IR 05000029/1989006
| ML20247B086 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 07/14/1989 |
| From: | Haverkamp D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20247B057 | List: |
| References | |
| 50-029-89-06, 50-29-89-6, IEIN-89-044, IEIN-89-44, NUDOCS 8907240035 | |
| Download: ML20247B086 (21) | |
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I U.S. NUCLEAR REGULATOR" COMMISSION j
REGION '
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i Report No:
50-29/89-06 Docket No:
50-29
Licensee No:
DPR-3 Licensee:
Yankee Atomic Electric Company 580 Main Street Bolton, Massachusetts 01740-1398 i
Facility Name: Yankee Nuclear Power Station Inspection at: Rowe, Massachusetts Inspection Conducted:
April 10 - May 21, 1989 l
Inspectors:
John Macdonald, Senior Resident Inspector i
M hael T. Markley, Resident Inspector
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Approved By:
% s k k w [ 4 _.- /
M/h Donald R. Haverkamp, Chieff' ~3C
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Dat'e Reactor Projects Section No.
Inspection Summary: Inspection on April 10 - May 21, 1989 Report No. 50-29/89-06 Areas Inspected: Routine inspection on daytime and backshifts by two resident inspectors of:
actions on previous inspection findings; operational safety; security; plant operations; maintenance and surveillance; engineering support;
radiological controls; licensee event reports; licensee response to NRC initi-
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atives; and, periodic reports.
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Results:
1.
General Conclusions on Adequacy, Strengths and Weaknesses in Licensee Program The licensee demonstrated strong operational response to the automatic reactor trip on April 23 and properly declared an Unusual Event. The
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plant was quickly stabilized and an ef fective post trip report was gene-
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ra t ed.,
A noted improvement was observed in the control of activities sup-
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porting reactor restart on April 24.
However, the licensee permitted a l
high radiation exclusion area door to be degraded for an excessive period of time.
Initial assessment of the potential safety impact was not char-acteristic of typical licensee responsiveness.
Senior manage.nent was ultimately effective in addressing this issue.
l 8907240035 890714 PDR ADOCK 05000029
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2.
Violations One violation was 4dentified during this inspection period which involved two examples of licensee failure to follow approved procedures as required by TS 6.8.1.
Both examples of the violation had been identified as un-resolved items in previous inspection reports.
The first example involved the improper circuitry restoration of a control room main fire panel module during surveillance testing which rendered the control room emergency air cleaning system functionally inoperable for approximately five-months in 1988 (50-29/89-06-01, Section-7.1.1).
The second example involved the omission of a restorative procedural step during non-return valve (NRV) logic testing which resulted in the inad-vertent closure of the No, 1 NRV and required a manual reactor. scram to be initiated (50-29/89-06-01, Section 7.1.2)
3.
Unresolved Items No unresolved items were identified this inspection period.
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Table of Contents TABLE OF CONTENTS PAGE 1.
Persons Contacted......................
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2.
Summary of Facility Activities....
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3.
Status of Previous Findings (IP 92701)...............................
3.1 (Closed) Unresolved Item 50-29/89-01-01, Review of Control Room Emergency Air Cleaning System (CREACS) Inoperability....
3.2 (Closed) Unresolved Item 50-29/89-02-01, Review of Licensee
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Resolution of Weaknesses Associa,ted with April 7, 1989 Reactor Startup............
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3.3 (Closed) Unresolved Item 50-29/89-02-02, Review of Personnel Error During Non-Return Valve (NRV) Logic Testing...........
4.
Operational Safety (IP 71707, 71710).............
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4.1 Plant Operations Review........
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4.2 Safety System Review.
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4.3 Inoperable Equipment..
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4.4 Review of Temporary Change Pequests and Mechanical Bypasses.....
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4.5 Review of Switching and Tagging Operations....
4.6 Operational Safety Findings...................................
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5.
Security (IP 71707).............................................
5.1 Observations of Physical Security......
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6.
Plant Operations (IP 71707, 93702, 82201)...........................
6.1 Automatic Reactor Scram on Low Main Coolant System Pressure.....
6.2 Notification of Unusual Event.........
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6.3 Reactor Restart Following Automatic Reactor Scram...............
7.
Maintenance / Surveillance (IP 71710, 61726, 62703, 61700).............
7.1 Licensee Failure to Properly Implement Surveillance Procedures..
7.1.1 Control Room Emergency Air Cleaning System Inoperability......................'................
7.1.2 Personnel Error During NRV Surveillance Testing.......
7.2 Boric Acid Flowpath Blockage..........
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Table of Contents PAGE 8.
Radiological Controls (IP 71707).....................................
8.1 Observation of Radiological Protection and Controls.............
8.2 Degraded High Radiation Exclusion Area Access Control Barrier...
8.3 Maintenance of Radiological Access Control Barrier..............
8.4 Detectable Activity In On-Site Septic System...............
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9.
Licensee Event Reporting (LER) (IP 90712, 92700).........
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9.1 LER 89-03, No. 2 Steam Generator Blowdown Monitor Inoperative...
