IR 05000271/1989022
| ML20006E751 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 02/09/1990 |
| From: | Eapen P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20006E744 | List: |
| References | |
| 50-271-89-22, NUDOCS 9002260313 | |
| Download: ML20006E751 (15) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report'No.
50-271/89-22 Docket'No.
50-271 License No. DPR-28 Licensee:
Vermont Yankee Nuclear Power Corporation RD 5, Box 169
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Brattleboro, Vermont 05301 Facility:
Vermont Yankee Nuclear Power Station Inspection At:.Vernon,' Vermont Inspection Conducted: December 12, 1989 - January 22, 1990 Inspectors:
Harold Eichenholz, Senior Resident Inspector John Macdonald,' Resident Inspector Thomas G. Hiltz, Reactor Engineer Approved by:
Nkb D /q/fD P. K. Eapen, Chief / Reactor Projects Section 3A 7Date Inspection Summary:
Inspection on December 12, 1989 - January 22. 1990 (Report No. 50-271/89-22)
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; licensee event re-ports; and, periodic reports.
Results:
1.
General Conclusions on Adequacy, Strength or Weakness in Licensee Programs The licensee exhibited a continuing commitment to resolution of NRC in-spection issues (Section 3).
The licensee response to the loss of the second of four drywell air hand-ling units was appropriate and well coordinated (Section 6.1).
~ Additionally the initial Fitness For Duty training properly addressed the program attributes and was completed prior to January 3,1990 (Section 5.2).
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90022603139 gO DR ADOCK
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i TABLE OF CONTENTS l
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1.
Persons Contacted....................................................
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2.
. S umma ry o f Fa c i l i ty Ac t i v i t i e s.......................................
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3.
Status of Previous Inspection Findings (IP 92701,92702*)............
3.1 (Closed) Unresolved Item 88-06-01: Review of Independent Surveillance Program Audit....................................
3.2 (Closed) Unresolved Item 88-14-02: Licensee Actions to Resolved Fi re Protection Program Deficiencies..........................
3.3 (Closed) Inspector Follow Item 88-02-03: Drywell Controls During
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Spent Fuel Movements..........................................
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3.4 (Closed) Unresolved Item 88-21-01: Review Auxiliary Operator Continuing Training Program Upgrades..........................
3.5 (Closed) Unresolved Item 89-09-01: Ir. corporation of Vendor Recommendations for MOV Cycling Limitations...................
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4.
Operational Safety (IP 71707,71710).................................
4.1 Plant Operations Rev1ew.........................................
4.2 Safety System Rev1ew.............................................
4.3 Inoperable Equipment............................................
4.4 Review of Tempora ry Modi fications...............................
4.5 Review of Switching and Tagging Operations......................
4.6 Operational Safety Findings.....................................
4.7 Conclusion......................................................
5.
Security (IP 71707)......................................
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5.1 Observations of Physical Security...............................
5.2 Fitness For Duty Program (TI 2515/104)..........................
6.
Plant Operations (IP 71707,93702,71710)............................
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6.1 Drywell. Air Handling Unit Inoperability.........................
6.2. Vital ac Motor Generator Set DC Drive Unavailability............
6.3 Standby Liquid Control System Safety System Walkdown............
7.
Maintenance / Surveillance (IP 71710,61726,62703)....................
7.1 Refuel Floor Radiation Moni tor Inoperability....................
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Table of Contents P
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PAGE 8.
Licensee Event Reporting (LER) (IP 93702)............................
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8.1 LER 89-24, Revision 1...........................................
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8.2 LER 89-25.......................................................
II 8.3.LER89-26.......................................................
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Review of Periodic and Special Reports (IP 71707)...................,
10. Ma nagement Meeti ng s (I P 30703).......................................
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- The NRC Inspection Manual inspection procedure (IP) that was used as inspec-
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tion guidance is listed for each applicable report section.
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DETAILS
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1.
Persons Contacted
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i Interviews and discussions were conducted with members of the' licensee j
staff and management during the report period to obtain information per-tinent to the areas inspected.
Inspection findings were discussed peri-l odically with the management and supervisory personnel listed below.
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Mr. R. Grippardi Quality Assurance Supervisor
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Mr. S. Jefferson, Assistant to Plant Manager Mr. J. Herron, Operations Supervisor Mr. R. D. Legere, Acting Maintenance Supervisor Mr. R. Pagodin, Technical Services Superintendent
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Mr. J. Pelletier, Plant Manager
Mr. D. Porter, Shift Supervisor Mr. R. Wanczyk, Operations Superintendent Mr. T. Watson, IV Supervisor Mr. W. Wittmer, Acting Maintenance Superintendent
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2.
