IR 05000271/1990001
| ML20034B765 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 04/18/1990 |
| From: | Pasciak W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20034B764 | List: |
| References | |
| 50-271-90-01, 50-271-90-1, NUDOCS 9004300339 | |
| Download: ML20034B765 (23) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No.
50-271/90-01 Docket No.
50-271 License No. DPR-28 Licensee:
Vermont Yankee Nuclear Power Corporation RD 5, Box 169 Brattleboro, Vermont 05301 Facility:
Vermont Yankee Nuclear Power Station Inspection At: Vernon, Vermont Inspection Conducted: January 23 - March 12,1990 Inspectors:
Harold Eichenholz, Senior Resident Inspector John B. Macdonald, Resident Inspector Thomas G. Hiltz, Resident Inspector Designate Edward H tier, Office of Nuclear Reactor Regulation Approved by:
( A 1_ 3 u-- u (d[Po W. Fasciak, Acting Chief, Reactor Projects Section 3A Date Inspection Summary:
Inspection on January 23 - March 12,- 1990 (Report No.
50-271/90-01)
Areas Inspected:
Routine inspection on daytime and backshif ts by resident in-spectors of:
actions on previous inspection findings; operational safety; security; events requiring telephone notification to the NRC; plant operations; maintenance and surveillance; licensee response to NRC initiatives; periodic reports; and organization and administration, i
l Results:
1.
General Conclusions on Adecuacy, Strength or Weakness in Licensee Programs The licensee continues to exhibit strong performance in the area of plant operations which was demonstrated by continuing and appropriate attention t
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to off-normal plant performance involving high drywell temperature (Sec-tion 4.6) and failed fuel indications (Section 7.1).
The development and diligent pursuit of issues involving inaccurate background investigations l
l indicated good management oversight in the area of security (Section 5,2).
l Operator performance in responding to the loss of the Turbine Building l
Closed Cooling Water System was especially noteworthy (Section 7.3).
The
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licensee's maintenance program demonstrated a proper questioning approach to plant safety by the conduct of an accelerated inspection frequency on the "B" Emergency Diesel Generator's vertical drive assembly aad the man-ner in which this equipment deficiency was resolved (Section 8,2).
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r 2.' ' Unresolved Items One unresolved item was identified during'this inspection period:
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The NRC will review Vermont'. Yankee's plans:for revising their.Emer-
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gency Diesel Generator overhaul program (Section 8.2).-
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TABLE OF CONTENTS-PAGE 1.
Persons Contacted.....................................................
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S umma ry o f _ Fa c i l i ty Ac t i vi t i e s.......................................
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3.
Status of Previous Inspection Findings (IP 92701,92702*)............-
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3.1 (Closed) Unresolved Item 87-12-03: Control Room Habitability l'
Review.....................................................-..'...
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3.2. (Closed) Violation 89-08-05: Failure to Conduct a; Proper Package-
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Search...................................................
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3.3 (Update) Unresolved Item 89-01-04: Resolve Design Basis Questions for Torus Vacuum Breakers...........................
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Operational Safety (IP 71707,40500).................................
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i 4.1 Plant Operations Review.........................................
4.2 Safety System Review............................................
5-4.3 Inoperable Equipment............................................
5-i 4.4 Review of Tempcrary Modifications...............................
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4.5 Review of Switching and Tagging Operations......................
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l 4.6 Operational Safety Findings............
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Security (IP 71707, 93702)...........................................
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5.1 Observations of Physical Security...............................
L 5.2 Inaccurate Background Investigations............................
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Events Requiring Telephone Notification to the NRC -(IP 93702)........
6.1 Improper Authorized Access......................................
9-6.2 Actuations of Primary Containment and Standby Gas Treatment Systems.......................................................
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Plant Operations (IP 71707,93702,35502,.40500).....................
7.1 Fu e l Fa i l u re I nd i c a ti o n s........................................
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7.2 Primary Containment I solation System Actuation..................
7.3 Loss of Turbine Building Closed Cooling Water...................
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Maintenance / Surveillance (IP 61726,62703,35502,40500).............
8.1 Emergency Diesel Generator (EDG) Operational Readiness D e nio n s t r a t i o n.................................................
8.2 EDG Maintenance.................................................
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Table.of Contents-
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9.
Review of-Licensee Response to' NRC Ini tiatives.......................
16-i 9.1 Review of NUREG-0737_ Commitments (TI 2515/65)...................
9.2 Installation of a Hardened Wetwell Vent (IP 92701)...............
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Review of Periodic-an'd Special Reports (IP 90713)....................
17-11. Organization and. Administration (IP 35701)............-................
12. Management Meetings (IP 30703).......................................
- The NRC Inspection Manual inspection procedure (IP) or temporary instruction
(TI)'that was used as inspection guidance is listed for each applicable report-
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section.
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i DETAILS.
l 1.
Persons Contacted Interviews and discussions were conducted with members of.the licensee's staff and management during the report period to obtain-information per-i tinent to the areas inspected.
Inspection findings were discussed peri -
odically with the management and-supervisory personnel listed below.
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Mr. W. Wittmer, Acting Maintenance Superintendent'
Mr. R. Grippardi, Quality Assurance Supervisor Mr. S. Jefferson, Assistant to Plant Manager.
i Mr. J. Herron, Operations Supervisor i
Mr. D. Legere, Acting Maintenance Supervisor Mr. R. Pagodin, Technical Services Superintendent Mr. J. Pelletier, Plant Manager Mr. L. Cantrell, Shift Supervisor i
Mr. R. Wanczyk, Operations Superintendent
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Mr. T. Watson, I&C Supervisor
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 80-94% of full rated power were con-ducted weekly to perform routine surveillances of control rod drives, main turbine valves and bypass valves'. Weekly power reductions, which occurred on January 28, February 4, 11, 18, 25, March 4, and 11, included rod pat-tern adjustments to facilitate 100% rated power operations or' confirm the presence of failed fuel in suspect locations.
