IR 05000271/1989012
| ML20248A392 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 09/25/1989 |
| From: | Blough A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20248A391 | List: |
| References | |
| 50-271-89-12, NUDOCS 8910020202 | |
| Download: ML20248A392 (17) | |
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U.S'. NUCLEAR REGULATORY COMMISSION l
REGION I
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Report No.
50-271/89-12-
- Docket No.
50-271 License No. DPR-28 Licensee:
~ Vermont-Yankee Nuclear. Power Corporation-
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RD 5, Box 169 Brattleboro, Vermont 05301
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Facility:
Vermont Yankee Nuclear Power Station Inspection At: Vernon, Vermont Inspection Conducted: July 18 - September 5, 1989
~ Inspectors:
Geoffrey E. Grant, Senior Resident Inspector
' John B. Macdonald, Resident Inspector Herbert J. Kaplan, Senior Reactor Engineer Approved by:
N Y-U~~h
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A. Randy Blough,gCtief, Reactor Projects Section 3A Date l
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Inspection Summary: Inspection on July 18 - September 5, 1989
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(Report No. 50-271/89-12)
Areas Inspected:
Routine inspection on daytime and backshifts by two resident i
inspectors of: actions on previous inspection findings; operationt.1 safety;
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security; plant operations; maintenance and surveillance; engineering support;
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licensee event reports; licensee response to NRC initiatives; and, periodic reports.
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Results:
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t 1.
General Conclusions on Adequacy, Strength or Weakness in Licensee Programs
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The licensee decision to replace the uninterruptible power supply (UPS) is
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a positive indication of a comprehensive approach to overall plant reli-i ability and safety, and demonstrates continuing conservatism in plant
't operations (Section 8.2).
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The licensee review of industry experience which resulted in the identi-
fication of a design discrepancy in the reactor building closed cooling
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water (RBCCW) system attests to an overall questioning approach to plant
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operations and safety.
Licensee safety committee involvement in this pro-cess provided a sound technical and safety basis review of the discrepancy and corrective actions (Section 8.4).
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2.
Unresolved Items One unresolved item was identified during this inspection period:
The NRC will review the licensee's final corrective actions address-
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ing an RBCCW system design discrepancy (Section 8.4).
8910020202 890925 PDR ADOCK 03000271 Q
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TABLE OF CONTENTS PAGE 1.
Persons Contacted............................
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I 2.
Summary of Facility Activities.......................................
3.
Status of Previous Inspection Findings (IP 92701,92702*)............
3.1 (Closed) Unresolved Item 88-14-06: Temporary Modification Program Deficiencies.4
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3.2 (Closed) Unresolved Item 86-14-01: Modify CILRT Procedure to Address Type B and C Leakage Results..........................
3.3 (Closed) Unresolved Item 88-20-03: Corrective Actions for Failed MOV Motor Housing Bo1ts.......................................
3.4 (Closed) Licensee Identified Item 88-20-01: Degraded Vital Fire Barrier at Main Steam Line Reactor Building Penetration.......
3.5 (Closed) Unresolved Item 88-19-04: Improvements to LER Development Process...........................................
4.
Operational Safety (IP 71707,71710).................................
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4.1 Plant Operations Review.........................................
4.2 Safety System Review............................................
4.3 I n o pe ra bl e Eq u i pme nt............................................
4.4 Revi ew of Tempora ry Modi fi cati on s...............................
4.5 Review of Switching and Tagging Operations......................
4.6 Ope rati onal Sa f ety Fi ndi ng s.....................................
5.
Security (IP 71707)..................................................
5.1 Observations of Physical Security...............................
6.
Plant Operations (IP 71707,93702)...................................
6.1 El e vated D rywel l Tempe rature s...................................
7.
Maintenance / Surveillance (IP 71710,61726,62703)....................
7.1 RCIC Inoperability..............................................
8.
Engineering and Technical Support (71707, 35502)....................
8.1 Spent Fuel Racks.............
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8.2 Uninterruptible Power Supp1y....................................
8.3 10 CFR Part 21 Report...........................................
8.4 System Configuration Discrepancy -
RBCCW.................
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Table of Contents-l
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I PAGE 9.
Licensee Event Reporting ( LER) (IP 93702)............................