Summary of Facility Activities Vermont Yankee Nuclear Power Station (VYNPS or the plant) continued full power operations during this report period. Throughout the period, short
term scheduled power reductions to 80-95% of full power were conducted weekly to perform routine surveillances of control rod drives, main tur-bine and by pass valves. On December 13, power was reduced to 97% with the reactor in a limiting control rod pattern. On December 16, power was reduced to 80% to exercise control rod and to make a rod pattern adjust-ment to facilitate 100% power operation. On December 18, reactor power was briefly reduced to 92% power following the observance of. increasing
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drywell temperature, pressure and particulate readings.
It was quickly H
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. determined that the increasing trends were resultant from the loss of dry-1:
well air handling unit, RRU4 as discussed in Section 6.1.
The reactor was
N returned to 100% power later on December 18. On January 6, power was re-I'
duced to 62% to accomplish a rod pattern exchange and conduct main steam
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isolation valve (MSIV) full closure surveillance testing. After satisfac-
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tory completion of these tasks, the reactor was returned to full power on January 7,1990.
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r On December 21, the licensee notified the NRC Operations Center via the.
Emergency Notification System (ENS) following a toxic gas monitoring (TGM)
l-system actuation, which by. procedure required control room operators to don self-contained breathing apparatus. This notification was made in
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accordance with 10 CFR 50.72 criteria. Additional 10 CFR 50.72 notifica-tions were made on December 23 and January 5, when spurious spikes of the west refuel floor radiation monitor resulted in Group III primary contain-ment isolation system actuations and on January 12, when the Ames Hill l
communications transmitter was removed from service to perform mainten-anc.
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During the report period Mr. Patrick Donnelly, VY Maintenance Superinten-
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dent accepted a position with the Yankee Atomic Electric Company. Mr.
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William Wittmer, who had been acting as Maintenance Superintendent while j
Mr. Donnelly was en special assignment to the Plant Manager, will remain in that position until a permanent successor has been named.
Effective December 17, Mr. Geoffrey E. Grant, senior resident inspector l
(SRI) at VYNPS, was promoted to Senior Operations Specialist in the Per-i formance. Evaluation Branch of the Office of Nuclear Reactor Regulation
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(NRR).
On December 31, Mr. Harold Eichenholz assumed the SRI position at VYNPS. Mr. Eichenholz was previously the SRI at Yankee Nuclear Power Sta-
tion.
I 3.
Status of previous Inspection Findings 3.1 (Closed) Unresolved Item 88-06-01: Review of Independent Surveillance Program Audit.
In Inspection Reports 50-271/87-23 and 88-03, the.
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inspectors documented several missed Technical Specification (TS)
required surveillances, as well as potential programmatic weaknesses
in the surveillance program. As a result of these issues the 11cen-
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see committed to perform an independent audit of the surveillance
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program.
The audit was accomplished by an independent engineering consulting firm and was completed on' December 21, 1988. The audit
- encompassed a review of TS to identify all required surveillances; a review of the Master Surveillance List (MSL) to ver.ify coincidence between required surveillances, MSL test number, and procedure num-ber; and a review of surveillance procedures to verify the test re-quirements were properly addressed.
The audit report identified de-ficiencies which were mostly administrative in nature.
The deficien-cies were categorized and properly disposed by the licensee.
The independent audit represented a significant licensee commitment to establish surveillance program fidelity and was typical of the
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licensee approach to issue resolution. The inspectors had no further c
questions. This item is closed.
3.2 (Closed) Unresolved Item 88-14-02: Licensee Actions to Resolve Fire Protection Program Deficiencies. On September 20, 1988 the licensee failed to post a firewatch following the removal of the recirculation pump motor generator (M-G) set foam suppression system from the auto-matic mode of operation. A notice of violation (IR 88-14, violation 88-14-01) was subsequently issued for failure to comply with TS re-quirements.
Inspector review of this event identified certain fire protection program deficiencies which may have contributed to this violation.
Specifically, TS and other design bases documents failed to provide definitive fire suppression system operability guidance.
Additionally, operator and shift engineer fire protection program training was determined to be inadequate.
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i The licensee concurred with the inspector observations and took appropriate corrective actions. A matrix is being developed which
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In the interim a standing order was issued regarding the
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recirculation pump M-G set foam suppression system.