On February 14, reactor power was briefly reduced-to 98% power in response to a loss of the turbine building closed cooling water system. -A failure of the turbine's mechanical hydraulic control system.on February 24, due to a failure of its electric pressure regulator, resulted in a main cool-ant system pressure spike and transient reactor power increase to 115% of
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rated power. Appropriate plant personnel response to both events pre-
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cluded plant trips.
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During the report period, Vermont Yankee conducted public meetings to pre-sent information and answer questions regarding their proposal to locate'a low level radioact1ve waste (LLRW) facility in Vernon, VT. A non-binding-article was included in the agenda of the Town Meeting to allow town government officials to assess town resident support for State of Vermont legislation that would designate a portion of Vermont Yankee's property in town as the preferred site for th? location of the LLRW facility.
The Town Meeting vote on the article '<on overwhelming support on March 5,.
1990.
An INPO evaluation by an 18 membar team occurred between January 29 and February 9.
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i A SALP Management Meeting was held in the.Vernon, VT town offices with a
Vermont Yankee on January 31. A Memorandum and Order was issued on
January 26 by the Atomic Safety and Licensing Board-.that admitted one con-
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tention by the State' of Vermont (an intervenor) in Vermont; Yankee's re-quest to recapture the construction period.
The contention centered around a maintenance team inspection finding (IR 89-80) regarding-lack' of a formal maintenance and surveillance program plan.
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Effective January 28, Mr.- John b. Macdonald, Resident Inspector (RI)- at -
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VYNPS, was assigned as Senior Resident Inspector at the Pilgrim Nuclear
'i Power Station.
On March 11 Mr. Thomas G. Hiltz assumed the RI position
at VYNPS. Mr. Hiltz was previously a Reactor Engineer in the Division of t
Reactor Projects in the NRC Region I Office (NRC:RI).
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Status of Previous Inspection Findings
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3.1 (Closed) Unresolved Item 87-12-03:
Control Room Habitability Review.
In October 1985, the NRC staff conducted a control room ventilation
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system design review to determine if appropriate design' features were.-
installed as-necessary to satisfy the habitability requirements of NUREG 0737, Item III.D.3.4, " Control Room. Habitability Requirements."
The review determined the control room. ventilation system at VYNPS t
was adequate, as documented in Inspection Report (IR).85-36.
How-ever, the report noted several. areas which reqJired,further review including issuance of technical specifications that demonstrate:
control room isolation on toxic gas and radiological-events; control room envelope inleakage less than 20 SCFM; and, acceptable equipment temperature limits.
On August 11, 1986, the NRC issued Amendment 96 to the VYNPS operat-ing license which addressed the operability requirements of the Toxic Gas Monitoring (TGM) system, and~which identified that the TGM will automatically isolate the control room during a toxic gas event.
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control room ventilation system must be manually isolated, if re-
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quired, following a radiological event.
Licensee off-normal proce-dure, ON-3153, Excessive Radiation Levels, requires the control room ventilation system be placed in emergency circulation mode if the source of abnormally high radiation levels cannot be readily deter-mined. Automatic isolation of the control room ventilation system
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was determined not to be required for analyzed design bases events,
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as long as the control room inleakage does not exceed the safety
analyses assumed limit of 20 SCFM.
In order to ensure the 20'SCFM inleakage assumption is maintained, the control room kitchen and sani-tary facility fans and dampers were provided automatic isolation functions when the control room ventilation system was automatically or-manually actuated. Additionally, the design. bases and historical
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performance data of the control room ventilation system indicate the system will maintain control room temperatures below 78 degrees.F
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(drybulb, 50% RH), which is the maximum assumed temperature for'the t
control room and service building as stated in Final Safety Analysis-
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Report (FSAR) Section 10.12.
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a The licensee response to-the NRC areas of concern regarding control room habitability have been appropriate. Any further questions which
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may be. posed regarding control room habitability will be addressed-t separately. This item is closed.
3.2 (Closed) Violation 89-08-05:
Failure to Conduct a Proper Package
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Search. On August 11, 1989 a licensee letter to the.NRC=provided the.
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immediate actions and.results of-an investigation in response to'this item. Based upon the information presented, the licensee: concluded that a violation:of the physiccl security plan and implementing pro-cedure had not occurred, and therefore requested withdrawal of the-apparent violation.
In a February 14, 1990 letter to the licensee, the NRC:RI office documented its reassessment of;this item and de -
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termination that the violation should be withdrawn. This item is-closed.
3.3 (Update) Unresolved-Item 89-01-04:
Resolve Design Basis Questions Torus Vacuum Breakers.
This issue reflected the NRC resident staff; and another_ licensee informing this licensee about a. potential-design deficiency with the Reactor Building to torus' vacuum breakers.
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vacuum breaker assembly is comprised of-two valves:
an instrument air operated butterfly valve and a swing check valve.
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are normally closed. However, the butterfly valves (V16-19-11A&B)~
fail open upon loss of instrument air leaving the check valve as a single barrier for maintaining primary containment integrity..Upon identification of the inability of the butterfly valves to provide
" fail-safe" containment isolation, the licensee declared the' butter-fly-valves inoperable and instituted Technical' Specification (TS)
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Action Statement 3.7,0.2 to verify closure of.the check valves by
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l logging their positions daily.
