IR 05000029/1990007
| ML20043G310 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 06/12/1990 |
| From: | Gray E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20043G309 | List: |
| References | |
| 50-029-90-07, 50-29-90-7, NUDOCS 9006200143 | |
| Download: ML20043G310 (16) | |
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V.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No:
50-29/90-07 Docket No:
50-29-Licensee No:
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Licensee:
Yankee Atomic Electric Company 580 Main Street Bolton, Massachusetts 01740-1398
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Facility Name: Yankee Nuclear Power Station Inspection at: Rowe, Massachusetts Inspection Conducted:
April 3 - May 14, 1990 Inspectors:
T. Koshy, Senior Resident Inspector
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T M. Markley, Resident Inspector
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Approved By:
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l E. H. Gray, Acting JNiief, Reactor Projects Section 3A Date Inspection Summary: Inspection on April 3 - May 14, 1990 (Report No. 50-29/90-07)
Areas Inspected:
Routine inspection on daytime and backshifts by two resident inspectors of: operational safety; plant operations; radiological controls;
maintenance and surveillance; security; engineering and technical support;
safety assessment / quality verification; licensee event reports; and periodic l
reports, j
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1.
General Conclusions on Adequacy, Strength or Weakness in Licensing Programs The plant management was very responsive to concerns regarding unrestrained
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. transient equipment and the removal of shipping caps on installed Rosemount
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transmitters. The receipt and storage of new fuel was performed in a safe
manner.
2.
Unresolved Items-
. Two items remained unresolved at the end of the inspection. The first
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involved the acceptability of component cooling water surge tank alarm setpoint and surveillance (section 8.1).
The other concerned a report-ability evaluation on the failure of commercial grade equipment in the safety related applications (section 8.2).
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Violation
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One non-cited violation was identified involving the unauthorized entry of an individual into a controlled access area (section 7.2).
.9006200143 900612 "
PDR ADOCK 05000029
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i EXECUTIVE SUMMARY
FlantOperations(Modules 71707, 81403, 60705)
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Yankee Nuclear Power-Station (YNPS) started coasting down for refueling on April 24, 1990. The pc.wer level is decreasing at the rate of $ percent rated power per week.
Licensee control over revision of auxiliary operator logs was
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inadequate in that it led to the omission of a motor operated valve status
verification for 2 days. (Section 4.3)
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L Radiological Controls (Module 71707)
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Radiological controls for the new fuel receipt, handling and storage was ade-
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quate.
Maintenance and Surveillance (Modules 61726,62703,61700,62700)
The surveillance activities performed by Electrical and Instrumentation & Con-trols (1&C) personnel were effective.
The unrestrained test gear in the switch j
gear area.was promptly corrected in response to the inspectors' concern, The
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technicians exhibited thorough knowledge of test equipment and the working pro-
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cedures.
- Secur'ity (Module /2707)
Security personnel identified an unauthorized entry into a controlled area.
This was.due to a malfunctioning access control device and a recent down grade l
of certain access authorizations. (Section 7.2)
q Engineering and Technical Support (Modules 71707,62703,37700)
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The inspector-identified the potential for significant chromateo water leakage
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from the component cooling water (CCW) system into the service water (SW) system-
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prior to main control board alarm of the CCW Surge tank level instrumentatioi.
system. This-is an unresolved item pending completion of the licensee technical
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evaluation. (Section 8.1)
- During a Plant Operation Review Committee (PORC) meeting discussion of an in-
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- operable pressure indicator, the inspector observed a potential problem with
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licensee awareness of 10 CFR 21 reportability requirements for dedicated com-
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ponents.
This item is unresolved pending further licensee evaluation. (Section 8.2)'
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Safety Assessment / Quality Verification (Modules 71707,40500)
'The NRR Project Manager riviewed the licensee quality asturance program for
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emergency diesel generator fuel oil relative to multi plant action item MPA
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A-15 (TI 2515/93) and moderator dilution requirements relative to multi plant action item MPA B-03 (TI.2515/94).
No unacceptable conditions were identified.