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clearly defines the parameters above. Additionally, operator and i
shift engineer fire protection program has been expanded.
Initial
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training has been improved and fire protection program training has
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L been added to the requalification training cycle.
The fire protec-L tion coordinator has been attending the training sessions to identify areas of potential improvement.
The licensee _ corrective actions have appropriately addressed this
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issue.
The inspector had no further questions.
This item is closed.
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3.3 (Closed) Inspector Follow Item 88-02-03: Orywell Controls During Spent Fuel Movements.
This issue concerns licensee radiological con-trols for access to and work in the drywell during spent or irradi-ated fuel movement.
Specifically, the inspector noted the need for a procedure and appropriate training to provide dctailed worker actions
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and responsibilities during ngrmal and potential emergency operations
during fuel movement.
The issue was also addressed during the maintenance team inspection
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and a new unresolved item was issued (88-80-02) which fully encom-
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passes the original item. Therefore, the original item will be closed and the issue of drywell radiological controls during fuel movement will be tracked via Unresolved Item 88-80-02.
This item is closed.
3.4 (Closed) Unresolved Item 88-21-01: Review Auxiliary Operator Continu-ing Training Program Upgrades.
During a review of non-licensed
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operator training program conformance with NUREG-1220, " Training Re-view Criteria and Procedures, the inspector noted weaknesses in the auxiliary operator (AO) continuing training program.
Specifically, A0 continuing training did not incorporate difficulty, importance, or frequency (DIF) factors _ into the job task analysis (JTA). These fac-
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tors, known as DIFs, are used to evaluate which tasks should be in-cluded in the continuing training program. The licensee has since incorporated these factors into the-systematic approach to the train-ing program for A0 continuing training and developed appropriate training modules. Inspector discussions with training personnel in-dicated that the two year cycle content list of the A0 continuing training program, developed from the job task analysis DIF matrix, was appropriate. The inspectors will continue to monitor A0 field performance as well as training programs in the future. This item is closed.
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3.5 (Closed) Unresolved Item 89-09-01: Incorporation of Vendor Recommenda-tions for MOV Cycling Limitations. On June 7, 1989, motor operated
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valve RCIC-21 failed during surveillance testing. Root cause an-
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alysis revealed that the actuator motor suffered accelerated electri-cal degradation from excessive cycling over a short duration during outage valve performance testing. This conclusion was based on physical evidence, as well as revised duty cycle limit information gained from the vendor. The updated vendor information limited the summation of valve stroke times in one hour +.o less than half the duty cycle without a rest period (nominally one hour).
The licensee s
has since_ incorporated this information into the appropriate MOV in-spections and performance test procedures.
t Of note, during inspector review of this event, was that the vendor had not made this information available through the issuance of generic industry documentation.
Rather, the licensae obtained the information from the vendor during causal analysis. Additionally, the licensee's improved vendor information program (AP 0312) requires increased licensee / vendor interaction to improve information acquisi-i, tion from the vendor. The inspectors had no further questions.
This item is closed.
4.
Operational Safety 4.1 Plant Operations Review
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The inspector observed plant operations during regular and backshift tours of the following areas:
Control Room Cable Vault Reactor Building Fence Line (Protected Area)
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Diesel Generator Rooms Intake Structure Vital Switchgear Room Turbine Building Control room instruments were observed for consistency between chan-nels, proper functioning, and conformance with technical specifica-tions.
Existing alarms and others received in the control room were reviewed and discussed with the operators.
Operator awareness and response to these conditions were reviewed. Operators were cognizant of board and plant conditions.
Control room and shift manning were compared with technical specification requirements.
Posting and con-trol of radiation, contaminated and high radiation areas were in-spected. Use of and compliance with radiation work permits and use of required personnel monitoring devices were checked.
Plant house-keeping controls were observed including control of flammable and other hazardous materials. During plant tours, logs and records were
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reviewed to ensure compliance with station procedures, to determine if entries were correctly made, and to verify correct communication of equipment status. These records included various operating logs, turnover sheets, tagout and jumper logs, and potential reportable i
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occurrence reports.
Inspections of the control room were performed on weekends and backshif ts including December 19-20, 1989 and January
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3, 8, 9, 10, 12, and 18, 1990.
" Deep backshift" inspections were conducted as follows:
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Date Time
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1/10-11/90 11:15 p.m. - 5:00 a.m.
1/18/90 10:00 p.m. - 12:00 a.m.