9.1 LER 89-14.......................................................
9.2 LER 89-09.......................................................
10. Review of Licensee Response to NRC Initiatives (IP 35502)............
10.1 IEB 80-06.......................................................
11. Review of Periodic and Special Reports (IP 71707)....................
12. Management Meetings (IP 30703).......................................
- The NRC Inspection Manual inspection procedure (IP) that was used as inspec-tion guidance is listed for each applicable report section.
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DETAILS 1.
Persons Contacted Interviews and discussions were conducted with members of the licensee staff and management during the report period to obtain information per-tinent to the areas inspected.
Inspection findings were discussed peri-odically with the management and supervisory personnel listed below.
Mr. P. Donnelly, Maintenance Superintendent
- Mr. R. Grippardi, Quality Assurance Supervisor Mr. S. Jefferson, Assistant to Plant Manager Mr. J. Herron, Operations Supervisor Mr. R. Lopriore, Maintenance Supervisor Mr. R. Pagodin, Technical Services Superintendent
- Mr. J. Pelletier, Plant Manager Mr. R. Wanczyk, Operations Superintendent Mr. T. Watson, I&C Supervisor
- Attendees at post-inspection exit meeting.
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2.
Summary of lacility 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 surveillance on control rod drives, main tur-bine and bypass valves. On July 29, power was reduced to 65% to accomp-lish a rod pattern exchange.
Full power was achieved on July 31. On July 30, failure of drywell air handling unit, RRU-1, caused drywell tempera-tures to rise above 160 F, which is the entry condition for the drywell temperature control emergency operating procedure (see section 6.1).
The plant remained in this condition through the coriciusion of the inspection period.
The licensee notified the NRC in accordance with 10 CFR 50.72 on July 25 when one of two reactor building floor drain sump isolation valves (LRW-83) failed to close. Notifications were also made on August 7 for the temporary loss of a civil defense siren during an electrical storm, on August 9 for reactor core isolation cooling (RCIC) system inoperability (see section 7.1), and on August 31 following the discovery of a reactor building closed cooling water (RBCCW) design deficiency (see section 8.4).
3.
Status of Previous Inspection Findings
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3.1 (Closed) Unresolved Item 88-14-06: Temporary Modification Program Deficiencies.
This unresolved item covered a general programmatic deficiency in the licensee's control of temporary modifications. The deficiency tas multi-faceted and was best characterized as a general lack of a cohesive pregram including: insufficient plant operations
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frediew committee (PORC) review, lack-of approved procedures for in -
ista11ation-and removal of modifications, and inadequate control of e
temporary scaffolding and shielding.
In response to.these issues the licensee completely revised AP 0020, " Control of Temporary Modifi-cations," (originally " Temporary Electrical Jumpers, Lifted Leads /
' Mechanical Bypasses") and AP 0019, " Control of Temporary Materials" (originally " Control of Temporary Loads on Piping, Equipment and-Structures"). These revisions were comprehensive and adequately ad-dressed each of the inspector concerns related to control.of tempor-ary modifications. Both procedures represent a marked improvement in the licensee approach.to maintenance of the plant design' basis. To
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ensure effective implementation of the revised program, site-wide
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training was provided on these procedures.
.Although. execution of the reviscd program appears to be effective, licensee development and implementation were slow. This item was identified by the-inspector and discussed with the licensee in Sep-tember 1988.
Implementation of a revised temporary modification pro-
. gram occurred in May 1989. The delay was due to. slow program de-velopment.and a licensee decision to await the end of the refueling outage (April 1989). The licensae invoked adequate interim compen-satory measures to ensure adherence to the intent of the new program.
The delay in implementation of the' revised temporary material h hielding and scaffolding) control program appeared to be due to a slow development process. This item is closed.
3.2- (Closed) Unresolved-Item 86-14-01: Modify CILRT Procedure to Address Type B and C Leakage Results.
This issue will be addressed during the NRC review the licensee's 1989 containment integrated leakrate
~ test (CILRT)'ro ults report and is being tracked under unresolved item 89-02-01. This item is closed.
~3.3 (Closed) Unresolved Item 88-20-03: Corrective Actions for Failed MOV Motur Housing Bolts. This item addressed failures of safety-related motor operated valves (MOVs) when the associated motor housing end bell bolts sheared off during valve stroking.