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TABLE OF CONTENTS
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Persons Contacted....................................................
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S umma ry of Fa c i l i ty Ac t i v i t i e s.......................................
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3.
' Operational Safety Verification (IP 71707)............................
3.1 Plant Operations Rev1ew.........................................
l 3.2 Safety System Review.............-...............................
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3.3 Review of Temporary Changes, Switching and Tagging..............
3.4 - Opera ti onal Sa f e ty Fi ndi ng s......................................
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3.5 Storage of Transient Equipment..................................
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Plant Operations (IP 71707,81403,60705)............................
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4.1 Plant Coastdown_to Refueling....................................
4.2 New Fuel Receiot.................................................
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4.3 Mi ssed Auxilia ry Operator Survei11ances.........................
4.4 Licensed Operator Requalification Exams.........................
4. 5. 0p e ra to r T ra i n i n g................ ;..............................
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5.
Radiological Controls (IP 71707).....................................
5.1 Radiological Controls for New Fuel Receipt......................
5.2 Primary Drain Collection Tank 0verflow..........................
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Maintenance / Surveillance (IP 61726,62703,61700,62700)..........,..-
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6.1 Rosemount Transmitters..........................................
7-6.2 Emergency Feed Water System Operability Test....................
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Security (IP 71707)...................
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L7.1 Observations of Physical Securi ty...............................
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7.2 Unauthorized Entry into Controlled Access Area..................
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Engineering / Technical Support (IP 71707,62703,37700)...........-....
8.1'. Component Cooling Water System.....................
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8.2 Nonconformance Report (NCR)......................................
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Table of Contents
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Safety" Assessment / Quality Verification (IP 71707,40500).............
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9.1' Verification.of Quality Assurance Request Regarding Multi-Plant.
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4; Action Item A-15 (TI 2515/93).................................
L 9.2 Verification of Licensee Changes Made to Comply with PNR El Moderator Dilution Requirements Multi-Plant Action Item B-03 i
(TI-2515/92)..................................................
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LicenseeEventReporting(LER)(IP90712)............................- 11
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i 10.1 L E R 9 0 0 01....................................................... ' 12.
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Review of Pe riocic Reports (I P 90713).................................
0 12. Management Meetings (IP 30703).......................................
- The NRC Inspection Manua1' inspection procedure (IP) or temporary instruction-(TI)1or'theLRegion I temporary instruction (RI TI) that was used as inspection o
L guidance is listed for each applicable report section, p
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DETAILS 1.
Persons Contacted Yankee Nuclear Power Stat,ign T. Henderson, Plant S Arin ' dent R. Mellor, Technical Drett Yankee Atomic Electrit pany (YAEC)
N. St. Laurent, Manager of Operatirens The inspector also interviewed other licensee employees during the inspec-tion, including members of the operations, radiation protection, chemistry, instrument and control, maintenance, reactor engineering, security, train-ing, technical services and general office staffs.
2.
Summary of Facility Activities Yankee Nuclear Power Station (YNPS, Yankee or the plant) has maintained continuous power operation since August 30, 1989. The plant began coast-down to refueling at 10:00 p.m. on April 24. A seven week refueling and maintenance outage is scheduled to begin June 23, 1990.
Effective May 13, 1990, Mr. Edwin H. Gray, Senior Reactor Engineer in the Materials and Processes Section of the NRC Region I Division of Reactor Safety (DRS) assumed cognizant Division of Reactor Projects (DRP) Section Chief responsibilitier, for YNPS.
On April 16 - 20, 1990, two NRC Reaion I specialist inspectors conducted a special safeguards inspection (NRC Inspection 50-29/90-08).
3.
Operational Safety Verification 3.1 Plant Operations Review The inspector observed plant operations during regular and bacK nift tours of the following areas:
Control Room Safe Shutdown System Building Primary Auxiliary Building Fence Line (Protected Area)
Diesel Generator Rooms Intake Structure Vital Switchgear Room Turbine Building Cable Tray House Spent Fuel Pit (SFP) Building Safety Injection Building The following items were checked during daily routine facility tours:
shift staffing, access control, adherence to procedures and limiting conditions of operation (LCOs), instrumentation, recorder traces,
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protective systems, control room annunciators, area radiation a v process monitors, emergency power source operability, operabili:/ of the Safety Parameter Display System (SPDS), control room logs, shift supervisor logs, and operating orders. On a weekly basis, selected Engineered Safety Feature (ESF) trains were verified to be operable.