9.2 LER 89-04, Inoperable Control Room Emergency Air Cleaning System........................................................
9.3 LER '89-05, Manual Reactor Trip Due To Personnel Error During Surveillance Testing..........................................
9.4 LER 89-06 High Radiation Exclusion Area Door Found Open.........
9.5 LER 89-07 Dropped Control Rods Result in Reactor Scram on Low Main Coolant System Pressure.....................
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10. Review of Licensee Response to NRC Initiatives (IP 92703)............
10.1 NRC Information Notice 89-44, Hydrogen Storage on the Roof of the Control Room...........
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11. Review of Periodic and Special Reports (IP 90713)....................
11.1 Annual Radiological Environmental Monitoring....................
12. Management Meetings (IP 30703,40700)................................
- The NRC Inspection Manual inspection procedure (IP) or temporary instruction (TI) or the Region I temporary instruction (R1 TI) that was used as inspection guidance is listed for each applicable report section.
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DETAILS 1.
Persons Contacted
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Yankee Nuclear Power Station l
N. St. Laurent, Plant Superintendent T. Henderson, Assistant Plant Superintendent R. Mellor, Technical Director l
Yankee Atomic Electric Company (YAEC)
J. DeVincentis, Vice President D. Maidrand, Assistant Project Manager The inspector also interviewed other licensee employees during the in-spection, including members of the operations, radiation protection, chemistry, instrument and control, maintenance, reactor engineering, security, training, technical services and general office staffs.
2.
Summary of Facility Activities Yankee Nuclear Power Station (Yankee, YNPS or the plant) operated at 100%
power until April 23, 1989, when the reactor protection system initiated an automatic reactor scram on low main coolant system pressure. The low MCS pressure resulted from Group C control rods dropping into the core during maintenance of the control rod drive mechanism (CRDM) cam motor which was not operating normally.
All safety-related equipment performed as designed. An Unusual Event was appropriately declared and terminated in accordance with station procedures.
The licensee was unable to con-clusively determine the cause of the dropped rods.
Proper maintenance and testing was completed.
Concurrence to restart was provided on April 24 and the reactor was made critical on April 25.
Full power was attained on the morning of April 27.
i Full power operation continued until May 19, 1989, when the licensee com-menced a planned load reduction to perform condenser tube cleaning. At the end of the inspection period on May 21, the licensee had completed the
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tube cleaning operation. The licensee began increasing load on May 22.
I No Region I specialist inspections were conducted during the inspection period.
3.
Status of Previous Inspection Findings i
3.1 (Closed) Unresolved Item 50-29/89-01-01.
Review of Control Room l
Emergency Air Cleaning System (CREACS) Inoperability.
In August 1988, the licensee determined the trip switch for two control room exhaust dampers and an exhaust fan, required to isolate to support
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8 CREACS system operation, had been miswired following March 1988 sur-
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veillance testing, such that the manual remote isolation function
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could not be accomplished.
As a result, the licensee declared the CREACS functionally inoperable for that five month period.
Inspec-tor review of this item indicated the licensee failed to follow ap-
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proved procedures, as required by TS 6.8.1, during the March 1988 l
performance of CREACS surveillance testing. The failure to properly l
implement approved procedures is described further in section 7.1.1.
l This unresolved item is closed.
3.2 (Closed) Unresolved 50-29/89-02-01. Review Licensee Resolution of-Weaknesses Associated with April 7, 1989 Startup.
Inspector review of the April 7,1989 startup noted weaknesses in the estimated criti-cal position (ECP) determination and the adequacy of the associated technical review, Weakness was also noted in the number of non-shift personnel in the control room during the startup evolution.
Inspector review of the April 25 startup indicated improved perform-ance, The reactor was made critical within acceptable reactivity tolerance. The ECP technical review requirements were revised to require operations shift supervisor review and assessment.
Improved control room access was observed during startup as assured by senior management.
Long term corrective actions include strengthening reac-tor engineering support for operations through the development of a computer program for ECP calculations.
This. item is closed.
3.3 (Closed) Unresolved Item 50-29/89-02-02, Review of Personnel Error During Non-Return Valve (NRV) Logic Testing. On April 6, 1989, the licensee failed to perform a restorative procedural step during NRV surveillance testing that resulted in the inadvertent closure of the l
No. 1 NRV and required the reactor to be manually scrammed.
Upon further review of this item, the inspectors determined the root cause of the unplanned NRV closure was licensee failure to follow proce-dures, as required by TS 6.8.1, during the performance of NRV sur-
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veillance testing. The failure to properly implement approved proce-dures is described further in section 7.1.2.
This unresolved item is closed.
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Operational Safety j
4.1 Plant Operations Review I
The inspector observed plant operations during regular and back-
shift tours of the following areas:
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Control Room Safe Shutdown System Building
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Primary Auxiliary Building Fence Line (Protected Area)
l Diesel Generator Rooms Intake Structure
Vital Switchgear Room Turbine Building j
Cable Tray House Spent Fuel Pit (SFP) Building j
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Control room instruments were observed for correlation between chan-nels, proper functioning, and conformance with technical specifica-tions. Alarm conditions in effect and alarms received in the control room were reviewed and discussed with the operators.