Continued operation in this condition
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is allowed by the TS. The licensee believed that the butterfly valves are primarily installed to support-the: vacuum breaker func-tion. Their secondary function is to provide a containment isola-tion. Therefore, the licensee concluded that the open failure posi-
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tion for these valves is correct, in that, it. supports their. intended primary function in providing vacuum relief if required. Once the licensee fulfilled the TS requirements, the remaining issues.became
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those of design basis sufficiency and the need to implement system modifications.
On February 16, 1989 in its letter (BVY 89-17) response to NRC Generic Letter 88-14, " Instrument Air Supply System Problems Affect-ing Safety-Related Equipment," Verment Yankee stated that they per-formed an investigation into the requirements of the Reactor Building to torus vacuum breakers. They concluded that the current operation and design is in compliance with the original design basis and
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Vermont Yankee's current licensing basis.
However, Vermont Yankee is
pursuing this issue further from a generic basis, through 1.he BWR1
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Owner's Group, and as it applies to other BWRs which are similar in; design-to Vermont Yankee.
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Subsequently, on February 5,- 1990, the NRC issued Amendment No.119'
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to the facil.ity operating license which clarified the operability requirements of the Reactor Building to Torus' vacuum breaker' system..
Specifically, wording was. removed-that required locking closed an-A inoperable vacuum breaker and wording was added that power operation
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may continue with a failed open vacuum breaker provided the other
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vacuum breaker in that line is verified to be closed and its. position
is logged daily.
Following the issuance of Amendment No. 119, the
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licensee documented.their' completed review of-the design bases and.
operability status of the: butterfly valves in their letter to;the NRC (BVY 90-016) dated February-15, 1990. The licensee concluded.that in the case where the butterfly valves fail open and would therefore not.
serve a containment isolation function, the check valves.would pro-vide this containment isolation function. They indicated this situ '
ation is acceptable as their configuration ~ allows both safety func-tions of this line to be met, using a highly reliable, passive check-valve to accomplish-the isolation function.
The licensee believes
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this was the intent of the original designer and. the reason for.the-special consideration given this penetration.
Therefore, these-
valves are in compliance with the current bases..
Based upon a February 14, 1990 review of the above noted information by the-Plant Operations Review Committee (PORC), the committee recom-mended to the Plant Manager that the subject valves-be declared oper-
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able.
The NRC: RI and NRC:NRR were informed of the:11censee's inten -
tions to declare the valves operable. -The licensee declared the valves operable on February 15, 1990.
.s The NRC:RI continues to believe that.the dual safety function of the Reactor Building to torus vacuum breaker system at VYNPS, and its susceptibility to loss of instrument air, warrants further NRC review to fully resolve all design, licensing, and operability questions.
This item remains open.
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:
i Control Room Cable Vault I
Reactor Building Fence Line (Protected Area)
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Diesel Generator Rooms Intake Structure Vital Switchgear Room Turbine Building
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Control room instruments were observed for correlation between chan-
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tions. Alarm conditions in effect and alarms received in the control
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room were-reviewed and discussed with the operators ;0perator aware--
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ness and response to these conditions were reviewed.. - Operators were found cognizant of board and plant conditions.
Control. room and-
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shift manning were compared with' technical specificationLrequire-ments.
Posting and control of radiation, contaminated and high radi-ation areas were inspected. Use of and compliance with radiation work permits and use of required personnel monitoring devices'werez 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 station proce '
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dures, 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-
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potential reportable occurrence reports.
Inspections'of the control room were performed on weekends and -backshifts including January 23-26, February 7, 8, 13, 15, 17,-23, and 26; and March 3'and.8,.
1990.
" Deep backshift" inspections were conducted as..follows:.
Date Time
01/23/90 3:00 a.m. - 5:00 a.m.
02/23/90 10:00 p.m. - 11:00 p.m.
Operators and shift supervisors were alert, attentive and responded appropriately to annunciators and plant conditions.
4.2 Safety System Review
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Portions of the Emergency Diesel Generator, Reactor Core Isolation-
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Cooling, Core Spray, Residual Heat Removal, Standby. Gas' Treatment, Residual Heat Removal Service Water, safety related electrical, and High Pressure Coolant injection systems were reviewed to verify pro-per alignment and operational status in the standoy mode. _The_ review-included verification that: (i) accessible major flow path valves were correctly positioned, (ii) power supplies were energized,
(iii) lubrication and component cooling was proper, and_(iv) com-ponents were operable based on a visual inspection of. equipment for
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leakage and general conditions.
No violations or safety concerns were identified.
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4.3 Inoperable Equipment Actions taken by plant personnel during periods when equipment was
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inoperable were reviewed to verify:
technical specification limits were met, alternate surveillance testing was completed satisfactorily and, equipment was returned to service only after repairs were com-pleted properly.
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-Date Out Date In System 01/29/90 01/29/90
"A & "B" Toxic. Gas Monitor (TGM)
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01/30/90-01/30/90
"A" EDG 02/13/90 02/18/90 RCIC 02/22/90 02/22/90
"A" & "B" TGM 03/01/90-03/01/90
"A" SBGT= System 03/09/90 ROOS *
Diesel Fire Pump
- The diesel fire pump remained out-of-service at'the conclusion of the inspection period pending the disposition of Non-Conformance.Re-
-port-(NCR) No.-'90-04.
This NCR documented the installation of parts without receipt. inspection during a. vendor overhaul of the pump. Thel parts used for the pump _ overhaul are required to pass _ proper strin-gent QC and procurement standards through. a receipt inspection.
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
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was created, safety evaluations were prepared in accordance'with 10-CFR 50.59, if required, and requests for modifications were reviewed-and approved prior to installation.
Implementation of.the requests was reviewed on a sampling basis.
90-03 -- Implemented on February 2 to document the opening of_the-supply breaker for the DC motor on the AC vital motor generator set.