The condition of plant equipment, radiological controls, security and safety were assessed.
On a biweekly frequency, the inspector re-viewed safety-related tagouts, chemistry sample results, shift turn-overs, portions of the containment isolation valve lineup and the posting of notices to workers.
Plant housekeeping and fire protec-tion were also evaluated.
Inspections of the control room were performed on weekends and back-shifts as follows: April 3, 4, 9, 10, 11, 12, 13, 14, 23, 24, 25, 25 and May 12.
Deep backshift inspection was performed on April 14 from 6:00 a.m. to 8:15 p.m. and May 12 from 9:30 a.m. to 10:15 p.m.
Operators and shift supervisors were alert, attentive and responded appropriately to annunciators and plant conditions.
No unacceptable conditions were identified.
Proper control room decorum was routinely observed.
3.2 Safety System Review The emergency diesel generators, EDG fuel oil, residual heat removal, and safety injection systems were reviewed to verify proper alignment and operational status.
The review included verification that (i) accessible major flow path valves were correctly positioned, (ii) power supplies were energized, (iii) lubrication and component cooling was proper, and (iv) components were operable based on a visual inspection of equipment for leakage and general conditions.
System walkdowns to assess the material condition of the ECCS HPSI and LPSI and the low pressure safety injection accumulator were per-formed.
Selected accessible valves were verified to be in the cor-rect position and locked when required by plant procedures.
3.3 Review of Temporary Changes Switching and Tagging Temporary change requests (TCRs), which were approved in support of implementing lifted leads and jumper requests and mechanical bypasses, were reviewed to verify that: controls established by AP-0018, " Tem-porary Change Control," were met; no conflicts with the Technical Specifications were created; the requests were properly approved prior to installation; and a safety evaluation in accordance with 10 CFR 50.59 was prepared if required.
Implementation of the requests was reviewed on a sampling basis.
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The switching and tagging log was reviewed and tagging' activities were inspected to verify plant equipment was controlled in accordance i
with the requirements of AP 0017,
" Switching and Tagging of Plant
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Equipment."
Licensee administrative control of off-normal system configurations
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by the use of TCR and switching and tagging procedures as reviewed
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above, was in compliance with procedural instructions and was con-
$1 stent with plant safety.
No unacceptable conditions were identi-fled.
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3.4 Operational Safety Findings An operatienti event occurred on March 20, 1990, involving an over-flow of the Prfmary Orain Collections Tank (PDCT).
It is discussed s
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in detail in section 5.2 of this report.
This was primarily due to
L an instrument malfunction. The NRC inspector reviewed that Plant Incident Report (PIR) No. 90-01 on the subject event.
This document indicated one of the root causes was the lack of auxiliary operators
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(AO) response to the increasing level trend in the PDCT level.
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inspector discussed this concern with the operational support staff.
The primary A0 log, performed every two hours indicated a level of
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72 percent in the tank.
The reviews after the event confirmed this i
to be an inaccurate indication.
The acceptance value for the-level was 30 to 90 percent.
a The inspector was concerned that the 90 percent acceptance value does not afford reasonable time to respond to the increasing level in PDCT.
The licensee is examining this matter to determine a suitable value when the new level sensor;and control system is installed.
Licensee management is reviewing other acceptance values in A0 log for an early detection of inoperative level-control systems.
These corrective actions were deemed acceptable.
3.5 Storage of Transient Equipment
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During a routine plant tour, the inspectors observed 3 pieces of l
switchgear equipment on rollers left unattended in the switchgear-room._ These-items are test gear for the~different types of switch-gear.
The. pieces of equipment in question were not restrained and were free to move in the case of seismic event, The NRC inspector
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brought this to the licensee's attention.