Operator. aware-ness and response to these conditions were reviewed. 0perators were found cognizant of board and plant conditions. Control room and shif t manning were compared with Technical Specification require-ments.
Posting and control of radiation, contaminated and high radi-ation areas were inspected.
The use of and compliance with Radiation Work Permits (RWPs) and use of required personnel monitoring devices were checked.
Plant housekeeping controls were observed including control of flammable and other hazardous materials.
During plant tours, logs and records were' reviewed.to ensure compliance with sta-tion procedures, to determine if entries were correctly made, and to verify correct communication of equipment status. These records in-cluded various operating logs, turnover sheets, tagout and temporary-change request logs, and event deportability evaluation requests.
Inspections of the control room were-performed on weekends and back-shifts including April 11-13,.17, 19, 27-28, May 8, 16-17, and 20.
Operators and shift supervisors were alert, attentive and responded appropriately to annunciators and plant conditions.
Personnel were knowledgeable of technical specifications, station procedures, and plant systems. The inspector noted the operating staff demonstrated good initiative in studying and practicing emer-gency operating procedures (EOPs) during reduced work activity times such as backshifts and weekends. ' Licensed personnel maintained posi-
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tive control and observation of operator trainee board manipulations.
Documentation of shift activities was good.
Inspector review of operating logs and turnover sheets indicated good characterization of operating history.
Off-normal conditions, surveillance completed, and equipment performance were appropriately documented.
4.2 Safety System Review The emergency diesel generators, EDG fuel. oil, containment isolation and high and low pressure safety injection systems were reviewed to verify proper alignment and operational status in the standby mode.
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The review included verification that: (i) accessible major _ flow path
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(iii) lubrication and component cooling was proper; and (iv) com-ponents were operable based on a visual inspection of equipment for leakage and general conditions.
No violations or safety concerns
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were identified.
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i 4.3 Inoperable Equipment Actions taken by plant personnel during periods when equipment was i
inoperable were reviewed to verify: Technical Specification limits f
were met; alternate surveillance testing was completed satisfactory-l 11y; and equipment return to service upon completion of repairs was
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proper. This review was completed for the following items:
May 1 - Present:
Incore nuclear instrumentation "B"
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drive was declared inoperable due to problem in withdrawing, advancing and stopping at appropriate positions.
(MRs 89-60, 89-61, 89-62).
Problems were also noted in the movable incore system in inspection report 50-29/
89-02.
May 16 The emergency boiler feed pump was de-
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clared inoperable to isolate and repair a steam-leak on valve HCV-825 (MR 89-1142).
April 11 - April 12 Low Pressure Surge Tank level indication
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instrumentation channel CH-LI-1 was declared inoperable due to anomalous instrumentation drift (MR 89-934).
No violations or safety concerns were identified.
4.4 Review of Temporary Change Requests and Mechanical Bypasses
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Temporary change roquests (TCRs), which were approved in support of l
implementing lifted leads and jumper requests and mechanical by-passes, were reviewed to verify that: controls established by AP 0018,
" Temporary Change Control", were met; no conflict 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 w s prepared if required.
Implementation of the requests i
was reviaed on a sampling basis.
The following requests were re-
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viewed:
TCR 89-169 -- implemented tn May 1, 1989 to remove path selec-
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tion capability for movable incore detector."B" drive. This l
TCR is still open pending corrective maintenance.
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TCR 89-146 -- implemented and restored on April 12,'1989 to sup-
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port maintenance on the low pressure surge tank level indication system.
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TCR 89-149 -- implemented on April 14 and restored on April 18
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to support maintenance on pressurizer wide range level channel indication PR-LI-705A.
4.5 Review of Switching & Tagging Operations The switching and tagging log was reviewed and tagging activities were inspected to verify plant equipment was controlled in accordance with the requirements of AP 0017, " Switching and Tagging of Plant Equipment". The following switching and tagging order was reviewed.89-443 -- issued and completed on April 12 to perform me.inten-
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ance on the low pressure surge tank level transmitter.
4.6 Operational Safety Findings Licensee administrative control of off-normal system configurations by the use of TCR and switching and tagging procedures as reviewed i
above, was in compliance with procedural instructions and was con-sistent with plant safety.
Licensee efforts to minimize active tem-porary change requests and mechanical bypasses is noteworthy.
During NRC inspection 50-29/89-01 section 4 (c)(2), the inspector identified a potential weakness in the safe shutdown diesel battery fastening. Subsequent review by the licensee determined that the current seismic fastening may need additional support bolting. -The licensee issued a nonconformance report to evaluate and implement proper battery fastening. relat ive to the Systematic Evaluation Pro-gram and good engineering practices. The licensee adequately satis-fied regulatory requirements.
The inspector had no further ques-tions.
5.
Security 5.1 Observations of Physical Security Selected aspects of plant physical security were reviewed during regular and backshift hours to verify that controls were in accord-ance 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, ano implementation of access controls including authorization, badging, escorting, and searches. No inadequacies were identified.