The circumstances surrounding-the necessity for the licensee to take this action were discussed in Section 6.2 of IR 89-22. -The'licen-
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see's Temporary Modification documents'the normal inservice condition-
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of the DC supply breaker normal _ inservice condition.as "Open".
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development of the temporary modification, its associated safety i
evaluation, and their subsequent review by the.PORC demonstrated pro-per licensee judgement on resolving equipment deficiency issues asso-ciated with continued plant operation.
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Additionally, several temporary modifications were close ! out during i
the report period. These were reviewed for completeness and adequacy of system restoration.
4.5 Review of Switching & Tagging Operations The switching and tagging log was reviewed and ttgging activities q
were inspected to verify plant equipment was controlled in'accordance 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-i cedures, as described in Sections 4.4 and 4.5, was in compliance with procedural instructions and was consistent.with plant safety.
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shift-inspections have consistently ~ found operators to be alert and'
attentive.- Operations are routi_nely conducted in a professional man-t ner in an atmosphere of quiet control and competence. With the ex--
t ception of isolated-instances, overall plant-cleanliness and material condition continue-to be good.
Periodically throughout the inspection period, drywell temperatures
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as obtained from an equipment qualified temperature-indicator (TI:
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16-19-30) slightly exceeded the 160 degrees F entry condition for emergency operating procedure (EOP) OE 3303, Drywel.1 Pressure and-Temperature Control Procedure.
For the temperatures indicated,.-0E-j 3103 only required continued drywell-temperature monitoring. All-drywell temperatures trended have remained belcw existing-equipment
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qualification program analysis.
This condition exists due to drywell
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air handling units (RRU-1.and 4) being inoperable, as detailed in Section 6.1 of IR 89-22. Because the drywell temperatures _have in-
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creased to the point that the upper average temperatures are approach-
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ing average EQ limits, the Operations Department-_ requested that the Engineering Support Department (ESD) conduct a review of the condi-tion.
The ESD generated recommendations associated with lowering ambient reactor building temperatures to increase the heat transfer rate across the reinforced concrete containment.
The recommendations
properly ref.locted minimum values specified in -the FSAR.
The inspec-
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tor took note of the fact that the Reactor Building contained Emer-
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L gency Core Cooling System batteries that are used to supply the Un-
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interrupted Power Supply units that supplies. power to both Low Pres-
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sure Coolant Injection subsystem valves.
Based upon a review of plant procedure OP 4210, Rev. 5, Maintenance.and Surveillance of Lead Acid Storage Batteries, the inspector determined that the minimum temperature criterion for the batteries is 50 degrees F.
Since the
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ESD recommendations reflected the FSAR minimum value of 55 degrees F for occupied areas in the reactor building, the inspector had no con-
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l cerns that the ESD recommendations could negatively impact on the l
subject batteries. A short duration outage in mid-March is planned
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l to restore the unavailable drywell air handling units to service.
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l addition, the PORC has been cognizant of the elevated temperature condition in the drywell. The ESD involvement in reviewing and pro-
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viding recommendations to address operational deficiencies continues-
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to be a notable licensee strength.
Licensee response to elevated i
I drywell temperatures has been appropriate and has demonstrated a pro-l per safety perspective.
No deficiencies were identified in licensee operations associated
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with the reviews covered in Section 4.
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Security 5.1 Observations of Physical Security Selected aspects of plant physical security were reviewed during regular and backshif t hours to verify that controls were -in. accord-ance with the security plan and approved procedures.
This. review included the following security measures:. guard staffing; vital and protected area barrier 1ntegrity; maintenance of isolation zones;:
and, implementation of-access controls, including authorization,.
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badging, escorting, and searches.
No-inadequacies were identified..
5.2 Inaccurate Background Investigations
The licensee notified the inspector on February 7 that they had dis-
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covered irregularities in three updated background investigations-
that were being performed as part of their Fitness-For-Duty Program that is required by 10 CFR 26.
Shortly thereafter, the licensee de-termined that unescorted access had been granted to a number.of in-dividuals who may have had inaccurate background investigations con-ducted. All background investigations in question were performed by a security screening contractor employed by the licensee. ' The licen-
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see's discovery resulted from an investigation' that they. initiated on
February 2, when Vermont Yankee security management personnel de-
-tected an irregularity that was contained ~1n an updated background
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investigation developed by their contractor. Based upon this inves-tigation, a total of nine cases were identified where flawed inves -
tigations were conducted that-resulted in access. authorization being granted.
Upon determination that three of these individuals were currently badged for unescorted site access, their access authoriza-tion was terminated pending a re-evaluation of their background in-
vestigation. Timely notification of this event was made by the -lic-
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ensee to the NRC Operations Center and the inspector.
l On February 7, the licensee directed that the Yankee Nuclear Services t
Division (YNSD) immediately perform an audit of the activities of,the security screening contractor for the purpose of assessing program-
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matic adequacy of the investigation process, evaluating established controls, and verifying compliance with licensee requirements.
This audit was initiated on February 8 and completed on' February 20. -The licensee attributed the cause of this event to -the -failure of a security screening contractor employee to adhere to. established rules and guidelines which would ensure accurate and factual investigation.
Further, no information had been uncovered that would have precluded granting access in any of the identified cases.
In accordance with j
10 CFR 73.71, the licensee issued Security Event Report 90-S01 on March 9, 1990. This report, which was comprehensively written and issued in a timely manner, properly documents the event and imple-mented corrective actions.
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Additionally, the inspector determined that in this area o'f security, I
the licensce's security management personnel were providi.ng good oversight of security contractor activities.
The licensee's inves -
tigation properly pursued the rblevant issues and was. conducted in.a diligent manner.
This event will be further reviewed by NRC:RI Physical Security-
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specialists at a later date.
6.