The present plant proce-
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dures do not have explicit guidelines for storing heavy equipment in f
the plant area. The licensee took prompt measures to restrain the above referenced equipment.
In order to prevent the recurrence, licensee procedure AP-0040 is being revised to provide clear guide-
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lines on storage of heavy equipment.
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4.
Plant Operations 4,1 Plant Coastdown to Refueling At 10:00 p.m. on April 24, the plant began coastdown to refueling.
The core XXI refueling and maintenance outage is scheduled to begin June 23,1990. The planned outage duration is seven weeks.
4.2 New Fuel Receipt One of the fuel shipment casks contair.ing fuel bundles B820 and BB22 had a tripped internal acclerometer. However, the internally mounted accelerometer was not tripped. The licensee promptly informed the fuel manufacture representative on site.
The representative followed the Combustion Engineering procedure 0C-01-02 for inspecting the fuel for physical integrity and concluded that the fuel was acceptable.
The NRC inspectors were concerned about the potential disturbance to the fuel pellet arrangement inside the fuel rods.
The licensee in pursuing this matter for a satisfactory response.
4.3 Missed Auxillary Operator (AO) Surveillances During routine surveillance at 8:00 a.m. on April 14, 1990, the primary plant side A0 noted that the required position verification for safety injection accumulator valve $1-MOV-1 was missing from the A0 logsheet form.
The individual brought this to the attention of control room staff who determined that it had been inadvertently de-leted in the revised-logsheet (Rev. 41).
Technical Specification 4.5,1(a)(2) requires the safety injection accumulator be demonstrated operable at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that the accumulator isolation valve SI-MOV-1 is open, The licensee initiated event re-portability evaluation (ERER) No, 90-19 which determined it was not reportable.
On April 12, 1990, Revision 41 of the A0 log sheet went into effect.
This revision rearranged the sequence of the log entries to agree with the order in which this surveillance is performed.
This revi-sion omitted the entry for verifying the status of the subject MOV.
Licensee operations management interviewed the 3 A0's that conducted surveillance between April 12, 1990 and April 14, 1990, The A0s con-firmed the performance of the subject surveillance but did not ques-tion the discrepancy of the log sheet, The licensee review of this event also noted that the MOV status light in the control room pro-vided additional verification of SI-MOV-1 position.
During the shift turnover, through the performance AP-2002, the off normal equipment /
system status is verified. The licensee concluded that the operabil-ity of the safety injection accumulater was not compromised and the only omission was a properly documented log.
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Since the operability was not challenged, the licensee considers this event to be not reportable under the LER.
Due to the safety signi-ficance of the event, Plant Incidence Report (PIR) No. 90-01 was pro-cessed for root-cause evaluation and corrective actions.
Inspector review noted the licensee's current logsheet procedure re-vision does not include an independent review requirement. The lic-ensee was responsive to the inspector's concern. Corrective actions to prevent recurrence include a procedural requirement for indepen-dent review of procedure revisions prior to PORC review.
4.4 Licensed Operator Requalification Examinations During this inspection period, the licensee training staff adminis-tered requalification examinations to 9 senior reactor operators (SRO) and 9 reactor operators (RO).
One R0 and 2 SR0s failed in the comprehensive written examination.
The candidates demonstrated weakness in the knowledge of Emergency, Operating Procedures.
These operators were promptly removed from the shift duties.
In accordance with the plant procedure AP-0500,
" Yankee Atomic Electric Company Operator Training Program," the lic-ensee called a meeting of the Training Advisory Committee.
The com-mittee required a retake of the examination after the candidates com-pleted a prescribed review. All candidates passed the retake ex-amination conducted on April. 25, 1990.
The licensee training department stated that a constant effort is being maintained to improve the skills of the operator.
The licensee raised the passing grade to a. score of 83 out of a possible 100 points.
The previous passing grade was 80. The operations staff appropriately removed the operators from shift duties until success-
-ful completion of the retake exams was accomplished.
4.5 Operator Training The NRC inspector, reviewed the operator training class on Reactor Vessel Internals conducted on May 5, 1990.