Licensee responsiveness to inspection concerns in this area was good.
In response to NRC inspection report 50-29/89-04, the licensee con-ducted training for storeroom personnel regarding security require-ments for deliveries.
Subsequent to NRC inspection 50-29/89-02, the licensee posted control signs to minimize passace from access control areas via doors intended to provide emergency egress only.
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the licensee issued a memorandum to the staff detailing the NRC ac-cess authorization program to clarify the plant position relative to the. program.
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Inspector review of securi.ty. program. implementation indicatedIgood performance.
Personnel' were attentive and ~ responsibly fit: for duty.
The inspector observed proper personnel:and' vehicle search procedures to be' implemented.
6.
Plant Operations 6.1 Automatic Reactor Scram on Low Main Cociant' System Pressure-At 6:30 p.m. on April 23, 1989 with the plant at 100' percent of-rated power, the control. room operators observed that the Group C control rods could not be moved in-either direction. The operating staff initial assessment determined that the control rod ~ drive me-chanism (CRDM) cam motor would not' operate. A maintenance. request (MR 89-996) was initiated and an electrician was called in to inves-tigate.
During initial troubleshooting an attempt was made to cycle the cam motor and move the Group C rods out one step. The " Power Range Loss of Power or Dropped Rod" panalarm was received and at least two of the four Group C control rods were abserved to have dropped into the core. Main coolant system (MCS) temperature and pressure immediately-decreased. At 6:53 p.m., the reactor protection system (RPS) initi-ated a low MCS automatic reactor' scram. The MCS pressure remained above the safety injection actuation signal (SIAS) setpoint. 'The number two emergency diesel generator (EDG) automatically started, as designed, when the turbine tripped subsequent to the reactor trip.
Inspector review of the automatic reactor trip indicated good licen-see operating crew performance. No evidence of personnel error was observed.
Safety related equipment performed as designed.
6.2 Notification of Unusual Event During the sequence of events' described in section.6.1 above, the licensee declared an Unusual Event at 7:10 p.m. on April' 23,1989 in accordance with procedure OP-3300, Rev. 6, " Classification of-Acci-
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dents".
l The states of Vermont and Massachusetts were notified as required.
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The NRC Operations Center was notified at 7:38 p.m. Unusual Event status was terminated at 7:15 p.m.
Inspector review of the event deportability' evaluation request (ERER No. 89-29),- control room records, and discussions with cogni-zant licensee personnel indicated good performance. The licensee
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responded appropriately and accurately identified an'd classified the Unusual Event. The control room operations staff followed appro-priate station procedures and concisely documented the sequence of events in the logbook.
The' inspectors had no further-questions.
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6.3 Reactor Restart Following Automatic Reactor Scram The post-trip report was reviewed by the Plant Operations Review Com-mittee'on April.24. The licensee was unable.to conclusively deter-mine thelcause of the dropped rods. :The most likely root cause was-determined to be a broken compression connector wire supplying power to the brake solenoid. This was attributed to having caused the brake to be applied during rod movement. Additionally, an. area'of bare insulation was noted on the insulation to one of the brake' coil-power wires..The licensee hypothesized that'a voltage drop or elec-trical'short may have occurred causing the-stationary grippers to.
drop the rods. Another possible.cause was determined to have been a.
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-possible momentary ~ opening of the contacts.or a dirty. contact.
The licensee repaired the degraded wiring and initiated post maintenance testing. All equipment and circuits tested correctly.
No inade-
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quacies were identified in the inspection and testing of the CRDM -
circuits and components.
One anomalous equipment condition was identified by the licensee dur-
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ing this event. The supply breaker for Motor Control Center (MCC) 4, Bus 2, was tripped open.
Proper loading was restored in a timely manner.
Licensee failure analysis indicated that the overload unit was trippirg early. The supply breaker was. replaced with_another unit.
Licensee management concurrence to restart was provided on April 24.
At 11:05 a.m. on April 25, the reactor was made critical and entered startup (Mode 2). The generater was phased to the grid (entered Mode
1) at 3:15 p.m. the same day.
The plant. achieved 100% rated power operation at 6:00 a.m. on April 27.
Inspector review of the repairs and startup indicated good perform-
ance. Although absolute root cause failure analysis could not be
'1 determined, tos licensee was effective in identifying the failure
mechanism. The licensee satisfied reporting requirements for Licen-l see Event Reporting (LER 89-07).
Issuance of maintenance requests
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and completion of post-maintenance testing was performed as: required.
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Resolution of technical issues was adequate.
During startup, the licensee accurately determined the estimated critical position (ECP).
Approach to criticality was achieved within acceptable reactivity'
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tolerances. The inspectors had no further questions.
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Maintenance / Surveillance 7.1 Licensee Failure to Prorerly Implement Surveillance Procedures The following two examples of licensee failure to. follow procedures during surveillance testing were identified as unresolved items'in previous reports.