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Operators and shift supervisors were-alert, attentive and responded appropriately to annunciators and plant conditions.
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4.2 Safety System Review Portions of the emergency diesel generator, reactor core isolation
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cooling, core spray, residual heat removal, standby gas treatment,
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residualrheat removal service water, safety related electrical, and high pressure coolant injection systems were reviewed to verify alignment and operational status in the standby mode. The review
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included verification that (i) accessible major flow path valves were
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correctly positioned, (ii) power supplies were energized, (iii) lubri-
- g cation and component cooling were proper, and (iv) components were
operable based on a visual inspection of equipment for_ leakage and
. general conditions.
4.3 Inoperable Equipment Actions taken by plant personnel during periods when equipment was inoperable were reviewed to verify:= technical specification limits were met; alternate surveillance testing was completed satisfactor-ily; and, equipment return to service upon completion of repairs was proper.
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Date Out Date In System 12/18 ROOS *
RRV-4 12/23 12/28
"A" refuel floor rad monitor 1/5 ROOS **
"A" refuel floor rad. monitor
- RRU-4 will remain out of service until the next reactor shutdown.
- The "A" refuel floor radiation monitor remained in by pass while being monitored by diagnostic recording equipment through the end of
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the report period.
4.4 Review of Temporary Modifications Temporary modifications were reviewed to verify that controls estab-
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lished by AP 0020 were met, no conflict with technical specifications were created, safety evaluations were prepared in accordance with 10 l-l
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CFR 50.59 if required, and requests were reviewed and approved prior
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to installation.
Implementation of the requests was reviewed on a i
sampling basis.
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4.5 Review of Switching & Tagging Operations I-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 0140, Vermont Local Control Switching Rules.
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4.6 Operational Safety Findings
Licensee administrative control of off-normal system configurations by the use of temporary modifications and switching and tagging pro-cedures, as reviewed in Sections 4.4 and 4.5, was in compliance with procedural instructions and was consistent with plant safety. Back-shift inspections have consistently-found operators to be alert and
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attentive. Operations are routinely conducted in a professional man-ner in an atmosphere of quiet control and competence. With the ex-cep'. ion of isolated instances, as discussed in Section 6.3, overall plant cleanliness and material condition continue to be good.
4.7 Conclusion
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No unacceptable conditions were observed in any of the above areas reviewed.
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Security
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5.1 Observations of Physical Security Selected aspects of plant phy'ical security were reviewed during
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regular and backshift hours to verify that controls.were in accor-N.
dance with the security plan and approved procedures.
This review
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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.
No inadequacies were identified.
L 5.2 Fitness For Duty Program (TI 2515/104)
On December 13, the inspector attended an initial licensee Fitness For Duty Program (FFD) training presentation.
The presentation con-sisted of lectures, videos, and visual aids. The session was
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authority. The training session effectively described the policy and l
procedures utilized to implement the FF0 program.
The potential hazards of controlled substance abuse with respect to individual and
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public health were discussed.- The Employee Assistance Program (EAP)
was addressed in detail including various scenarios in which an in-
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dividual would seek or be requested to seek EAP assistance.
Super-visory training on behavioral observation is presented as a separate
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management training program and therefore was not included in the FFD
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presentation attended by the NR0 inspector.
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The presentation was well attended (approximately 100 persons). The.
instructors were well prepared and encouraged open end frank question
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and. answer sessions.
The inspector had no further questions.
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6.
Plant Operations
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F 6.1 Drywell Air Handling Unit Inoperability J
On December 18, RRU-4, the second of four drywell air handling units failed.
Previously, on.luly 30,1989, RRU-1 was secured following a fan motor shaft failure.
The RRU-4 experienced an electrical fault-
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in the fan motor.
Loss of RRU-4 resulted in an increase in drywell i
temperature, pressure and indicated particulate activity on the dry-
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well-radiation monitor. Control room operators briefly reduced resc-tor power to 92% until drywell temperature and pressure stabilized.
Increased particulate levels were due to changing drywell air flows resulting from the failed RRV.
Following evaluation of these para-meter changes, power was increased to 100% on the same day.
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'Due to the extremely cold winter atmospheric ambient temperatures, the effect of the loss of the second drywell air handling unit was minimized.
However, the licensee initiated daily drywell temperature prot'ile trending.