The training was tailored to emphasize the needs for the upcoming refueling outage.
Technical specification;1imits on heating, cooling and safety limits were dis-cussed in detail.
The vessel internal assembly, disassembly, and fuel movement were also highlighted to facilitate refueling opera-tions.
The training was presented in an effective manner which solic -
ited a high level of class participation and shared knowledge.
No discrepancies were observed.
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5.
Radiological Controls Selected radiological controls and other work evolutions were observed to i
determine the adequacy of program implementation commensurate with the radiological hazards and importance to safety and to determine conformance
to licensee procedures and other regulatory requiremants.
Independent surveys were performed by the inspector to verify the adequacy of radio-i logical controls and instructions to workers.
5.1 Radiological Controls for New Fuel Receipt
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Inspector review of radiological controls for new fuel receipt indi-
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The inspector observed radiation protection (RP) personnel perform adequate surveys to properly assess radiological integrity of the fuel assemblies and shipping containers.
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Inspector review of air sampling and smear counting results indicated i
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Personnel access and fuel handling was tightly controlled. No unacceptable radiological conditions were identified.
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On March 20, 1990 at 1:15 a.m. the Waste Gas Loop Seal Radiation Monitor alarmed. An auxiliary cperator (AO) was dispatched to the site to investigate the problem. Water was observed overflowing the
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chemistry primary sample sink.
The licensee conducted a prompt
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evaluation of the airborne radioactivity release through the primary i
vent stack (PVS) and contained the spread of contamination around the r
sample sink.
The A0 found that the waste gas compressors were not
operating.
These compressors reduce the pressure in the system by t
pumping the gas to decay drums.
i The Primary Drain Collection Tank (PDCT) is equipped with a level
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control system. A high level signal of approximately 70 percent actuates the pump that transfers the contents to Waste Holdup Tank L
and/or Activity Decay and Dilution Tank.
During the event, this level control system failed to operate due to a stuck level switch.
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This led to the overflow of the PDCT through the sampling line into the chemistry sample sink.
The fluid also filled up the suction line
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of the waste gas compressors. _The fluid filled line prevented the pressure switch from sensing the increasing gas pressure in the waste
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Gas pressure continued to rise in the system and led to the release through the PVS via the loop seal.
The total gaseous offsite release 9.9 E+3 microcuries was well within regulatory limits.
The effluent release rate was 1.1E-10 micro-curies /mi of equivalent Xe-133 as the most restrictive isotope..The
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10 CFR 20.106 limit for radioattive Xe-133 effluent released to un-restricted areas is 3E-7 microcuries/ml.
The 10 CFR 50.72(b)(?)(iv)(A)
limit for four_ hour reportability is 6E-7 microcuries/ml of Xe-133.
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The release was less than 1*4 of regulatory limits, such.that the
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I radiological safety significance of the release is minor. The licen-
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see is in the process of resolving the instrument failure. The exist-ing level instrument cannot be calibrated as it would require degas-sing the PDCT and the associated piping to perform this activity.
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The immediate corrective action has recovered the functions of the existing level control systems. The licensee is evaluating an up-to-i date level sensing system that is testable during normal operation
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and can be maintained in calibration. This modification is currently scheduled for the upcoming outage.
The inspector concluded that
while the immediate corrective action was effective, the licensee
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plan to upgrade the level sensing system is appropriate,
6.
Maintenance / Surveillance
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The inspector observed and reviewed maintenance and surveillance problem
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investigation activities to verify compliance with regulations, adminis-
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trative and maintenance and surveillance procedures as well as the follow-
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ing: codes and standards, proper QA/QC involvement, safety tag use, equip-L ment alignment, jumper use, personnel qualification, radiological controls c
for worker protection, fire protection, retest requirements, LCOs, evalu-l ation of test results, removal and restoration of equipment, deficiency
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review, resolution and reportability per the plant Technical Specifica-
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tions.