In accordance with the' NRC enforcement criteria, as stated in 10 CFR Part 2, Appendix C, Supplement 1, these procedural noncompliance are considered collec'ively as one violation.
Because the licensee has planned or. 71emented appropriate corrective acticas to prevent occurrence of similar-events as reported in the associated LERs,. written response to this violation is not required.
7.1.1 Control Room Emergency Air Cleaning' System Inoperability.
In accordance with NUREG 0737 task action plan item III.D.3.4, " Control Room Habitability", licensees were re-quired to ensure control room operators would be adequately protected against the effects of a toxic or radioactive gas release, such that the reactor could be operated or shut down safely under design basis accident conditions without personnel receiving radiation exposures in excess of 5 Rem to the whole body (10 CFR 50, App A, general design criterion 19).
In response to this requirement, the licen-see installed the control room emergency air cleaning sys-tem (CREACS) during the 1981 refueling outage.
The CREACS is a non pressurized, high efficiency filtered, 100%
recirculation air cleanup system.
It utilizes portions of the normal control room ventilation ductwork and various control room ventilation damper i.solations, necessary to support CREACS operation, are manually initiated'via the control room main. fire panel.
Specifically, Module SM-30 q
in the main fire panel provides manual isolation.of the
control room exhaust fire dampers, FD-B and FD-E.
l On March 7, 1988, following a design change to the CREACS (EDCR 88-46), the licensee performed CREACS acceptance testing which verified the capability. to menually isolate.
dampers FD-B and FD-E.
The testing was delineated in OP j
4644, " Functional Test of the Fire Detection Instruments-
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tion".
Step C.V.2.a. of the procedure directed the removal or " lifting" of the electrical lead at terminal point 5 of l
the SM-30 module position E-2.
The ensuing steps directed
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the damper switch be placed in the " operated" position and
directed the acknowledgement of anticipated system re-
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sponses.
Step C.V.5.b directed the restoration or "reland-
ing" of the electrical lead to its original circuit con--
i figuration. The lead was properly lifted and system re-
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sponse to damper switch repositioning was as designed.
However, the lead was incorrectly relanded to terminal I
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l#0200 9-point 6 versus point 5.
The wiring discrepancy'was not..
detected by the independent verification _revier required by.
the procedure. :In the miswired configuration the SM-30 module position E-2 function to provide manual remote. iso-
.lation of fire dampers FD-B'and.FD-E, upon CREACS initia-tion, was defeated.. Control room air in-leakage during'.
g CREACS' operation with the fire' dampers in the open position i
would substantially exceed.the maximum assumed' design in-leakage of 34.0 CFM (0.06 volume changes per hour). There-fore, the inability to manually isolate the. dampers remotely...
as designed, rendered the CREACS' functionally inoperable.
The wir4g error was not' identified.and corrected'until the manual damper trip switch failed to operate during survei.1-lance testing of the fire ' detection and suppression ~ systems performed on August 16,:1988.
However, had a radiological event occurred during the period of CREACS inoperability, permanently installed radiation-monitoring.equipmentLwould have alerted operators to continuing elevated control: room radiation levels which would have indic_ated a-CREACS mal -
function. The control room area radiation monitor.provides an above normal alert alarm at-10 mR/hr and a high alarm at 40 mR/hr.
Further, following activation,ofe emergency plan, continuous port'able control ' room area radiation.
monitoring would be initiated.- Due to the ability to-promptly identify abnormally high control. room radiation-levels.. ample-time would have been considerr' available to ensure proper control. room isolation by locally closing the affected dampers without exceeding G0C 19 limit's.
It should be noted that the fire protection system automatic isolation.
y function of the dampers:provided by SM-30 module position:
E-1 was unaffected and, remained operable.
An event deportability evaluation request (ERER No. 88-0005)
was initiated on August'16, to determine if 10 CFR 50.72 or 10 CFR 50.73 reporting criteria-had been met.
Based on a
review of applicable TS which addressed only control room
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ventilation system emergency shutdown requirements but which did not provide CREACS operability requirements, the l
licensee concluded that the event was not-reportable. ~How-L ever, the licensee initiated an internal plant information
report (PIR 88-08). The PIR accurately' depicted the event i
and identified proper initial corrective actions.
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The inspectors performed an independent review of this event and subsequent-licensee corrective actions. 'The in-spectors concurred with the root cause analysis and deter -
mined the-licensee corrective actions to be. comprehensive and complete. However, the inspectors took exception with the licensee deportability determination and requested the licensee revisit the. issue.
Following further review the licensee determined that the event was, in fact, reportable and on February 24, 1989 made an after-the-fact notifica-tion to the NRC Operations Cente.. 'On March 26, 1989 the licensee issued LER 89-04 (see section 9) which properly;
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addressed this event.. Further, the licensee submitted a
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proposed change to TS (proposed change No. 209, dated April-14, 1989)-to reflect the operability and surveillance ~re-quirements applicable to CREACS. The proposed change re-mains under NRR review.