Periodically throughout the inspection period, drywell temperature as obtained from equipment qualified temperature indicators TI16-19-30A and B slightly exceeded the 160 degrees F i
entry condition for emergency operating procedure (EOP) OE 3103, i
"Drywell Pressure and Temperature Control Procedure." For the tem--
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peratures indicated, OE 3103 only required continued drywell tempera-ture monitoring. All drywell temperatures trended have remained be-low existing equipment qualification program analysis.
Although t'ne unavailability of two RRVs is not adversely impacting operations during the winter months, the condition will' ultimately
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require a plant shutdown when the outside air temperature rises. The licensee is currently planning a short duration maintenance outage in early March.
During this maintenance outage, the licensee plans
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to restore the drywell air handling units.
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Licensee response to the loss of the second RRU has been appropriate.
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Daily drywell temperature trending was re-initiated and operators appropriately identified conditions for entry into the E0P to respond
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to the observed drywell temperature condition. Additionally, plant L
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anagement expeditiously conducted contingency and planning meetings m
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to address this equipment loss. The inspector nad no further ques-
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tions.
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6.2 Vital AC Motor Generator (MG) Set DC Drive Unavailability
On January 17, operators identified that the AC vital motor generator (MG) set DC drive tachometer belt had broken. The tachometer belt
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provides speed feedback to the DC motor speed controller. Therefore, on a loss of the AC prime mover, the DC motor would sense a continu-
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ous low frequency condition (without the DC drive tachometer belt)
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and overspeed in an attempt to recover frequency. -To prevent such a
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scenario, the licensee opened the DC motor supply breaker. This ac-tion eliminates DC power to the MG set and has no impact on normal
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operations. With offsite power available, loss.of the normal AC power supply to the MG set would result in the automatic transfer to the' alternate power supply (MCC9A). The transfer would create a brief power interruption which may cause the recirculation pump MG set scoop tubes, the feedwater regulators and electrical pressure regulator to lock up.
Lock up of these components could cause an automatic reactor scram.
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During a loss of offsite power with'the diesel generators available,
power to the vital MG set would be lost for approximately 13 seconds until the diesel generators speeds up to accept emergency loads.
Additionally, as a result of previous equipment environmental quali-
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fication analysis which questioned the operability of the vital MG set DC drive in a post accident environment, the licensee had in-i stalled the alternate AC source via MCC9A.
Therefore, vital MG set response without DC power available in a design bases accident scenario has been previously evaluated.
However, if AC power re-mained unavailable for an extended period of time as,in a beyond de-sign bases station blackout scenario, control room RCIC reactor vessel level and pressure control would be unavailable, and would
necessitate operation of the RCIC system remotely from the Alternate
Shutdown Station.
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The licensee has fully evaluated the potential impact of removing the DC drive capability of the AC vital MG set.
It should be noted the vital MG set is not described in technical specifications.
It is
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described in the electrical distribution section of the FSAR (Section 8.4) but is_ not described in the safety design bases section.
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also appeared that an FSAR revision was not made to describe the de-sign bases of the alternate AC power source for the vital MG set in-stalled via engineering design change request (EDCR)85-404.
The broken DC drive tachometer belt is scheduled to be replaced during
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the short duration outage scheduled for early March. The inspectors will further review the licensee actions related to the vital MG set during future NRC inspections.
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6.3 Standby Liquid Control System Safety System Walkdown j
The inspectors performed a periodic detailed walkdown of all access-
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ible portions of the Standby Liquid Control (SLC) system.
All valves were properly positioned and labeled and appropriate flow-
paths were aligned.
Equipment supports and hangers were properly
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L made up and base plate bolting was observed to be fully engaged.
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Instrumentation was observed to be properly calibrated at approved setpoints and critical heat tracing circuitry was observed to be
functioning properly.
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Circuit breaker cubicles and lighting panels were inspected to ensure
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proper position and freedom of movement.
During these' inspections s
small amounts of loose dust and grit were observed in electrical com-
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partments located on the reactor building ground level. As a result
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the inspector along with the Operations Supervisor conducted an ex-
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tensive inspection of electrical panels throughout the reactor build-inD. The inspection revealed that only the compartments on the a
ground floor had any dust buildup.
The inspector concluded this was a result of dust generated from floor resurfacing activities i.) torus area and the lower level of the corner-rooms. The licensee has in-
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creased dust containment measures and is establishing increased in-spections of electrical compartments during the work effort.
The licensee was responsive to the inspector observations and it re-
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flected management expectations of exceptional plant housekeeping standards. -The inspector had no further questions.
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7.