Events Requiring Telephone Notification to the.NRC The circumstances surrounding events which required NRC notification via-d the dedicated ENS line were reviewed.
A_ summary of the inspector's find-ings follows or is documented elsewhere as noted.
6.1 Improper Authorized Access
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A notification to the NRC in accordance with'10 CFR 73.71(b)(1)~,
Appendix G.I(b) was made at 6:40 p.m. on February 7 that' an-event-involving improperly authorized access of individuals to protected
.4 and vital areas. occurred as a result of. inaccurate background inves-tigations performed by a security screening contractor. This event is discussed further in Section 5.2.
6.2 Actuation of Primary Containment Isolation and Standby Gas Treatment Systems i
At 2:53 p.m. on February 9, the NRC was notified in accordance with.
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10 CFR 50.72(b)(2)(ii) that inadvertent actuations of. the Primary
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Containment Isolation and Standby Gas Treatment Systems occurred at l
2:45 p.m. due to personnel error.
This event is discussed further in
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Section 7.2.
7.
Plant Operations 7.1 Fuel Failure Indications i
Throughout this report period, off gas levels, as sampled at the-l Steam Jet Air Ejector (SJAE) were in the range of 36,000 to 41,000 l
uCi/sec.
Inspection Report 89-09 identified an increase in off gas system radiation levels beginning in June 5,1989. The increased radiation levels were indicative of a small failure (s) of fuel pin cladding.
Fuel failures in the' core cycle appear to have occurred immediately following the rod pattern exchanges from sequence A2 to
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81 on June 3, 1989 and possibly Al to B2 on September 23, 1989.
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According to Vermont Yankee, the exact cause of the fuel failures is
unknown, although the symptoms appear to indicate pellet-clad it.ter-action versus crud induced localized corrosion. Attachment I to-this inspection report provides a graphical representation of off gas
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levels experienced at the SJAE during the current operating cycle.
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The off gas level at the SJAE on March 12, 1990'was 39,300 uCi/sec',-
l and is the result of xenon and krypton fission gas'nuclides. An
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isotopic analysis, allowing for. a 30 minute decay period, indicates' a
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release rate of~14,600 uti/sec.
This value is well belowlthe'applic '
able TS 3.8.K.1 limit of 160,000 uCi/sec. The air dese due.to: noble-gases released in gaseous' effluents, as calculated by the licensee:
based on the methodologies in the Offsite Dose Calculation' Manual,,
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were very small and~ remained considerably less:than the'11mits speci-f fied in TS 3.8.F.1.
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Extensive actions have been -taken by the licensee in managing 'this
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condition.
In late December 1989, the licensee acknowledged that they have a program in place for monitoring fuel performance and'
taking appropriate action. However., no formal' program documents
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existed. On January 26,.1990, plant management and:the PORC approved for issuance the document " Fuel Performance Monitoring Guidelines and Failed Fuel Action Plan." The document reflects the licensee's fuel integrity performance goal ofino fuel' failures.
Licensee management.
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recognizes that to preclude implementation of the failed fuel action
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plan necessitates concentrated efforts in: (1) procuring fuel with high quality cladding', (2) refueling housekeeping, (3) post-refueling >
housekeeping, (4) strict reactor water chemistry control during all modes of-operation, and (5) strict. compliance with fuel vendor _ hand-ling and operating guidelines. The document describes. fuel _ perform-ance monitoring activities. As part of the coolant sampling-and analysis program', the licensee uses the EPRI CHIRON computer code, which analyzes reactor coolant water and predicts the number of'
failed fuel rods in the core.
Currently, the analyses-predicts that-approximately ten failed rods exist in the core.
Based upon the SJAE off gas activity being in the range of'10,000 to
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60,000 uCi/sec, the licensee is implementing' Action Level IV activi-i ties of the plan which includes:
(1) augmented sa'mpling and analysis of off gas and iodine data, (2) ensuring equipment is available for sipping during the next refueling outage, (3) assessing the use of rod pattern changes to reduce off gas-level, (4) preparing for a fuel reconstitution / replacement effort at the next refueling, (5) utiliz-
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ing the shutdown iodine filter during power maneuvers, and (6) locat-
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ing and repairing steam leaks. Action Level V responsibilities, which are implemented when off gas levels are in a range of 60,000.to 80,000 uCi/sec requires the licensee,-in part, to consider the effect of fission product activity on plant maintenance and evaluate the prudence of a power derate or early shutdown. Action Level VI re-
sponsibilities include, in part, requiring the plant operators to initiate a power reduction to maintain off gas levels below 80,000 F
uti/sec.
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Licensee response to'this condition continues to be excellent. ~A good level of involvement by'on-site and off-site organizationsLis-routinely observed in addressing issues associated with the' plant-operating with failed fuel.
This includes assessments by PORC, as-part of their review responsibilities,f to detect potential = safety;
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hazards. The on-site Reactor Engineering and Chemistry Departments _
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effectively communicate the results and analyses performed in accord-
ance with the established program and procedures.
TheLettablishment'
i and implementatian of the licensee's' plan represents good management-and a conservative approach to' minimizing 1the. negative effects on the
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l plant, personnel, and environment due to the operation of the plant:
with failed fuel.
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7.2 Primary Containment Isolation System Actuation At 2:45 p.m. on February 9, with the. reactor operating at 100% of
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rated power, a Primary Containment Isolation System (PCIS) Group-III-actuation occurred.
The actuation, which isolates primary and-secondary containment ventilation and-initiates.the Standby Gas
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Treatment System, was inadvertently caused by Radiation' Protection (RP) technicians. At' the time of occurrence the RP. technicians were performing a quarterly calibration on one. of the Reactor Building Ventilation (RBV) Radiation Monitors.