The root cause for the CREACS inoperability was personnel
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error during the restoration of module SM-30 circuitry con-
figuration following the March 7, 1988 system acceptance i
testing. The procedure developed ~ sufficient instruction to i
properly restore the circuitry and also required indepen-dent verification of proper as-left condition..However, the performing individual failed.to follow the procedure during restoration of module SM-30 and the individual per-l forming the independent verification of' the restoration
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failed to identify the circuitry discrepancy.
Failure to follow approved procedures, during CREACS testing, is a-I violation of TS 6.8.1 requirements (50-29/89-06-01, example 1).
7.1.2 Personnel Error During NRV Surveillance Testing On April 6, 1989, during functional testing of the.NRV con-trol system logic, a restorative procedural ~ step was not
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performed which resulted in the inadvertent closure of the
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No. 1 NRV and necessitated the initiation of a manual reac-j tor scram. This event was previously documented in detail,
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in inspection report 50-29/89-02, sections 6.1 and 7.2.
Upon further review of the event and the associated LER j
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(LER 89-05, see section 9), the inspectors determined'the
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root cause of the NRV closure was personnel inattentiveness to procedurai adherence.
Failure to follow procedures, i
during NRV testing, is a violation of TS 6.8.1 requirements -
(50-29/89-06-01, example 2)
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The licensee initiated prompt corrective actions lto this1
'I event. The plant staff was counselled on the importance'of d
procedure compliance and attention.to detail 'during the.
performance of surveillance testing. The affected proce-dure was reviewed and enhanced,. in. format, to' improve. human -
factors considerations.
Other surveillance procedures are being similarly reviewed. The inspectors had no further
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questions.
7.2 Boric Acid Flowpath Blockage During a boric acid transfer on May 4, he licensee discovered a flow-path blockage which prevented the boric acid transfer pump.from estab-lishing a suction path from=the boric acid mix tank (BAMT).through
,CS-MOV-629. 'This flowpath provides the-normal boration pathway to the. low pressure surge tank (lPST) and the safety injection (SI)-
tank.
The licensee satisfied technical specification-(TS) required flow-
. path op'erability by verifying direct pathways from'the SI tank to the
- harging pumps and from the BAMT to the. charging pumps. Maintenance-request number 89-1074 was issued.
Corrective maintenance' involved using a Multiamp variable current test device for approximately one and one-half hours at which time the blockage cleared.
Inspectar review of the event indicated-adequate licensee perform-ance. Maintenance and system restoration was completed in a timely manner. Appropriate TS action statements were entered and exited as required. The licensee determined the blockage to have occurred as a result of boric acid precipitation on the discharge side ~ of the valve, apparently as a result of insufficient flushing of that line after presious system usage. The pathway between CS-MOV-629 and the j
transfer pu'np was not heat traced.
Licensee examination of CS-MOV -
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629 indicated that no through-valve leakage had occurred.
Routine boric acid transfers require flushing the line with demineralized water.
Discussions with cognizant licensee personnel-indicated that'
heat tracing is not necessary with proper flushing.
No evidence of operator error was apparent.
Procedures requiring flushing.were com-pleted as required, but advitional flushing may have been warranted.
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Post-maintenarie testing on May 4 was satisfactorily completed and J
the system was returned to service.
The inspector had no further
. questions.
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I 8.
Radiological Controls 8.1 Observation of Radiological Protection and Controls Radiological controls were reviewed on a routine basis relative to industry radiological standards, administration and radiological con-
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trol procedures, and regulatory requirements.
Selected work evolu-j tions were observed to determine the adequacy of program implemen-
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tation commensurate with the radiological hazards and importance to
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safety.
Independent surveys were performed by the inspector to
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verify the adequacy of radiological controls and instructions to workers.
I 8.2 Degraded High Radiation Exclusion Aree Accesa Control Barrier On April 20, 1989, the licensee identified a high radiation exclusion I
area (HREA) access control door which was open and unattended.
Ex-amination of the door indicated that the locking mechanism had failed.
The licensee secured the door and initiated an aggressive j
safety assessinent and surveillance testing schedule. A maintenance i
request was submitted and repairs were completed the next morning.
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I The licensee evaluation indicated that no inadvertent e.posures or
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unauthorized personnel accesses had c ccurred.
The licensee deter-mined the failed door to be operatle based on initial performance j
testing.
Subsequent surveillance t9 sting (every one-half hour) veri-fied the door to maintain physic:1 access control integrity except that the locking mechanism would fail when tested under certain con-ditions (i.e. willful attempts to defeat the barrier).
The'licen-rae deemed this to have been acceptable based on continued surveil-lance testing and minimum plant staffing during backshift hours.
Inspector review of the licensee evaluation and corrective actions indicated generally good performance.
Although the licensee was ag-gressive and effective in ensuring personnel access was tightly con-trolled, additional compensatory measures were not taken when sur-ve111ance testing demonstrated the door locking mechanism to be de-graded in performing its design function.
Licensee management ac-i knowledged the inspector concerns and stated that a more secure lock-ing mechanism or continuous surveillance was merited. The inspector noted senior licensee management to have been fully involved in all phases of the incident.