Maintenance / Surveillance
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7.1 Refuel Floor Radiation Monitor Inoperability On December 23 and again on January 5, the "A" refuel floor radiation monitor experienced spurious upscale spikes which resulted in group III primary containment isolation system actuations and subsequent.
standby gas treatment system initiations. All systems responded as designed and the events were not operationally significant.
However, in-an attempt to identify the root cause of these spikes the licensee conducted extensive causal analysis of the events.
Follow-p ing the December 23 event, there were indications that the initial and ensuing spikes were the result of intermittent perturbations in electrical cable and connector performance. The connectors were
[s cleaned and the unit was functionally tested satisfactorily.
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radiation monitor was returned to service on December 28. However, following the January 5 event, there were no. indications of equipment malfunction.
The monitor was maintained in the bypass position and
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L diagnostic recording equipment was connected which will determine whether further spurious spikes originate from the trip unit, indi-cator, or the sensor / converter.
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The licensee suspects the sensor / converter due to its recent end of
life. failures (see Section 8.2).
However, due to the apparent rela-
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tionship between the monitor spiking and connector performance in the i
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December 23 event, the licensee will-continue to record diagnostic data until a more definitive failure determination can be completed.
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The inspector will follow licensee's action in this regard during i
future NRC inspections.
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The abuve detailed diagnostic efforts for the refuel floor radiation monitor is typical of licensee's conservative actions to troublesh07t
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equipment failures.
8.
Licensee Event Reporting (LER)
1he inspector reviewed the Licensee Event Reports (LERs) listed below 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.
8.1 LER 89-24, Revision 1
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LER 89-24, Revision 1,_" Missed Residual Heat Removal (RHR) System
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Valve Leakage Due to Incomplete Procedure Review," readdresses the September 13, 1989, licensee discovery that valve RHR-18 had not been leak' tested during the 1989 refueling outage, as required by the In-service Test (IST) program. As previously documented in Sections 7.1
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and 10.5 of Inspection Report 89-17, RHR-18 is the inboard contain-
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ment isolation valve on the RHR shutdown cooling system.- The shut-down cooling system is also equipped with an outboard containment
isolation valve, RHR-17. Continuing licensee review of design bases documentation, including Technical Specifications-(TS) and ASME B31.1, 1967 edition, concluded pressure isolation valves (PIVs) were not a TS design bases and the shutdown cooling system was designed and constructed in excess of applicable Code requirements. Addi-
-tionally, recently issued generic letter, GL 89-04, " Guidance on De-veloping Acceptable IST Programs," required that only PIVs listed in TS be included in the IST program. Therefore, the licensee has de-termined RHR-18 should not be included in the IST program and has-taken actions to remove the testing requirement. This LER provided extensive information into the licensing bases for the plant and i
properly identified the ASME code required safety features of the hz shutdown cooling subsystem of the RHR system. No deficiencies were identified.
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8.2 LER 89-25 I
LER 89-25, " Inadvertent Primary Containment Isolation System Actu-
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ation Due to a Failure of a Reactor Building Ventilation Monitor Sen-
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sor/ Converter," addresses the November 28, 1989 Group -III primary containment isolation system (PCIS) actuation and subsequent Standby Gas Treatment System (SBGT) initiation. The Group III PCIS isolation
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signal was received following the upscale failure of the "B" reactor building ventilation radiation monitor.
Investigation revealed the
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signal sensor / converter component of the radiation monitor had ex-
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perienced an end-of-life failure. The failed component was subse-L quently replaced and the radiation monitor was returned to service on November 29. Corrective actions taken to address end of life radi-ation monitor component were appropriate.
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8.3 LER 89-26 LER 89-26, " Inadvertent PCIS Actuations Due to Spikes on a Refuel Floor Radiation Monitor," addresses the December 23, 1989 and January l
5, 1990 Group III PCIS actuations and SBGT system initiations.
The Group III PCIS isolations were the result of spurious upscale spikes
of the "A" refuel floor radiation monitor. The maintenance aspects
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of these events are detailed in Section 7.1, No deficiencies were identified.
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9.
Review of Periodic and Special Reports
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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 NRC-required data; (2) that test results and supporting information were 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 re-l viewed:
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Monthly Statistical Report for Plant Operations for the Month of
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l December 1989.
1.
Feedwater leakage Detection System Monthly Performance Summary for
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December 1989.
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t 10. Management Meetings 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-tion and prior to report-issuance.
No proprietary information was iden-tified as being included in the report.
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