IR 05000271/1998004
| ML20249A123 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 06/04/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20249A115 | List: |
| References | |
| 50-271-98-04, 50-271-98-4, NUDOCS 9806160082 | |
| Download: ML20249A123 (73) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
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Docket No.
50-271 Licensee No.
DPR-28 Report No.
98-04 Licensee:
Vermont Yankee Nuclear Power Corporation Facility:
Vermont Yankee Nuclear Power Station Location:
Vernon, Verrnont Dates:
March 15 - May 2,1998 j
inspectors:
William A. Cook, Senior Resident inspector Edward C. Knutson, Resident inspector Ronald L. Nimitz, Senior Radiation Specialist, Region i Thomas F. Burns, Reactor Engineer, Region i Approved by:
Curtis J. Cowgill, lil, Chief, Projects Branch 5 Division of Reactor Projects l
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9906160092 990604 i
PDR ADOCK 05000271
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EXECUTIVE SUMMARY l-Vermont Yankee Nuclear Power Station l'
NRC inspection Report 50-271/98-04 This integrated inspection included aspects o' licensee operations, engineering, f
maintenance, and plant support. The report covers a seven week period of resident inspection; in addition, it includes the results of announced inspections by a regional radiation protection specialist.
Operations The shutdown for the refueling outage was well controlled and the operators
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L demonstrated good teamwork and communications. (Section 01.1)
On two occasions, known equipment problems were not discussed in the procedure
or during the pre-evolution brief, and presented minor challenges to the operators l
when they occurred. (Section 01.1)
Maintenance incorrect orientation of a peripheral fuel bundle was identified by an alert operator e
during an unrelated activity, and was the result of 1) an error in the fuel loading
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schedule, and 2) an error in performance of the fuel movement procedure. These errors were mitigated by the fact that they were identified and corrected during the ongoing fuel movement operation, and that -n moortunity for discovery of the
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condition still existed during the full <-
.... nc ano.. Sase of refueling. Licensee
' corrective actions were prompt s', ;,spror 'ee, whir, resulted in a non-cited violation consistent with sectier J d ' ' of the.J ~ enforcement Policy. (Section M1.2)
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l VY's was successfulin identifg vg r i sking fuel bundle through in-core e
mpping. Continuation of in-core is (,
, the leaking fuel bundle had t>een
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- <opropriate steps to identify the cause identified was conservative. VY s
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of the fuelleak. (Section M2.1)
The short circuiting of the "A" main static
.tery that occurred on April 1 was the
- result of inadequate supervisory oversight of preparations to perform the battery charge. VY's immediate corrective actions were appropriate, and the initial root -
cause evaluation was adequate. Failure to perform steps of the governing procedure in series was a violation of technical specification 6.5. (Section M2.2)
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The NRC concluded that the licensee followed the plant technical specifications for
the degraded primary containment penetration associated with the recirculating
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water sample line. Also, the licensee provided an appropriate basis for the retraction of the event notification. (Section M8.1)
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Enoineerina VY's initial investigation of the effect of the fault on the "A" main station battery e
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VY subsequent analysis, based on additional information obtained from the battery manufacturer, concluded that the battery had not been damaged as a result of the event. The NRC determined that this conclusion was adequately founded. (Section E2.1)
At the close of the inspection period, VY was developing a modification to upgrade
two blowout panels in the main steam tunnel, to be completed during the current outage. Engineering actions to resolve the remaining issues associated with BMO 96-18, " Main Steam Tunnel Slowout Panel," and the reactor building HELB concerns were continuing. VY's resolution of these issues to support startup from the current refueling outage will be examined as part of the continuing NRC inspection.
-(Section E2.2)
Thorough licensee review of an in-service test procedure identified that two thermal
relief check valves had not been adequately tested in the forward direction during the 1996 refueling outage. Immediate operability determinations were prompt and adequately founded. (Section E2.3)
Plant Suonort l
VY radiological control coverage of significant work involving torus modifications
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was weak. Additionally, industrial safety issues involved in the work were not i
immediately recognized and addressed by the licensee until brought to management's attention by the inspector. Upon notification, the licensee suspended work activities pending improvements in radiological control coverage, and took action to address the industrial safety concerns. (Section R1.1)
Overall, the licensee implemented an ALARA program that met the requirements of
the regulations. Notwithstanding, ALARA planning and preparation activities were limited in scope and areas for improvement were identified. (Section R1.2)
Applied radiological controls for ongoing work activities were generally well
implemented. The licensee implemented generally effective external and internal exposure control programs. A licensee-identified condition involving high radiation areas with inadequate barricades and postings was appropriately addressed.