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6.1 Rosemount Transmitters
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During a routine tour of the plant on April 12, 1990, the inspectors observed.three Rosemount Transmitters in service with plastic ship-ping caps. The 3 transmi;ters 'were SC-FT-204A, CH-FT-2 and SSS-FIT-10.
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The licensee was informed of this discrepancy.
The Rosemount Transmitters are shipped with 2 conduit entry points for electrical connections.
The factory seals these entry points with plastic caps to preclude moisture and dirt intrusion. 'The manu-
facturer's instruction is to remove the plastic cap at the unused conduit entry point and seal it with a stainless steel pipe plug dur-ing installation.
The. licensee conducted a walkdown of all the plant areas and identi-
fied 19 Transmitters with a plastic cap at the unused conduit entry.
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The manufacture's letter in response to this concern stated that the installation of the steel plug is critical only for environmentally qualified transmitter applications. All these transmitters were located in mild environments and are not subject to high moisture.
The licensee confirmed this problem to be limited to transmitters in the mild environment.
The licensee procedure for environmentally qualified transmitters, OP-6219, Maintenance of Rosemount 1153B Transmitters Revision 4 specifically addresses the installation of the plug.
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i During the installation of these plugs, the licensee observed a potential problem with the installation.
The transmitters located in the mild environment are not equipped with a moisture seal at the primary conduit entry.
If the unused conduit entry is fully sealed, any condensation in the conduit can accumulate at the transmitters.
The recent inspections did not reveal any condensation in the trans-mitters. However, the licensee is evalutting the prudence of sealing unused conduit entry when the primary entry is not equipped with a moisture seal, inspector review noted licensee actions were adequate and appropriate.
6.2 Emergency Feed Water System Operability Test On April 4, 1990 the licensee performed OP-4211, Rev. 28, " Emergency Feed Water System Operability Test." After the completion of the test, the shift supervisor reviewed the results and concluded that the system is operable.
Subsequent review was done by the Inservice Inspection (ISI) coordinator in the review cycle. The ISI coordina-tor observed that the demonstrated flow for P-79-1 motor-driven emer-gency feedwater pump was 51 gpm which is higher than the acceptance value of 42 to 50 gpm. The pump was promptly declared inoperable by the Shift Supervisor even though the results did not reflect inferior pump performance.
The anomaly was discussed in detail at a Plant Operation Review Committee (PORC) meeting. Since the improved flow was due to a change in valve alignment to prevent backflow through the idle pump, PORC concluded that it was acceptable to raise the acceptance value to 51 gpm.
The licensee's action on this finding was thorough and timely.
No unacceptable conditions were identified.
7.
Security 7.1 Observations of Physical Security Selected aspects of plant physical security were reviewed during regular and backshif t hours to verify that controls were in accord-ante with the security plan and approved procedures.
This review included the following security measures: guard staffing, vital and protected area barrier integrity, maintenance of isolation zones, and implementation of access controls including authorization, badging, escorting, and searches.
During a back shift entry by the NRC inspector, the security guard on duty inquired if the conerol room should be informed of the inspec-tor's arrival onsite. Yhe NRC inspector discussed this matter with the security managemert. The licensee agreed to reemphasize the regulatory requiremen; in 10 CFR 50.70 (4) which requires abstaining from announcing NRC presence at the site during inspection visits.
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7.2 Unauthorized Entry into Controlled Access Area In response to concerns during NRC Region I security inspection No.
50-29/89-13, the licensee reevaluated personnel access authorization needs for various plant areas. As a result, some individuals had access authorizations downgraded to limit accessibility of areas where a demonstrated need was not evident.
At 3:00 p.m. on April 7,1990, licensee security personnel identified an unauthorized individual attempting to enter a controlled access area.
Identification of this incident was due, in part, to a mal-functioning access control device.
Security personnel responded and initiated an investigation which determined that the malfunctioning access control device had previously afforded the subject individual unauthorized entry into the controlled access area.
The licensee
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tested other access control devices and determined that no other mal-functioning devices existed.
The licensee subsequently determined that the individual had been authorized for the controlled access area prior to downgrading of his access authorization.
Cognizant licensee security supervision documented the incident as~a security loggable event.