Immediate licensee corrective actions included tightening these bolts on all accessible safety-related MOVs. The licensee initiated a 10 CFR 21 report de-tailing the generic aspects of this issue in February 1989.
Long-term corrective actions included procedure revision, bolt replace-ments and failure mechanism analysis. The licensee revised OP 5220,
"Limitorque Operator Inspection," to include guidance on motor end bell bolt tightening. A revision to the preventive maintenance pro-gram included a semi-annual check of bolt tightness on all accessible susceptible MOVs. The existing balts on the majority of the sus-ceptible MOVs were replaced with bolts of high strength material.
The remaining MOVs will undergo bolt replacement during the 1990 re-fueling outage. The licensee is continuing research into the ulti-mate failure mechanism which caused the bolts to shear.
Licensee actions to date have been sufficient to prevent recurrence of this
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failure mode.. Closeout of this Lissue, pending final results of the failure mechanism'research, is addressed in the licensee's internal commitment. tracking system and will be reviewed by the inspector during routine inspection. activities This item is closed.
3.4 (Closed) Licensee Identified Item 88-20-01: Degraded Vital Fire Barrier at Main Steam Lice Reactor Building penetrations. -This item concerned the 1ack of. fire rating for the main steam line penetration from the reactor to turbine buildings. By letter dated February 2,-
1989, the licensee requested an exemption to 10 CFR 50, Appendix R-requirements <regarding this penetration. By letter dated June 26,-
1989, NRC:NRR granted the exemption request. This item is closed.
3.5 (Closed) Unresolved Item 88-19-04: Improvements to LER Development'
Process.
This item addressed a trend in licensee event report (LER)
development resulting in the licensee not meeting the 30 day limit for LER submission. The licensee revised internal routing and review practices to improve the development process. Specific responsibili-
. ties were assigned to senior engineering support-department person-nel. 'The effectiveness of the' revised system is being monitored by
the licensee's ccmmitment tracking system. The. inspector observed that LERs developed since this item was opened have met the 30 day limit. This item is closed.
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Operational Safety 4.1 Pla +. Operations Review The inspector observed plant operations during regular and backshift tours of the following areas:
Control Room Cable Vault Reactor Building Fence Line (Protected Area)
Diesel Generator Rooms Intake Structure Vital Switchgear Room Turbine Building Control room instruments were observed for correlation between chan-nels, proper functioning, and conformance with techaical specifica-tions. Alarm conditions in effect and alarms received in the control room were reviewed and discussed with the operators. Operator aware-ness and response to these conditions were reviewed. Operators were found cognizant of board and plant conditions.
Control room and shift manning were compared with technical specification require-
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l 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 were checked.
Plant housekeeping controls were observed including control of flammable and other hazardous materials. During plant tours, logs and records were reviewed to ensure compliance with station proce-dures, to determine if entries were correctly made, and to verify
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correct communication of equipment status. These records included'
L various operating logs, turnover sheets, tagout and_ jumper' logs,> and potentir.1 reportable occurrence reports.
Inspections'of the control room were performed on weekends and backshifts including July 19,.
'ugust 15,16,17,22,23,24, and 31, 1989.
" Deep backshift" inspections w e conducted as follows:-
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Time July 19 9:00 p.m. - 11:00 p.m.
August 31 9:00 p.m. - 11:30.p.m.
Operators-and' shift supervisors were found to be alert and attentive and responded appropriately to annunciators and plant conditions.
4.2 Safety System Review The emergency diesel generators, reactor core isolation 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. proper alignment and operational status in the standby mode. The review included verification that (1) accessible. major flow path valves were cor-rectly positioned, (ii) power supplies were energized, (iii) lubri-cation and component cooling was proper, and (iv) components were operable.-based on a visual inspection of equipment for leakage and general conditions. No violations or safety concerns were identi-fied.