Review of the licensee exposure assessment verified that no inadvertent exposures occurred.
On April 20, 1989, the licensee completed an event deportability evaluation report (ERER No. 89-28) which resu' tea in issuing LER 89-06 (as discussed in Section 9).
Aco ss controls subsequent to the incident adequately satisfied Tect.nical Specification requirements.
The inspectors had no further questions.
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8.3 Maintenance of Radiological Access Control Barrier Inspector review of the access control incident detailed in section 8.2 above indicated that the failed aor had been repaired the pre-vious week on April 13, 1989.
That repair had been completed for MR 88-1553, "Old PCA Door, lock does not open from inside" which was submitted by radiation protection personnel on September 16, 1988.
The room provides two points of entry and egress; one is the afore-mentioned door and the other is an electric rollup door, which could be de-energized from outside the room with an electrical switch.
The inspectors expressed concern that seven months lapsed before egress through the degraded door was restored.
Inspector review of radio-logical survey data taken on April 21, 1989, indicated that general area dose rates in the room were as high as 2 Rem / hour.
The initial assessment completed by members of the radiation protec-tion, maintenance, and safety departments did not demonstrate a pro-per safety perspective in addressing this issue.
The licensee ex-planation for the untimely resolution of the maintenance request was that no locksmith was employed onsite and that a new lock assembly had to be ordered.
It was not indicated on the MR that the door con-trolled access to a locked high radiation exclusion area even though the MR was submitted by a member of the RP staff.
Similarly, the cognizant radiation protection engineer (RPE) was not aware of the MR.
The licensee assessmant of the personnel safety hazard concluded that the door not affording egress was "not a safety concern" because the rollup door was normally energized.
The licensee detailed several access control procedures and practices as well as resource-ful egress options.
The. inspector considered the assessm?nt conclu-sions unsatisfactory in that they were inconsistent with the intent of the regulations and good access control practices.
The inspectors met with senior management to discuss tre unsatisfac-tory resolution of this issue Management acknowledged the inspec-tors concerns and initiated prompt corrective measures.
Radiation protection personnel were directed to be more concisa in describing the radiological considerations when submitting MRs. A radiation protection MR logbook was initiated to ensure tracking and timely resolution of MRs affecting radiological safety.
The inspectors noted all safety concerns were identified and are currently being appropriately addressed. The inspectors had no further questions.
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1l 8.4 Detectable Activity In On-Site Septic System On April 21, the. licensee performed routine semi-annual radioanalysis
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of an effluent sample drawn, from the on-site leeching field which
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indicated the presence of trace amounts of low level activity...On
April 28, a bottom sample was. drawn from the leeching field septic
tank. ' The tank has a capacity of 7,000 gallons and at _present it is-l approximately 2/3 full. The.results' indicated the presence of the following activities.
ISOTOPE ACTIVITY UNCERTAINTY Co-60 3.15E-7 uCi/g
+/- 2.02E+00%
Cs-134 1.93E-8 uCi/g
+/- 8.17E+00%
Cs-137 5.81E-8 uCi/g
+/- 6.8E+00%
On May 11, a composite sample of the septic tank was drawn.
The re-sults indicated the presence'of the following activities.
ISOTOPE ACTIVITY UNCERTAINTY Co-60 2.95E-8 uCi/g
+/ 7.0E+00%
Cs-137 7.54E-8 uCi/g
+/- 2.12E+01%
F-18 7.83E-6 uCi/g
+/- 1.66E+01%
Both samples were counted for 50,000 seconds, at a 95% confidence level.
The levels of activity identified do not present a' health concern and are well below the allowable septic disposal. limits by release into sanitary sewage systems, as stated in:10 CFR 20.303.
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The routine semi-annual sampling of the septic system effluent was initiated in 1979.
The licensee stated that no detectable activity had been indicated during previous sampling of this septic system, which was installed in 1979.
An archival record search revealed that sludge from a retired on-site leeching field had been transported to Barnwell', South Carolinia in 1976 & 1978, as LSA (low specific activity) waste. The earliest retrievable record.of pre-shipment sampling was in 1974.
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The licensee is investigating potential radionuclides point (s)' of
entry into the septic system.
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Inspector review of this event indicated that the licensee was con-servative and thorough in assessing the issue. The' sampling methodo_-
logy and analysis technique were technically. sound in characterizing.
the. radiological' significance.
Review of analysis results indicated good counting statistics as evidenced by the long count _ times.
The
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inspectors will continue to evaluate licensee performance'in this
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9.
Licensee Evnt Reporting ('LER).
The inspector reviewed the below-listed licensee event reports (LERs) to determine that with respect'to the general aspects of:the events:
(1) the.
report was submitted in a timely manner; (2) description of the-events was accurate; (3) root cause analysis was' performed; (4) safety implications were considered; and (5)~ corrective actions implemented or planned were sufficient to preclude recurrence of a similar event.
l 9.1 LER 89-03, No. 2 Steam Generator Blowd un iSnitor Inoperative This LER describes a February 9-10, 1989 event in which the No. 2 steam generator blowdown' monitor was inoperable for greater than eight hours without a continuous monitor being placed in service, as required by the TS 3.3.3.1 action statement. On the morning of February 9, a chemistry analysis of the blowdown effluent indicated abnormal results.