Upon further review on M3y 8, security man-agement determined the incident met one hour reportability require-ments of 10 CFR 73.71(b)(1), Appendix 6 (b) and (c).
Security in-formed the control room at 4:30 p.m. on May 8, and the NRC Operations Center was notified at 5:01 p.m. the same day.
In addition to equipment testing, security management counseled the cognizant supervisor who made the initial reportability evaluation.
Other security supervisors were similarly retrained in those situ-ations requiring NRC notification. The malfunctioning control device was removed from the access control system.
i The licensee was effective in self-identifying the incident and evaluating personnel performance.
Licensee corrective actions and actions to prevent recurrence were adequate and were completed within the scope of this inspection.
Inspector review determined the inci-dent was isolated in nature and not repetitive as set forth in 10 CFR 2,_ Appendix C, Sections V.A and V.G.1 for a non-cited violation of 10 CFR 50.73(b)(1), Appendix G(b)(c) (50-29/90-07-03).
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Engineering / Technical Support 8.1 Component Cooling Watar System The inspectors reviewett selected plant systems against design bases,.
operational safety and environmental concerns.
The component cooling water (CCW) system cools plant equipment which contains radioactive
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fluids. lhis is a closed loop cooling system acting as a barrier to
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reduce the possibility of a leakage of radioactive fluids to the en-vironment.
The component cooling water is in turn cooled by the plent service water.
The water in CCW system is chromated for corrosion control.
The Yankee Rowe design requires a higher pressure for the CCW system than the service water system which cools it.
This arrangement pro-vides the potential for a high pressure to low pressure leakage into the service water system.
The radiation monitor and automatic isola-tion of the vents due to high radiation in the CCW system limits the radiological concerns.
This radiation monitor is critical for de-tecting leakage.
This monitor is currently calibrated once in every 18 months and is not governed by any technical specification surveil-lance. The CCW system surge tank level is indicated in the control room.
The high level alarm for the tank is set for 8 feet and the j
low level alarm is set for 5 feet.
The present setpoints could per-mit a leakage of 21,000 gallons of water into ar from the CCW system prior to level alarm. That is, the CCW could leak into the service water system or the primary system could lea' into the CCW system when the radiation monitor is not in servic9.
The licensee manage-ment plans to review this system to reevalJate the set points at which operations should be alerted of a leak.
This is an unresolved item (90-07-01).
8.2 Non Conformance Reports (NRC)
The NRC inspector attended a Plant Operation Review Committee (PORC)
meeting conducted on April 30, 1990.
The licensee discussion on NCR 90-011 regarding an inoperable pressure indicator, did not demonstrate a thorough knowledge of the 10 CFR Part 21 requirements for licensee dedicated commercial grade components in safety related applications.
In response to the inspectors' concern, the licensee initiated an NCR action item on the reportability evaluation program to formally ad-
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dress the issue of reporting the failure of licensee dedicated com-ponents. This is an unresolved item (90-07-02),
9.
Safety Assessment / Quality Verification 9.1 Verification of Qua_lity Assurance Request Regarding Multi-Plant Action Item A-15 (TI 2515/93)
On March 31, 1988, the NRC issued temporary instruction TI 2515/93 to review and verify requirements for design, construction and operation of structures, systems and components important to safety (including consumables).
A review perfcrmed at a NRC Region IV licensed nuclear power plant, found that diesel generator fuel oil was not on the Q-list for the facility.
This review brought into question whether diesel generator fuel oil was included in the quality assurance (QA) plans for other operating plants.
In January 1980, the NRC requested all licensee's
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i check their QA programs with respect to emergency diesel generator
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(EDG) fuel oil or provide justification for not doing so.
This
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generic activity is identified as MPA A-15.
i The NRC licensing project manager in the Office of Nuclear Reactor Regulation (NRR) reviewed the licensee EDG fuel oil and QA programs relative to the above criteria.
The inspector verified that EDG fuel oil was included in Appendix D,Section IV of the YNPS QA manual.