4.3 Inoperable Equipment
' Actions taken by plant personnel during periods when equipment was inoperable were reviewed to verify that: technical specification limits were met; alternate surveillance testing was completed satis-factorily; and, equipment return to service upon completion of re-pairs was proper. This review was completed for the following items:
Date Out Date In System 7/30 ROOS *
Drywell HVAC - RRV-1 8/9 8/10 RCIC'
9/3 ROOS *
"A" UPS (* remained out of servi. at the conclusion of the inspection period)
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i 4.4 Review of Tempurary Modifications
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Temporary modifications were reviewed to verify that controls: estab--
lished by AP 0020 'were met, no. conflicts with technical specifica-
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E tions were created, safety evaluations were prepared in accordance with 10 CFR 50.59 if required, and requests were reviewed and ap-proved prior to installation.- Implementation of the_ requests was_
reviewed on a sampling basis. The following requests were reviewed:
89-044 -- Implemented on August 3, to bypass the RCIC-15 open limit-switch and fully backseat the' valve to isolate potential packing-steam leakage. The TM was restored on August 29 when RCIC-15 was re-moved from its backseat (see section 6.1).89-046 -- Implemented on August 8, to bypass the HPCI-15 open limit switch and fully backseat the valve to isolate potential packing steam leakage. The TM remained active through the conclusion of the inspection period (see section 6.1).89-047 -- Implemented on August 8, to bypass the RCIC-16.open limit
- witch and fully backseat the valve to isolate potential packing steam leakage. The TM remained active through the conclusion of the-inspection period (see-section 6.1).
Additionally, several temporary modifications were closed out during the' report. period. These were reviewed for completeness and adequacy of system restoration. No inadequacies were identified.
.4.5 Review of Switching and Tagging Operations The switching and tagging log was reviewed and tagging activities were inspected.to verify that plant equipment was controlled in ac-cordance with the requirements of AP 0140, Vermont Local Control Switching Rules.
Switching and tagging orders were reviewed for the majority of the inoperable equipment specified in section 4.3.
No inadequacies were noted.
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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 attentive.
Operations are routinely conducted in a professional marner in an atmosphere of quiet control and competence. With the exception of isolated instances, overall plant cleanliness and mate-rial condition continue to be good. No deficiencies were identified in licensee operations associated with the reviews covered in Section 4.
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Security 5.1 Observations of physical Security i
Selected aspects of plant physical security were reviewed during regular and backshift hours to verify that controls were in accord-ance with the security plan and approved procedures. This review included the following security measures:
guard staffing; vital and protected area barrier integrity; maintenance of isolation zones; and, implementation of access controls, including authorization, badging, escorting, and searches. No inadequacies were identified.
6.
Plant Operations 6.1 Elevated Drywell Temperatures On July 30, reactor recirculation unit, RRU-1, one of four drywell air handling unit coolers, was secured following indications that the RRU fan drive belt had failed.
Indications which led the operators to conclude that the belt had failed were a drop in fan motor current from a normal operating current of 28A to an essentially no load cur-rent of 16A, and, a reduction in the differential temperature across the RRU heat exchanger from 40 degrees F to 9 degrees F indicating the absence of forced air flow circulation.
Loss of RRU-1, in con-junction with normally warm summer ambient atmospheric temperatures and a minor drywell steam leak, elevated drywell temperatures above the 160 degree F entry condition for emergency operating procedure (E0P) OE 3101, "Drywell Pressure and Temperature Control Procedure."
Drywell temperature is obtained from environmentally qualified tem-perature indicators TI-16-19-30A and B located approximately 180 de-grees apart at the 270 ft. elevation. On July 30, the licensee in-itiated trending of the drywell temperature profile.
Indicator TI-16-19-30B increased to 167 degrees F on July 30 and trended upward to a high resdin, sf 173 degrees F on August 3 and has since trended downward to 16i degrees F on September 4.
Indicator TI-16-19-30A increased to 155 degrees F on July 30 and trended upward to 160 de-grees F on August 3 and has remained at or below 160 degrees F through the conclusion of the inspection period. Although the plant has remained in OE 3103 since RRU-1 was secured, the procedure only directs continued monitoring of drywell temperatures for current conditions.