.A followup sample, which was not analyzed until the morning of February 10, indicated the monitor was actually moni-toring feedwater that was leaking past a stuck open check valve in a cross connect line. The' licensee concluded the root cause of this reportable event was'the failure of the cross-connect line check
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valve to properly seat, huwever, independent inspector review of this event identified the root cause to be personnel error.
The in-spectors concur that the failure mechanism that permitted feedwater to infiltrate the blowdown line was the failed check talve. Not-withstanding, the TS action statement requirements could have been met initially had the cognizant chemists expeditiessly resampled the blowdown effluent, confirmed the presence of feedwater and taken the appropriate actions to ensure closure of the check valve or alternate continuous monitoring of the blowdown within theLeight hours allowed by TS. A supplement to the LER was planned to be issued following final determination of the check valve failure. The licensee has com-mitted to revisit the root cause determination and incorporate cor-rections, as appropriate, in the supplement.
The inspectors had no further questions at this time.
9.2 LER 89-04, Inoperable Control Room Emergency Air Cleaning System This-LER addressed the licensee failure to properly restore an elec-trical circuit during CREACS surveillance testing, which rendered.the CREACS inoperable for five months in 1988 (March - August). This event was previously identified as an unresolved item in inspection
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report 50-29/89-01. The item has been resolved as a violation in this inspection report (see sections 7.1 and 7.1.1 for more detail).
This LER fulfilled the above criteria with the exception of timely reporting as detailed in section 7.1.1. The inspectors had no further questions.
s 9.3 LER 89-05, Manual Reactor Trip Due to Personnel Error During Surveil-lance U sting This LER addressed the April 6,1989 manual reactor scram following the unplanned closure of an NRV during surveillance testing.
This event was previously identified as an unresolved item in inspection report 50-29/89-02. The item has been resolved as a violation in this inspection report (see sections 7.1 and 7.1.2 for more detail).
This event fulfilled the above criteria and no deficiencies were noted.
9.4 LER 89-06, High Radiation Exclusion Area Door Found Open This LER described the condition in which a radiological access con-trol door was identified as being open and unattended contrary to Technical Specification requirements.
This incident is discussed in detail is section 8.2 of tnis report.
The licensee adequately cor-rected the mechanical failure.
No inadvertent exposures occurred.
The LER fulfilled the above criteria and no reporting deficiencies were identified.
9.5 _LER 89-07, Dropped Control Rods Result in Reactor Scram on Low Main Coolant System Pressure This LER described the April 23, 1989, event in which the reactor scrammed while trouble shooting cam motor problems with the Group C control rods.
The reactor scram, maintenance, and restart are dis-cussed in detail in section 6.0.
Technical considerations were ade-quately addressed.
Concurrence for restart received the proper level of management review.
The LER fulfilled the above criteria and no reporting deficiencies were identified.
10.
Review of Licensee Response to NRC Initiatives 10.1 NRC Information Notice 89-44, Hydrogen Storage on the Roof of the Control Room This NRC information notice details potential safety problems associ-ated with the onsite hydrogen storage near safety related equipment /
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structures. Ar requested, the licensee provided the NRC with a writ-
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ten response detailing hydrogen storage locations, quantities and distances from safety related equipment and structures.
Included in
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the response was-a description'of.the hydrogen storage facility de-sign features, requirements for the use of non-sparking tool.s and.
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physical : cylinder ' restraints.
Inspector review verified hydrogen storage to'_ exist ~as ~ described in the response. At Yankee Nuclear Power Station, the hydrogen storage area is different from the conditions described in the notice.
In-spector observation of hydrogen storage and use indicated no apparent-safety hazard. No administrative.or procedural inadequacies were-identified.. Safety control signs were appropriately posted.
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tive ventilation existed such that hydrogen would be evacuated from-
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the area in the event of.a. leak. The inspector had no fur.ther ques--
tions.
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i 11. Review of periodic and Special Reports:
I Upon receipt, the inspector reviewed periodic and special reports sub-mitted pursuant to Technical Specifications.
This review verified, as._
applicable: (1) that the reported information was valid and included the.
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NRC required data; (2) that test results and supporting information were q
consistent with design predictions and performance.. specification; and
(3) that planned corrective actions were adequate'for resolution of the problem. The inspector 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, April and May, 1989.
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Annual Radiological Environmental Monitoring Report for 1988.
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BYR 89-79 indicated it was properly submitted in accordance with technical
'I specification 6.9.5 (a).
Review of. analysis results provided no indica-
.i tion of an inadvertent radiological release offsite.
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12. Management Meetings l
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At periodic intervals during this. inspection, meetings were held with
senior plant management to discuss the findings. A summary of findings for the-report period was also discussed at the conclusion of the inspec-l.
tion and prior to report issuance.
No proprietary information was iden-
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tified as being included in the report.
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