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Additionally, station procedure OP-9208, " Sampling of Diesel Fuel
Oil," requires that viscosity, water and sediment tests be performed as well as visual examinations of each fuel oil delivery.
The pro-cedure specifies that visual examinations and specific gravity tests be performed prior to offloading.
Inspector review determined that the program met the requirements of MPA A-15.
No unacceptable con-ditions were identified.
9.2 Verification of Licensee Changes Made to Comply with PWR Moderator Dilution Requirements Multi-Plant Action Item B-03 (TI 2515/94)
On March 31, 1988, the NRC issued temporary instruction TI 2515/94 to review changes and verify administration controls, plant modifica-tions and commitments in response to DDR Information Memorandum No.
7, "PWR Moderator Dilution," issued October 4, 1977. This letter
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requested the licensee submit an evaluation of potential moderator
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dilution accidents for the facility including an assessment of fac-turs which affect the capability of the operator to take corrective action.
The letter further requested the licensee inform the NRC if, based on the evaluation results, design or procedural corrective actions are required to preclude the occurrence or consequences of postulated boron dilution accidents. This generic activity is iden-tified as MPA B-03.
The licensee responded in letters dated September 26, 1977, March 26, 1981, and July 8,1981. Also, a safety evaluation, dated July 22, 1981, was submitted to the NRC characterizing protection for postu-lated boron dilution events in all modes assuming the worst single active failure.
Review by the NRC licensing project manager in the Office of NR1 de-termined that the licensee demonstrated adequate protection agaiist postulated boron dilution events in all operational modes for the worst single active failure.
Inspector review determined that the requirements of MPA B-03 were met.
No unacceptable conditions were identified.
10, Licensee Event Reporting (LER)
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The inspector reviewed the below 14ted licensee event reports (LER) to determine whether: (1) the repon was submitted in a timely manner; (2) description of the evan.t. vos accurate; (3) root cause analysis was
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performed; (4) safety implications were considered; and (5) corrective
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actions implemented or planned were sufficient to preclude recurrence of a similar event.
10.1 LER 90-001. " Loss of Power to Emergency Bus No.1 Results in Auto-matic Start of Emergency Diesel Generator No.1," addresses the March
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o 8,1990, incident where the bus tie breaker BT-1B unexpectedly tripped during surveillance testing of the No. I high pressure safety injec-
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tion pump (HPSI). This resulted in a loss of power to the No. 1 Emergency Bus (EBus) and an automatic start of the No.1 emergency r
diesel generator. (EDG). However, the EDG output breaker (EG-1)
failed to automatically close to provide power to the EBus.
The lic-ensee restored power to the EBus by resetting and closing the tie breakers BT-1A and BT-18.
The output breaker EG-1 failure was deter-mined to be caused by binding on the breaker trip shaft which pre-vented the trip shaft from completely resetting.
The tie breaker BT-1B' failure was not repeatable and licensee therefore could not determine the root cause. The licensee repaired EG-1 and inspected the other output breakers to ensure similar conditions did not exist.
No deficiencies were identified in the licensee response'to this in-cident or the reportability.
11. Review of Periodic Reports Upon receipt, the inspector reviewed periodic reports submitted pursuant
.to. Technical Specifications.
This review verified, as applicable:. (1) that
-the reported information was valid and included the NRC-required data;.
(2) that test results and supporting information were consistent with de-sign predictions'and performance specification; and (3) that planned cor-rective actions were adequate for resolution of the problem. The inspec-tor also ascertained whether any reported information should be classified as an: abnormal occurrence.
The following reports were reviewed:
Monthly Statistical Report for plant operations for the months of
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March 1990;.
Annual Radiological Environmental Operating Report, YAEC letter BYR
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90-062, dated April 27, 1990; 10 CFR 20.407, Personnel Monitoring Report, YNSD letter 90-034, dated--
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March 19, 1990.
-No unacceptable conditions were identified.
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Ma.= cement Meetinos
.At periodic intervals during this inspection, meetings were held with
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senior plant management to discuss the findings. A summary of findings
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for the report period was also discussed at the conclusion of the inspec-
. tion and prior to report issuance.
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