Review of the chromalox temperature indicating system (installed to detect high energy line breaks) identified abnormally high tempera-tures of approximately 250 degrees F at the card file 2 channel 3 indicator. This indicator is located in close proximity to TI-16-19-30B (190 degrees, 270 ft. elevation) and above the RCIC steam sup-ply line and isolation valve, RCIC-15. On August 3, the licensee implemented a temporary modification to bypass the RCIC-15 motor ac-tuator open limit switch and fully backseat the valve in an attempt
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. t'o isolate a potential. packing ' leak. On August 8, a similar tempor '
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ary modification was implemented to backseat the HPCI steam supply.
isolation valve, HPCI-15, (the. temporary modifications are discussed in'Section 4.4).
No appreciable drywell temperature reductions were
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realized as a result of.the valve backseating.
In conjunction with efforts to isolate the steam leak, the licensee performed an analysis.of the impact of elevated 1ocalized and average
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drywell temperatures on equipment environmental qualification (EQ)
assumptions. The analysis assumed equipment in 'the area of the sus-pected leaks was subjected to a temperature of 300 degrees F and the remainder of thr., equipment within the drywell was subjected to. the average elevated temperatures.
The analysis concluded the EQ equip-
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ment'could be exposed to the assumed elevated temperatures for the remainder of the current operating cycle without adversely impacting a;
qualified life.
-Licensee response to this situation has been appropriate. Control room operators properly identified the conditions for entry into the drywell temperature control E0P and drywell temperature trending was promptly initiated. Although unsuccessful, attempts to isolate the steam leak have been adequately evaluated and performed in a con-
' trolled manner. Environmental. qualification implications of elevated drywell temperatures were properly addressed. The inspector had no further questions.
7.
Maintenance / Surveillance 7.1 RCIC Inoperability On August.9, the RCIC trip and throttle valve, RCIC-1, failed to trip the RCIC turbine when actuated from control room panel (CRP) 9-4.
Isolation valve RCIC-131 was subsequently closed to secure the tur-bine and RCIC was declared inoperable.
Initial troubleshooting per-
-formed under MR 89-2540 identified valve stem binding caused by in-terference contact between the valve stem and actuator shaft coupling and the valve yoke. Maintenance personnel machined the yoke and chamfered the coupling corners to dimensions specifieo by the vendor and reassembled the valve. The valve subsequently tripped properly three times during cold condition testing.
However, when steam was admitted the valve failed to trip.
Further troubleshooting identi-fied that the additional force imparted on RCIC-1 by the steam during the open stroke sufficiently increased the valve coastdown travel to essentially backseat the trip valve disk. Therefore, there was in-sufficient exposed area on the top of the valve disk for steam pres-
.sure to impart the force necessary to trip the valve. The trip valve disk position is established by controlling the RCIC-1 open position clearance of the shaft coupling and sliding nut (in yoke assembly) to 0.010-0.040 inches. This is accomplished by adjustment of the valve
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motor actuator.open limit switch ' setting. The open limit switch set-ting was last adjusted on May 13, 1988 (inspection report 50-271/
88-06, MR 88-0870) following a similar RCIC-1 failure to trip. At that~ time an open limit switch setting of 0.030 inches was estab-lished in accordance with vendor instructions by manually positioning RCIC-1 with the handwheel. The valve was subsequently tested satis-factorily and returned to service. This process, however, failed to compensate for continued valve travel beyond the limit switch setting caused by the combined effects of motor coastdown and upward force imparted by steam pressure. Therefore, in actual' operation, the coupling was contacting the sliding nut and yoke.
Upon discovery of.
this condition, maintenance personnel performed the appropriate limit switch setting adjustments to ensure the recommendco clearances were maintained.
Ultimately, a static or cold cleara me of 0.153 inches was necessary to ensure the vendor recommended clearance was main-tained during electrical stroking. The valve was subsequently
. tripped twice satisfactorily from the control room and RCIC was de-clared operable on August 10.
Maintenance department response to this event was noteworthy. The licensee conducted a comprehensive root cause analysis which ulti-mately identified ambiguities in vendor limit switch setting in-structions.
Proposed and enacted corrective actions were properly evaluated and vendor concurrence w?s obtained. The inspectors had no further questions.
8.
. Engineering and Technical Support 8.1 Spent Fuel Racks
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The inspector reviewed documentation relating to the fabrication and installation of ten new type 304L stainless steel, high density spent fuel racks. At the time of the inspection, nine of the ten egg-crate type racks had been installed in the fuel pool. The outside perime-ter of the tenth rack (Rack 1) was visually examined by the inspector and found to be free of observable damage. The structural fillet welds appeared to be relatively smooth and free of gross surface ir-regularities, typical of welds made by the Tungsten Inert Gas Pro-cess. The racks contain Boral (B4C in an aluminum matrix) sheets as a poison material for absorbing neutrons.