(Section R1.3)
VY implemented an effective radioactive material and contamination control program
during the outage. (Section R1.4)
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VY maintained an effective program for the training and qualifications of contractor
radiation protection personnel. The licensee implemented an effective audit of ongoing radiological controls activities. (Section R5.1)
A pathway was identified which provided unmonitored access from the turbine
building into the reactor building vital area. Appropriate compensatory measures were established, and long term corrective action is under development. The pathway for unmonitored vital area access was a violation of the VY Physical Security Plan. (Section S2.1)
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TABLE OF CONTENTS EXEC UTIVE SU M M A RY.............................................. ii Summary of Plant Status
............................................1 1. Operations
....................................................1 O1 Conduct of Operations.................................... 1 01.1 Reactor Shutdown Observations........................ 1
Miscellaneous Operations issues............................. 2 08.1 (Closed) Unresolved item 95-05-01
......................2 08.2 (Closed) Unresolved item 97-01-01...................... 3 11. M ai nt e n a n c e................................................... 3 M1 Conduct of Maintenance................................... 3 M1.1 Maintenance Observations............................3 M1.2 Fuel Bundle Orientation Error........................... 4 M2 Maintenance and Material Condition of Facilities and Equipment....... 5 M2.1 Failed Reactor Fuel due to Interaction with Foreign Material..... 5 M2.2 Main Station Battery Short Circuit During Restoration from Test Di s ch a rg e........................................ 6 M2.3 Observations of Plant Material Conditions.................. 8 M8 Miscellaneous Maintenance issues............................ 9 M 8.1 Follow-up Licensee Event Report Retraction for Event Report 32833
...............................................9 111. E n g in e e rin g.................................................. 1 1 E2 Engineering Support of Facilities and Equipment.................11 E2.1 Main Station Battery Operability Following Fault............11 E2.2 (Update) Inspector Follow.Up Item 9 6-1 1 -0 2............... 12 E2.3 Inadequate installation Testing Identified for Thermal Overpressure R elie f Valve s..................................... 13 E2.4 (Closed) Inspector Follow-Up Item 9 6-1 1 -01............... 14 E8 Miscellaneous Engineering issues............................ 15 E8.1 (Closed) Inspector Follow-Up Item 97-02-06............... 15 IV. Plant Support
................................................16 R1 Radiological Protection and Chemistry (RP&C) Controls............ 16 R1.1 Refueling Outage Radiological Controls (Program Changes)..... 16 R1.2 Refueling Outage Radiological Controls (ALARA Planning and Prep aratio n)...................................... 16 R1.3 Refueling Outage Radiological Controls (Internal and External Exposure C ontrols)........................................ 19 R1.4 Refueling Outage Radiological Controls (Control of Radioactive Materials and Contamination)
.........................21 R5 Staff Training and Qualification in Radiation Protection and Chemistry.. 21 v
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R5.1 Radiological Controls Training and Qualification............. 21 R7 Quality Assurance in RP&C Activities......................... 22 R8 Miscellaneous issues.................................... 23 l
R8.1 Plant Tour Observations............................. 23 R8.2 Meeting With Station Management..................... 23 R8.3 (Closed) Unresolved ; tem 9 6-1 1 -0 3......................' 2 4 S2 Status of Security Facilities and Equipment..................... 24 S2.1 Potential Pathway for Unmonitored Vital Area Access........ 24 l'
V. Management Meetings
..........................................25 X1 Exit Meeting Summary................................... 2 5 X3 Review of Updated Final Safety Analysis Report (UFSAR)........... 25 ITEMS OPENED, CLOSED, RECLASSIFIED, AND DISCUSSED.................. 26 I
PARTIAL LIST OF PERSONS CONTACTED............................... 27 i
LIST OF ACRONYMS USED
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l Report Details Summarv of Plant Status j-At the beginning of the inspection period, Varmont Yankee (VY) was operating at 87
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percent power, in a gradual power reductior due to fuel depletion (coastdown). On March 20, a planned reactor shutdown was cc.idu:ted to begin the cycle twenty refueling outage.
The main generator output breaker was eponed at 10:50 a.m. and the reactor mode switch was placed in the shutdown position at 11:37 a.m. The plant was cooled down and cold L
shutdown conditions were established at 10:25 p.m. Major outage activities include
replacement of strainers on the emergency core cooling system torus suction lines, shortening of the torus downcomers, refurbishment of the torus internal coating, reactor refueling, including sampling to identify the location of the leaking fuel (discussed in
. inspection report 50-271/97-11), in vessel equipment inspections, replacement of four control blades and seven local power range monitor detector strings, inspection of the high pressure turbine, and installation of a no-load disconnect switch for the main generator.
