determinations and corrective actions.
Particularly notable were licensee actions with respect to Unit 1 LER 90-005 which was reported under the provisions of 10 CFR 21.
Use of BARTS (Butterfly Analysis and Review Test System) to troubleshoot a-chronic problem with certain safety related valves identified a possible generic root cause.
The root cause was a casting defect involving excessive material in an intended (design) recess such that
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the surface _had been machined flat by the manufacturer.
The defect
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was confirmed to be present in the other five applicable valves.
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identical parts were found to be used in certain other valves but no similar performance problems were identified.
The inspector found the licensee's actions associated with this item to be.a good example of problem investigation, root-cause analysis and corrective action.
The following Unit 1 incident report was reviewed:
Incident Report 1-90-029 Six Cases of Termination Radiation Exposure Reports Overdue.
The inspector found the licensee's investigation, root-cause analysis and corrective action to be detailed and thorough.
8.2 Quality Concerns Resolution Program A review of the licensee's Quality Concern Resolution Program (QCRP)
was conducted to assess if the program is adequately addressing employee concerns and whether corrective actions, when-required, are adequately implemented. The charter of the QCRP is to address all concerns and allegations that licensee employees (both permanent and contract) may have concerning nuclear safety or quality issues.
The QCRP is administered in accordance with Quality Assurance
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Instruction 15.13. The program provides several different methods-in which an employee can communicate concerns including the mail, telephone or walk in.
In addition, all Quality Control (QC)
Inspectors are required to'be given exit interviews before leaving the site. When a concern is received, the Manager of Quality Assurance assigns an investigator to resolve the concern.
The inspector selected a sample of 18 investigations for further review.
The inspector-found that the QCRP is generally effective and is being utilized by employees with concerns.
Several concerns have been substantiated. The inspector found that the lack of timeliness and the limited scope of some of the investigations has a negative impact on investigation quality. These weaknesses in timeliness and
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- L scope of certain individual investigations were discussed with cognizant licensee personnel. The apparent acceptance and use of-the QCRP by licensee personnel as an effective method to raise concerns is considered to be a strength.
9.
Followup of Previous Inspection Findings The NRC Outstanding Items List was reviewed with cognizant licensee personnel.
Items selected by the inspector were subsequently reviewed through discussions with licensee personnel, documentation reviews and
- e field inspection to determine whether licensee actions specified in_the Ols had been satisfactorily completed.
The overall status of previously
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identified inspection findings was reviewed, and planned / completed
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licensee actions were discussed for the items reported below, t
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9.1 (Closed) Unresolved Item.(50-334/86-07-01):
Improve Emergency
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Diesel Generator (fDUT reliability.
This item consolidated four previously identified NRC concerns into one item.
Subitem I was t
closed in Inspection Report 50-334/89-22; 50-412/89/21.
Subitem 3
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was combined with-other identified concerns with EDG air start system and is being tracked as part of NRC Unresolved Item No. 50-334/
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88-22-02. - Subitem 2 concerned the reliability of the EDG air start system solenoid operated valves (SOV). Subitem 4 concerned the lack.
of documented evaluation of the cause of the failure of an EDG Manual Start Relay (MSR). -The review of Subitems 2 and 4 is documented in Section 7.1.
9,2 (Closed) Violation (50-334/89-13-01; 50-412/89-14-01):
Inadequate control of Measuring and Test Equipment (M&TE).
Significant problems had been identified concerning the control of M&TE. Of particular concern was the issuance of beyond calibration
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M&TE for use on safety related equipment. There were also numerous examples of the fsilure to properly mark and segregate past-due M&TE, to return M&TE to issue points at the completion of the recorded work activity, to record all work performed using a piece of M&TE in the
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Daily Issue and Use Record Log, and to use the Daily Issue and Record Log when checking out M&TE.
In addition, a weak practice was identified concerning the lack of requirements to lock up the M&TE issue points when unattended.
The inspector reviewed the corrective actions of the licensee to prevent recurrence, The review of this item is documented in Section 7.2.
9.3 (Closed) Unresolved Item (50-334/88-02-01): Overcooling concerns following reactor trip. A Unit 1 reactor trip occurred from full power following the inadvertent trip of a reactor coolant pump. The inspector reviewed the licensee's evaluation of the event.
The review of this item is documented in Section 7.3.
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k 9.4~ (Closed) Unresolved Item (50-334/89-04-02): ' Unit 1 Safeguards
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Building Ventilation-Cohcerns. The_ inspector had previously noted'
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that11oss of offsite power scenarios had the potential to result in D'
' elevated temperatures'in two subcompartments of the Safeguards Building.
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The inspector--reviewed the licensee's evaluation of this concern.-
p LThis review is documented-in Section 7.4.
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10.. Exit Meeting g
LPeriodic meetings were held with senior facility management during'the;
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course of this. inspection:to. discuss the inspection scope and findings. Az summary of inspection findings;was further discussed with the licensee at the conclusion-of; the. report period on April 6, '1990.
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