Inspector review of licensee specification VY-ME-SI for the subject racks indicated that the racks were designed in accordance with the criteria described in the NRC Standard Review Plan Section 3.8.4, Appendix D, " Technical Position on Spent Fuel Racks." Applicable sections of ASME III, Subsection NF-1980 ED and Summer 1981 Addenda
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were used as guidance in the fabrication. The responsibility for analysis, design and fabrication as described in engineering design change request (EDCR)86-412 was assigned to Nuclear Systems (NES),
Danbury, Connecticut. The actual fabrication was performed by an NES
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subcontractor; U.S. Tc
' Die, Allison Park, Pennsylvania. A review
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of documentation indic 2d that the racks had been fabricated from corrosion tested, type 304L stainless steel in accordance with ASTM A240 specification. Weld wire conformeo to type 308/308L chemistry
& accordance to SFA 5.9 specification.
Correspondence was provided by the licensee ?. hat indicated Yankee Atomic metallurgical personnel had reviewed pertinent ASME Section IX welding procedures and' qualification,.as well us, material and non-destructive examination (NDE) certifications.
In cddition, numerous surveillance visits were made to U.S. Tool & Die during fabrication to assure that the required quality assurance (QA) program was being followad.
Inspector review of the reports and discussions with a key participant indicated that the surveillance' were very thorough with emphasis on workmanship, dimensional verification, performance test-ing, Boral traceability, personnel qualification, and weld inspec-tion. Where problems were found, the reports indicated that Yankee Atomic personnel provided necessary evaluation and followup to assure proper corrective action. The licensee identified deficiencies in a few of the racks in which many of the cells failed the drag test.
Corrective action consisted of hydraulically expanding the cells ac-
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cording to written instructions.
The inspector reviewed the specific document package for Rack No. I and confirmed that it contained most of the essential recort, such as
dimensional and performance tests; material traceability rems for the rack material, weld metal and Boral plates; and, visual m pene-trant test reports. Other essential records not included in the
document package were readily produced by the licensee. The chemical
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analysis reports of the Boral plates indicated conformance to ASTM-80 type 3 (B minimum 76%).
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Old spent fuel rack removal.and installation of new racks were per-formed in accordance with procedure FCI-86-412-002. The inspector concluded from a review of this procedure and accompanying records that the operations were performed in a detailed and controlled man-ner as evidenced by QA verification.
It should be noted that the last rack (#1) had not been installed because it failed to comply with the required two inch clearance between the rack and fuel pool wall. This condition is apparently due to the unevenness of the existing wall, is presently being evaluated for disposition, and will be addressed in EDCR-86-412 Change Notice 3.
Licensee options in-clude omitting rack No. 1 from the pool or reorienting it by 90 de-
grees.
Reorientation would require modification of the cask pad.
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The inspector noted that an augmented emergency cooling system, as required by the Safety Evaluation, is presently being designed.
In-stallation is planned for January 1994. Also, installation of the Boral plates for long term surveillance, as committed to by the lic-ensee, is presently in progress.
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Findings The. inspector determined that the spent fuel rerack program was l '
closely controlled and well executed. The new racks had been fabri-l-
cated and installed under a comprehensive QA program covering all aspects of the process.
Engineering documentation was good and in-
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dicated conformance-to applicable specification and code require-L t-ments. The inspector noted that an index for each of the record l
packages would facilitate information retrieval. No deficiencies were identified.
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8.2 Uninterruptible Power Supply On Sept.aber 3 the "A" train of the uninterruptible power. supply (UPS-1A)' alarmed and tripped. The unit was subsequently declared-inoperable. Maintenance activities to ascertain and correct the failure were in progress at the end of the inspection period. The UPS-1A last failed on June 5,1989 (IR 89-09 Section 7.1) and has failed several times in the past two years along with some failures of UPS-1B.