The outege was originally sche fuled to last for 41 days; however, delays due to a variety of causes, and outage work scope increases, have resulted in lengthening of the outage.
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At the close of the inspection period, VY projected that startup would occur near the end of May.
I. Operations
Conduct of Operations'
01.1 Reactor Shutdown Observations
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Insnaction Scone (71707)
The inspector observed portions of the steam plant and reactor shutdown from the control room.
b.
' Observations and Findinas
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- The inspector noted that the power reduction was well controlled. The control room j
operations staff functioned as a team, and the shift supervisor led discussions of significant operations (such as operation of the reactor mode switch and transition of. steam loads to the bypass valves) prior to them being performed. However, there were two instances when abnormal equipment responses unnecessarily I
challenged the operators, in that they should have been anticipated based on previous shutdowns. Specifically, an automatic main turbine trip occurred within a minute of decoupling the main generator from the grid, caused by a high moisture separator level signal. The same sequence of events occurred during shutdown for the previous refueling outage, and investigation had identified the likely cause as being level instrument response to the transient, rather than an actual high level
' Topical headings such as 01, M8, etc., are used in accordance with the NRC standardized reactor inspection report outline. Individual reports are not expected to address all outline topics.
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condition. Although this dou not cause a major transient during plant shutdown (reactor power is below the automatic turbine trip / reactor trip scram setpoint), it required the operators to deal with an unexpected plant response. Given that corrective action for the instrument anomaly was still being evaluated, the inspector considered that a note in the procedure, indicating that this event may occur, would have been appropriate. The second problem was that the "C" main feedwater pump tripped due to low flow, within one minute after the reactor was manually scrammed. This was due to the slow response time of the pump recirculation valve (a known condition), and had occurred during previous plant shutdowns. However, the possibility that the pump would trip was not discussed as part of the pre-evolution brief.
The inspector considered that these two problems constituted operator workarounds. However, the Operations Department Workaround List did not include either of these problems. The inspector discussed this with VY management, who agreed to review the issue.
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Conclusions The shutdown was well controlled and the operators demonstrated good teamwork and communications. On two occasions, known equipment problems were not discussed in the pmcedure or during the pre-evolution brief, and therefore presented minor challenges to the operators when they occurred. VY agreed to reexamine their bandling of these operator workarounds.
Miscellaneous Operations issues 08.1 (Closed) Unresolved item 95-05-01: Adequacy of Emergency Operating Procedures (EOPs) Revision Process. This unresolved item was initiated in Inspection Report No. 50-271/95-05, dated April 7,1995, and highlighted some examples where the adequacy of the licensee's EOP verification and validation (V&V) process could be challenged. As stated in Report No. 95-05, VY management shared the inspector's concern and initiated a self-assessment to examine the problem. Since April 1995, the NRC staff has examined a number of operator licensee candidates and observed licensee emergency preparedness drills. In addition, since late 1996, the licensee has been conducting a design basis documentation review which has identified a number of "old design issues" which have had direct and indirect impact on the EOPs. Individually and collectively, these minor errors in EOPs have not had a significant impact c ' *he safe operation of the facility, and do not necessarily constitute an inadequate EOP V&V process. The licensee's assessment identified no specific cause for the EOP errors and concluded that cor'tinued use and routine scrutiny of the EOPs by the VY staff and operators has led to further improvements and refinements in the overall objectives and methodologies in the procedures.
The adequacy of the VY EOPs and associated V&V process will continue to be scrutinized by the NRC as a matter of routine review and followup, via the core inspection program. Since the identified errors in the EOPs constitute violations of
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minor significance and are not subject to formal enforcement action, unresolved item URI 96-05-01 is closed.
08.2 (Closed) Unresolved item 97-01-01: Deviation from Emergency Operating Procedures (EOPs) Compliance. This unresolved item was initiated to track the licensee's resolution of their deviation from an emergency operating procedure
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during the conduct of their January 1997 planned emergency exercise. As stated in inspection Report No. 50-271/97-01, the VY staff utilized a reactor pressure vent (RPV) pathway other than one specified in EOP step ALC/CF-4. Subsequent review by the VY staff concluded that they should have followed the EOP guidance to ensure the adequacy of the prescribed vent pathways. VY's deviation from the EOPs was due to a training error, lack of knowledge of the specific design, and lack of procedural understanding. However, the problem was identified by VY during the exercise critique, and appropriate and effective corrective action was taken. This failure to follow EOPs during a planned exercise (for training) did not constitute a violation of regulatory requirements. Therefore, unresolved item URI 97-01-01 is
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closed.