This calibration is required
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by TS Surveillance Requirement 4.2.C.
Calibration of the RBV Radiation Monitors was-performed in accordance with plant procedure OP-4503, Rev. O, Source Calibration of.. Reactor Building Ventilation'and Refueling Area Zone Radiation Monitors.
This procedure specifies the appropriate actions for. operators;to bypass the' applicable PCIS trip relay associated with the monitor being calibrated, and the removal by the operator of the' trip-bypass condition after ensuring that the monitor.has been returned to nor-mal. On the day in question, the actuation of the PCIS occurred be-cause an RP technician. removed the trip bypass condition prior to the monitor returning to a normal: or untripped condition when its detec-tor was removed from the source calibrator.
The licensee informed ~
the inspector that the cause of this event was the failure of the RP technicians to use the required procedure to perform the assigned
task. However, the surveillance activity was performed with the knowledge and consent of the Shift Supervisor (SS).
Plant equipment response to'PCIS Group III actuation was' normal.
Following the actuation, plant personnel verified that there were no unacceptable radiological conditions, the isolation was reset and systems were returned to normal. The licensee made the required notification to the NRC as part of the event response.
This event will be the subject of a Licensee Event Report (LER No. 90-01).
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The inspector will follow-up the licensee's event. analysis, root cause determination, and development of corrective measures during a.
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subsequent review of the LER.
The inspector had no further_ ques-tions.
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7.3 Loss of Turbine Building Closed Cooling Watel On February 14 at 10:05 a.m., Turbine Building' Closed Cooling Water
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(TBCCW) "A" pump tripped and could not be restarted from the Control-RoomPanel-(CRP).1 Since all CRP indications for the pump were ex-tinguished, the operators immediately informed electrical maintenance
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personnel that they suspected that-the control power fuse for the~
pump was blown.
There are two 100 percent capacity pumps in the
TBCCW' System, however, the "B" pump was out-of-service for mainten-
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ance at the time of this event. The TBCCW System provides cooling
water for the heat removal from'various equipment in the Turbine'
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Building, such as, reactor feed pumps, condensate pumps, isolated phase bus coolers and the station air compressors.
At the time the event occurred, the inspector was in the control room and was able to monitor operator response. An auxiliary operator was
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dispatched to the TBCCW Pump "A" electrical breaker, the station air compressor cooling was switched to the alternate cooling system
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supply (i.e. service water), and control room operators. began to
closely monitor condensate and reactor-feed pump-temperatures. At 10:10 a.m., the SS directed the initiation of a plant load! reduction
at a rate of 1 percent of rated power per minute.
This action was taken to help reduce the heat load on the TBCCW system;until normal-flow could be restored.
The TBCCW Pump-"A" control power fuse was.
replaced by maintenance personnel and the pump started by 10:12 a.m.
The power reduction was terminated at 98 percent of rated power at 10:13 a.m. after no abnormal temperature indications on equipment was verified.
The plant was returned to full power operation at 10:30 a.m.
The maximum equipment temperature increase observed was a 6 degree F on one of the reactor feed pumps.
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Just prior to the trip of the "A" pump, an I&C technician was per-forming an instrument calibration on a pressure switch at the common discharge piping downstream of the two TBCCW pumps. The work was being performed in pressure switch PSL-104-46-1B in accordance with plant procedure OP-5355, Cooling Water Systems.
It was noted by the
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I&C technician that at the time a voltage check was being made across pressure switch contacts, tne "A" pump tripped.
The licensee is con-ducting an investigation into the pressure switch calibration process to ascertain if it is the root cause of the blown control circuit-fuse.
Further, the licensee is in the process of developing a Plant Information Report to document the event, the immediate corrective actions, and appropriate corrective actions to preclude recurrenc v
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Control room personnel and plant support staff response to this. event I
was excellent.
Supervisory control room personnel demonstrated pro-per command and control and operators were diligent-in_ monitoring-equipment condition. Good communication between departments was evi-1 dent and procedural instructions pertaining to this upset condition
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were consulted and adhered to'.
The Shift _ Engineer was involved and properly integrated into the process.
Licensee follow-up action-appears to be appropriate and timely.
The inspector had no concerns regarding this event.
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8.
Maintenance / Surveillance 8.1 Emergency Diesel Generator (EDG) Operational Readiness Demonstration The licensee conducted a routine monthly surveillance test'on the "A"
EDG on January 30,-1990.
This test was performed in accordance with plant procedure OP-4126,-Rev. 23, Diesel Generators; Surveillance,.
which conforms'to the requirements of TS 4.10.A.1.a.
Specifically, the diesel is manually started and loaded-to between 2500 and 2750 kw for an eight hour period. According to the procedure'OP-4126, cylin-der-exhaust temperature data is recorded on an hourly basis on Form VYOPF 4126.02.- The inspector noted that after four hours into the EDG run, the No. 11 cylinder was recorded as indicating'1200 degrees F and the No.12 cylinder was recorded as indicating 255 degrees F.
Additionally, the indication for the No. 5 exhaust temperature was-noted by the plant operator to have wide swings in.the range of 900
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degrees F.
The acceptance criteria on the form VYOPF-4126.02 stated that:
" Variation between cylinders should not exceed-300 degrees F" and " Maximum temperature of any cylinder is 1100 degrees F."=
The other cylinder readings were indicating.an exhaust temperature in the-range of 710 to 970 degrees F.
The completed Form 4126.02 for January 30, 1990 indicated that Maintenance Request (MR)_90-016_was issued to identify the thermocouple (TC) instrumentation discrepan-cies for Cylinder Nos. 11 and 12.
This MR was -issued.on January 3,.
1990. MR No. 90-0311 was issued on January 30, 1990 to' resolve the spurious readings provided by Cylinder No. 5 TC instrumentation.