The licensee completed a review of-a UPS reliability study undertaken in response to repetitive problems and failures of the UPS. The cur-rent licensee intent is to replace both UPS units during the 1990 outage. The licensee is considering both static and rotating units as possible replacements.
Evaluation of the various options will be complete by the end of 1989. The decision to replace the UPS is a g
positive indication of the comprehensive licensee approach to overall plant reliability and safety, and demonstrates continuing conserva-tism in plant operations.
8.3 10 CFR Part 21 Report In accordance with 10 CFR Part 21, the licensee submitted a report on July 27, 1989 concerning the stem failure of v'alve RHRSW-89A (resi-dual heat removal service water system) which was previously dis-cussed in IR 50-271/89-07 Section 7.2.
The valve stem failed on April 5, 1989, at the transition between the shaft and stem flange.
Licensee evaluation concluded that the failure of the stem was the result of several factors including improper heat treatment of the 410 SS alloy used in the stem which increased its susceptibility to stress corrosion cracking (the material property changes that oc-curred when the shaft was forged were not recognized); presence of a sharp stress riser at a point of high stress in the stem; mismatched stem and disc nut; and, incorrect orientation of the valve. The mis-matched stem and disc part problem was due to part number control difficulties between Walworth and Aloyco. This occurred when Aloyco replaced Walworth as the firm supplying the components.
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Following an August 1988 failure'of the stem, parts were replaced E
'with a stem from Walworth supply and a disc-nut from Aloyco supply.
The parts difference resulted-in a smaller surface area for contact between the stem and disc nut.
As detailed in IR 89-07,-the licensee. completed several short term corrective actions to address this issue.
Long term corrective ac-tions committed to in the 10 CFR Part 21 report include:
The licensee will either revise stem ordering information to
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require a.radiused corner on the stem or revise appropriate plant documents to indicate that replacement stems will be modiried before use.
The licensee directed Yankee Atomic Quality Assurance Department
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to review the vendor audit program relative to special process control at the sub-vendor level and, if necessary, revise the auditing-program.
Licensee procurement procedures will be revised to include a
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requirement for a vendor surveillance to be performed to verify consistency in part numbering in the event a change of ownership takes place.
The licensee also requested Yankee Atomic to develop a program
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to address material and design concerns related to the use of high strength steels.
The licensee will perform ultrasonic testing on the RHRSW-89A
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stem on a bimonthly basis until the orientation of the valve is corrected. This testing method has the capability of detecting any indications of a stem crack in its early stages.
The licensee will evaluate the need to inspect / replace valve
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components in RHRSW-898. Any corrective action warranted will be completed no later than the 1990 outage.
ihe RHRSW-89A valve will be rotated to the correct orientation
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at the first convenient opportunity, which is currently sched-uled during the next refueling outage.
The inspector reviewed the subject report and asscciated corrective actions, and found no deficiencies. The licensee approach to long-term resolution of this issue is comprehensive and attests to a sound engineering program.
8.4 System Configuration Discrepancy - RBCCW On August 31, 1989, the licensee determined that a reactor building closed cooling water (RBCCW) system configuration discrepancy existed between the VYNPS final safety analysis report (FSAR) and the current
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I as-built condition. Table 7.3.1 of the FSAR indicates that the motor
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operated valve (MOV) RBCCW ~17 should be powered from an "ac emer-i gency power bus."
The current configuration for RBCCW-117 has the
. valve powered from a non-emergency (safety) power source.
The licen-see determined that this discrepancy constituted c condition'outside
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the plant design basis and'made the_ appropriate 10 CFR 50.72 notifi-s
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cation. The licensee convened the plant operations review committee (PORC) to review and approve a justification for continued operation (JCO) covering this condition.
The JC0_was accepted primarily based upon the fact that the valve is not safety related; that the RBCCW system is a closed loop inside of containment; and, the RBCCW. system insic;e of containment is seismically qualified. The JC0 was accepted by PORC on an interim basis pending further investigation of the de-sign basis and development of a modification that would correct the
' discrepancy'and power RBCCW-117 from an emergency power supply.
These actions were to be completed and ready for PORC review by Sep-
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tember 8, 1989.
The licensee determined that no -interim compensatory measures were necessary regarding this condition. The licensee basis
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for this was that the valve was not safety-related; RBCCW was a closed seismic loop inside containment; and, the valve had no asso-ciated limiting condition for operatic,a (LCO) in the plant technical specifications (TS).