11. Maintenance M1 Conduct of Maintenance M1.1 Maintenance Observations a.
Insoection Scooe (62707)
The inspector observed portions of plant maintenance activities to verify that the correct parts and tools were utilized, the applicable industry code and technical specification requirements were satisfied, adequate measures were in place to ensure personnel safety and prevent damage to plant structures, systems, and components, and to ensure that equipment operability was verified upon completion j
of post maintenance testing.
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b.
Observations. Findinos and Conclusions The inspector observed all or portions of the following maintenance activities:
Control rod drive 6eolacement. observed March 23
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Observed from outside the drywell at the contu,1 rod drive tunnel.
Observed one removal, lift transfer into storage box, and movement of replacement
unit back into containment. No problems noted, good RP coverage, advising people
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f in the vicinity of changing radiological conditions as the old drive approached the area boundary, i
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in-core fuel minnina. observed March 25 e
. Done by GE with VY reactor engineering oversight.
Observed from the refueling bridge and from tha monitoring station directly adjacent the cavity; also observed three movements of the sipping equipment, which simultaneously sampled four fuel bundles. The contractor later determined that the
.four-bundle sipping hood was not functioning properly, and switched to a single bundle hood. The Isaking fuel bundle was identified using this equipment, and subsequently confirmed using the repaired four-bundle hood.
Reactor fuel movements. observed March 31
Reactor refueling was performed in two phases, with the first to establish a core configuration to support local power range monitor detector replacements and control rod blade changeouts. The first phase consisted of 256 fuel bundle movements. The inspector observed five fuel bundles being removed from the core and transferred to the spent fuel pool. The inspector noted that independent verification of grapple engagement by two independent means was consistently
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being performed prior to fuel movement.
M1.2 Fuel Bundle Orientation Error a.
Inspection Scope (92902)
During operations to change out reactor control rod blades on April 24, an operator.
noted that the fuel bundle in core location 29-04 (on the periphery of the core) was
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channel fastener pointing away from the core, however, the bundle at 29-04 was positioned 90 degrees out of this orientation. The bundle had been installed in this
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location earlier in the_ outage, during the initial fuel shuffle. The inspector reviewed selected portions of VY's activities to identify the cause of this event and corrective action implementation.
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Observations and Findings Control rod blade change out activities were stopped pending resolution of the ll issue. At the time, one control rod (22-27) had been fully withdrawn in preparation
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for blade change out, however, it was not in close proximity to the fuel bundle at l
- luue. ' An event report was generated and a four hour non-emergency notification was made to the NRC operations center in accordance with 10 CFR 50.72.
VY determined that bundle 29-04 had been incorrectly oriented as a result of an error in the fuel movement schedule. That is, the bundle had been oriented in accordance with the fuel movement schedule, however, the specified orientation was incorrect. The fuel movement schedule is generated by a computer program, however, the orientation is a manual data entry. This personnel error was not L1__--_-------______________________
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. caught during the independent verification of the schedule. In addition, the fuel L
movement procedure, OP-1101, " Management of Refueling Activities and Fuel L
Assembly Movement," includes steps for the refueling SRO to verify proper orientation of the peripheral bundles prior to insertion into the core. However, the procedure also contains a caution that there shall be no deviation from the fuel loading schedule. Since OP-1101 is a " reference use" procedure whereas the fuel loading schedule requires that the operators initial for each fuel movement, the fuel loading schedule is the portion of the procedure that is in continuous use. The inspector concluded that the error in the fuel loading schedule was the principle cause of the fuel bundle orientation error. Additionally, the full core verification that -
is performed after the conclusion of all fuel movements was verified to have been j
correct; therefore, a proceduralized verification was in place that would have (
identified the error.
As corrective action, VY reverified the entire fuel loading schedule, including a j
check against the vendor's cycle management report. Personnel involved in fuel
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movements were briefed on the event, including discussion of how to verify bundle orientation. The fuel loading schedule was modified to include indication of
< peripheral bundles, and the operator aid that is used on the refueling bridge to record
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Conclusions Incorrect orientation of the peripheral fuel bundle was identified by an alert operator during an unrelated activity, and was the result of 1) an error in the fuel loading schedule, and 2) an error in performance of the fuel movement procedure, OP-1101.