Following discussions with personnel in the I&C and Operations De-partments, and after reviewing maintenance records, the inspector-learned that there has been indications of poor operational and main-tenance history on both EDG's exhaust TC. instrumentation. The prin-cipal contributor to anomalous instrumentation performance appears to be the existing design configuration that incorporates an interme-diate terminal box utilized by the TC wiring that'is susceptible to engine oil exposure. On February 23, 1990, the I&C Department issued MRs 90-0506 and 90-0507 to replace the TCs on "A" and "B" EDGs, re-spectively such that the intermediate terminal box 'is bypassed by the use of TCs with longer sheathed wiring.
Based upon an Engineering Support Department analysis conducted on February 16, 1990, the
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thermocouple reconfiguration and replacement can be considered a one-
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for one replacement and: performed-under the jurisdiction of a MR.-
While these MRs are waiting availability of parts from the EDG ven-dor, corrective maintenance was; performed for MRs 90-0016'and 90-0311-,
The February 28, 1990, monthly surveillance test performed on the "A"-
EDG demonstrated that the TCs were performing satisfactorily.,Not-withstanding the instrumentation problem due~to the. terminal box and engine oil exposure problem, there appears to~be other TC failures-that cannot be explained by the aforementioned problem and may be indicative of a longstanding reliability. issue.
The inspector questioned the._ licensee-about how discrepancies associ-ated with. failure to meet the surveillance-testing acceptance'cri-teria for the cylinder exhaust temperatures are resolved. 'According to representatives of the Operations Department ~, the surveillance test results are reviewed by the on-duty Shift-Engineer.
Following a review of these results,-the engine performance is analyzed =and a.
disposition as to what corrective measure are-warranted is developed.
Given-the operational history, the EDG start and load information,t and engine performance data, the subject' discrepancy.was.-judged to be instrumentation failure. On a monthly basis, the Operations Depart-ment issues a Plant Performance Monitoring Memorandum, which. dis-
cusses plant performance trends 'and shows trending of operational--
j data.
For both EDGs, cylinder tempentures are-trended over the -
prior eight month period. Where instrumentation discrepancies
exists, MR numbers are assigned.
l The licensee was responsive to inspector questions and~provided needed documentation and technical assistance in a timely. manner.
No deficiencies were identified during the-inspector's review-of the licensee demonstration of EDG operational readiness.
8.2 EDG Maintenance i
As a result of a licensee initiated evaluation. program to determine the acceptability of extending the vendor recommended inspection and l
overhaul frequency from 12-18 months to-22-24 months.for the "B" EDG, j
Vermont Yankee identified the need to establish an accelerated verti-
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cal drive assembly inspection frequency on this EDG, 'The NRC's.re-l view of this item is documented in Section 7.3.of-Inspection Report-l 89-21.
J Based upon a November 1-3, 1989 inspection, the licensee determined d
that excessive clearance tolerances in the upper vertical drive i
assembly were observed that were indicative of potential bearing wear
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or spacer wear between the thrust and roller bearings. At that time, the licensee issued MR 89-3389 to control the maintenance activities that restored all clearances in the vertical drive assembly com-ponents to design tolerances and established the need to conduct an investigation into the root cause of the condition that was causing
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wear on the drive components.
Following the establishment of an in--
spection schedule that would continue until the April 1990 planned
periodic overhaul, the EDG was placed in service. The first periodic -
inspection occurred after 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> of EDG-operation and was. conducted-in December 1989. 'All component tolerances remained constant at de-sign values.
The next scheduled-inspection was to follow an addi--
tional accumulated operation period ofL16 hours and was to occur in
the latter part of February 1990.
However, the~ scheduling of an
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annual inspection.of EDG air receiver tanks on January 31, 1990-l afforded the licensee an opportunity-to inspect the vertical drive.
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assembly after an additional 8 gerational hours.
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The January 31, 1990 inspection determined thatfexcessive clearance--
tolerances in the upper vertical drive-assembly had reoccurred.
The Maintenance Department-recommended that the EDG remain out-of-service;
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to disassemble, inspect, and rebuild the upper vertical drive assembly.
The inspection; showed that the excessive clearances were theiresult-
of component wear in the stacked bearing assembly.-.This,was attri-
buted to a loss of'a designed interference fit between the pinion shaft and thrust bearing inner race that allowed turning of the thrust bearing inner race, spacer ring and spacer relative to adja-
cent components and one another.
The loss'of the interference fit, that would allow for the turning of the thrust bearing's inner race, could result from the following factors:
(1) torsional stresses due to load changes on the drive shaft; (2) inadequate initial inter-ference fit during initial manufacturing; and (3)' inadequate com-pression on the stacked bearing assembly. The licensee indicated
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that their reassembly of the upper vertical drive unit addressed the-
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latter two issues by careful dimensional checks-and hot assembly techniques. The entire maintenance process on this EDG was witnessed:
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and supervised by a vendor technical representative, who ensured that the reassembly followed current manufacturing practices. Apparently,-
when this EDG was originally assembled in 1968, a mixture of prac-tices were in use, such that, potential causal factors addressed.
above could well have existed in the original manufacturing ' process.
Based on the previous discussion, the licensee requested, on February 12, 1990, the assistance of the EDG vendor (Fairbanks Morse) in i
assessing the ability of the EDG to have' performed its design func-tion. Additionally, the vendor was provided with the licensee's. root cause analysis for review and comment.. Testing of the "B" EDG and
inspection of the clearances on the vertical drive assembly were con-t ducted prior to the EDG being declared operational on February 3, 1990.