- Findings The' inspector reviewed the. design discrepancy and licensee corrective actions. The licensee identified'the discrepancy during a review of the results of a safety system functional inspection (SSFI) which had-been' performed recently at a similarly designed plant (GE BWR Mark 4). The licensee practice of timely reviews of industry experience is noteworthy and attests to a questioning approach to plant safety and operations.
Licensee response t6 the deficiency was appropriate.
The condition was reviewed and assessed from a safety perspective. A corrt.ctive action plan was formulated and executed in a timely man-ner. Overall PORC involvement in this process has improved, and, in this instance, provided a' sound technical and safety basis review of the design discrepancy and associated corrective actions.
The inspector noted that the application and condition of RBCCW-117 was previously addressed by the licensee in response to the TMI Ac-tion Plan (NUREG 0578 Item 2.1.4 and NUREG 0737 Item II.E.4.2).
The inspector requested that the licensee review the responses to these
. items, as applied to RBCCW-117, to ensure continued conformance. The results of this licensee review and the completion of the final cor-rective actions associated with the RBCCW-117 power supply remain as an unresolved issue (50-271/89-12-01).
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9.
Licensee Event Reporting (LER)
The 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.
9.1 LER 89-14 The LER 89-14, " Reactor Core Isolation Cooling System Inoperability Due to Motor Burn Out on RCIC-21 Valve", addresses a system inoper-ability when the RCIC-21 valve failed to close during a surveillance test on June 7, 1989. Details of this event appeared in Se-tion 7.2 of IR 89-09.
This was a well written and comprehensive LER.
Event description, cause and corrective actions were well documented. The causal analysis demonstrated an insightful engineering assessment.
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Planned long term corrective actions appear adequate to prevent re-currence. The licensee initially determined that this event was not reportable based upon the fact that the valve had failed only after reaching the open positica and therefore RCIC remained operable. The inspector noted that RCIC had been declared inoperable by operators in accordance with 10 CFR 50.72 on June 7.
The RCIC system remained inoperable during RCIC-21 repairs until June 8 when valve and system operability were declared. The inspector discussed these details with various levels of licensee management. The licensee subse-quently determined that this event was reportable under 10 CFR 50.73 requirements. With the exception of lack of timeliness noted above, this LER fulfilled the above criteria.
9.2 LER 89-09 The LER 89-09, " Lack of Redundancy in Residual Heat Removal Service Water Systems," addresses a deficiency in a design change to the residual heat removal (RHR) service water (SW) system which caused a loss of subsystem independence due to susceptibility to a potential single mode of failure.
Details of this event appeared in Section 8.1 of IR 89-09. This LER presents an extensive analysis of the background of this design deficiency. The causal analysis and cor-rective actions are comprehensive. No deficiencies were identified.
10.
Review of Licensee Response to NRC Initiatives 10.1 IEB 80-06 Inspection and Enforcement Bulletin (IEB) 80-06, " Engineered Safety Feature (ESF) Reset Control" requested various licensee reviews and actions regarding reset controls for ESF equipment. The licensee response to this bulletin vas forwarded to NRC:RI in a letter dated
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June 13, 1980. Inspector review of the response and associated-in-spection activities were described.in IR 50-271/82-01. The inspec-
.i tion determined no inadequacies existed. All' inspection activities
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L associated.with the subject bulletin are complete. This issue-is closed.
L 11.
Review of Periodic and Special Reports.
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 l
NRC-required data; (2) that test results and supporting information were consistent with design predictions and performance' specifications; and (3) that planned corrective actions were adequate for resolution of the probl er,.
The inspector also ascertained whether any reported information should be classified as an abnormal occurrence.
The following reports were reviewed:
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Monthly Statistical Report for plant operations for the months of
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July and August 1989.
Feedwater leakage detection system monthly performance summary for
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July 1989.
No inadequacies were identified.
12. Management Meetings
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At periodic intervals during this inspection, meetings were held with senior plant management to discuss the findings. A summary of findings for the report period was also discussed at the conclusion of-the inspec-tion and prior to report issuance.
No proprietary information was iden-tified as being included in the report.
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