The licensee's response to this unexpected activity was good and re-flected the conservative aggressive manner-in which equipment defi-ciencies are pursued. A good level of QA oversight and involvement by a vendor technical representative were noted. The staff personnel
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supervising and performing the work were very knowledgeable and skilled.
Housekeeping practices and conditions at the worksite re-flected strong licensee performance.
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During the inspector's review of maintenance activities associated
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with the EDGs, the inspector determined that the Maintenance Depart-ment was in the process of. proposing changes to the existing EDG overhaul program'- Currently, the preventive maintenance program for
the EDGs consist of major electrical. and mechanical overhauls and =
inspections at a periodicity.of 12-to.18 months. The overhauls'are i
normally scheduled concurrent with a refueling outage although-they
have been conducted in the past with the plant at power' utilizing thei-t 7 day outage time allowed for the Limiting Condition for Operator (LCO)-specified by TS 3.10.B.1.
The-Maintenance Department is re-
directing the program in' two significant ways. 'The first deals with expanding the program to include more predictive maintenance.prac-tices in addition.to the' existing preventive maintenance inspections.
The second aspect includes extending the periodicity.of. major over
hauls to every 22 to 26 months which would require removal of.the
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EDGs from. service, one at a-time, during' full power operation.;
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q This~1atter aspect was potentially-of some concern:to the inspector..
The inspector informed the licensee that a detailed safety basis
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needs to be developed that would demonstrate _ clear safety benefits that would be derived from back-to-back voluntary entry into the LCO-that allows removal of these important. safety related pieces of
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equipment from service during power operations. The' licensee
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acknowledged the inspector's comments.and concerns, and indicated
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that they were in the process of developing a written safety. basis
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that would be presented to the PORC for its review. <The-licensee's
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plans for changing their EDG overhaul program is-identified as an-
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unresolved item (50-271/90-01-01).
9.
Review of Licensee-Response to NRC Initiatives l
9.1 Review of NUREG-0737 Commitments i
The NRC's Region I office has inspection responsibility for selected action plan items.
These items have been broken down into numbered descriptions (Enclosure 1 to NUREG-0737, " Clarification of-TMI Action Plan Items").
Licensee. letters containing commitments to-the NRC were used as the basis for acceptability, along with the NRC
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l clarification letters and inspector judgment. The following item was
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reviewed.
- TMI Task Action Plan Item II.K 3.18.C, ADS-Modification.
The
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inspector reviewed the circumstances and supporting documenta-tion and verified that this item was completed-by VYNPS.
In-spection details pertaining to the installation of the modified ADS design are contained-in Section 3.2 of. Inspection Report 87-16. TS Table 3.2.1 was revised to reflect this modification to the ADS system, as was the FSAR.
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The inspector reviewed plant procedures OP-2112, Rev. 9, " Auto Blow-down System," and OP-4343, Rev.- 16, ADS System Logic Test, to verify
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that the modification had been-incorporated into station procedures.
and its logic system is tested to verify operability.
No discrepan-cies were identified. This item is closed.
9.2 Installation of a Hardened Wetwell Vent - Generic Letter 89-16 In response to the NRC's: issuance of Generic Letter 89-16:
Instal-
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lation of a Hardened Wetwell Vent, the licensee specified in-an-
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October 30, 1989-letter to the NRC their voluntary commitment to in-stall a hardened vent.
The licensee indicated that they expect to establish specific design crit'eria such that they can install.this enhanced containment overpressure protection capability by the end of.
s the 1992 refueling outage.
The NRC in its January 19, 1990 letter, accepted the licensee's proposed schedule and their commitment'to make the requisite modifications'under the provisions of 10 CFR 50.59, 10.
Review of Periodic and Special Repory Upon receipt, the inspector reviewed periodic and special reports sub-
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mitted pursuant to Technical' Specifications-(TS).
This. review vertfied, 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 spe.cification; and
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(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 January and
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February 1990 (TS 6.7. A.3).
Semi-annual Effluent and Waste Disposal Report for Third and Fourth
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l Quarters of 1989 (TS 6.7.C.1).
l 11. Organization and Administration
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During the inspection period, the inspector reviewed changes to the 11cen-
see's staff or organization structure as described below. The review in-cluded verification that the licensee's on-site organization structure is as described in the facility TS and verification that personnel qualifica-
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tion levels are in conformance with ANSI N18.1-1971, as described in TS Section 6.1.
The licensee announced their plans for a reorganization that adjusts the onsite organization structure for the VYNPS.
This action is the result of a licensee evaluation that indicated a more efficient and streamlined structure would enable the station staff to function more effectively.
The adjustment consists of the Operations, Maintenance, and Instrument and I
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Engineering Support departments-reporting to the Technical Services Super-intendent.
The licensee,.on March 2, 1990,. submitted a TS Change Request which proposes an administrative = update that deletes the organization charts from the TS and incorporates them into the Yankee: Operational Qual-ity-Assurance Manual. This action is consistent with Generic Letter 88-06 and reflects the approval granted.by the NRC:NRR to other-licensees for similar requests.
The intended changes were reviewed on a preliminary basis by NRC:RI and NRC:NRR personnel. As a result of this review, the NRC staff granted oral approval to the licensee to implement their planned reorganization since.
no deficiencies were identified and a Technical Specification Change Re-quest is in process. The licensee will implement the planned reorganiza-tion on March 15, 1990.
There were no unacceptable conditions identified.
12. Management Meetings The inspectors attended a meeting at VYNPS on. January 31, 1990 between the NRC and the Vermont Yankee Nuclear Power Corporation to discuss the re-sults of the NRC Systematic Assessment of Licensee Performance (SALP) re-view cnnducted for the. period of July 1,_1988 through S9pt. ember _30,1989.
This assessment is documented in SALP Report 50-271/88-99.
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 identi-fied as being included-in the report.
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