. Although each of these errors constitutes a violation of regulatory requirements, they are mitigated by the fact that they were identified and corrected during the
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l ongoing fuel movement operation (as opposed to being identified after the operation was complete). In addition, an opportunity for discovery of the condition still existed during the full core verification phase of refueling. Licensee corrective
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actions were prompt and appropriate. Accordingly, this non-repetitive, licensee-
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identified and corrected violation is being treated as a non-cited violation, consistent with section Vll.B.1 of the NRC Enforcement Policy. (NCV 50-271/98-04-01)
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Maintenance and Material Condition of Facilities and Equipment M2.1.. Failed Reactor Fuel due to Interaction with Foreion Material a.
Inanection Scope (92902)
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A failure of a reactor fuel pin cladding was identified by VY on October 26,1997, as discussed in inspection report 50-271/97-11. The inspector reviewed selected portions of VY's activities to identify the fuel bundle containing the failed fuel pin prior to commencement of reactor refueling, a_____-________._ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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b.
Observations and Findinos in-core fuel sipping was performed to identify the location of the leaking fuel bundle.
' This technique uses a hood that sets on top of a fuel bundle and allows coolant from within the bundle to be sampled for isotopic analysis. The activity was performed by GE contractors, under the supervision of VY reactor engineering personnel. On March 28, the leaking fuel pin'was determined to be in fuel bundle VJ5333, located in core position 23-18. This location was consistent with the
. results of the power suppression testing that was performed shortly after the leak had developed. VY informed the NRC of the positive identification of the leaking fuel bundle in accordance with applicable regulations. Although only a single leak
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was suspected, VY proceeded with sipping of all la-core fuel bundles that would remain in service during the upcoming cycle. No other leaking fuel was identified.
The subject fuel bundle was moved to the spent fuel pool for disassembly and examination. The leak was found to be from a cladding breach in a single fuel rod.
The cause of the breach was interaction of the fuel rod with a piece of foreign material (fretting),' which was still present. VY had the material analyzed at an off-
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consistent with a wound metallic gasket. At the close of the inspection period, VY's review of this event, including efforts to identify the specific source of the material, were still in progress, c.
Conclusions i
- VY's efforts to identify leaking fuel through in-core sipping were successful in identifying a single leaking fuel bundle. Continuation of in-core sipping after the L eaking fuel bundle had been identified, so that all fuel bundles that were to remain l
in service were checked, was conservative. VY was taking appropriate steps to g
. attempt to identify the foreign material which was the cause of the. leak.
M2.2 Main Station Batterv Short Circuit Durina Restoration from Test Discharas i
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Insoection Scoon (g3702. 929021 On April 1,1998, the VY maintenar.ce staff was conducting the system restoration phase of Operating Procedure (OP)-4215, " Main Station Battery Performance /
Service Test," revision 8, dated March 27,1998, following the 8-hour discharge test on the "A" main station battery. Battery restoration was to be performed in
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accordance with Section 5, " Post Discharge Recharge Charge." At approximately 1:40 p.m., non-insulated portions of the positive and negative cables in the cable vault room were accidentally crossed, causing a direct short across the battery.
This short resulted in the cables being arc welded together at the point of contact, f acial electrical flash burns to the individuals involved in the activity, and the overheating and burning of cable insulation at the negative battery terminal. The
. Inspector reviewed the circumstances leading up to the event and the licensee's response to the event.
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Observations and Findinos
-.From interviews with the personnel involved in the incident, the inspector
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determined the sequence of events that led to the incident. Late on the morning of April 1, a test discharge of the "A" main station battery was completed, and
' electricians began preparations to recharge the battery in accordance with section 5,
" Post Discharge Recharge Charge," of OP-4215. Three electricians were involved in this task: A VY electrician, who was acting as the first line supervisor; a lead contract electrician; and, a contract electrician.' The VY electrician was going to connect the installed battery cables to a deenergized spare battery charger (procedure step 5.3). These cables are routed to the battery room (located on the next floor up) and terminate in a junction box. The two contract electricians were going to install a shunt circuit (used to measure charging current) on jumper cables that are used to connect the charger (via the cables in the junction box) to the battery (step 5.4). Their intent was to connect the jumper cables to the charger cables in the junction box in parallel with installing the shunt circuit (connected to the positive cable at the connection between the jumper and the junction box cable),
so that these intermediate connections could be insulated prior to making the final-connections to the battery. Attaching the jumper cables to the junction box cables -
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and connecting to the battery were both included in step 5.5.
The contract electricians completed the negative cable intermediate connection
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(with the exception of insulating it), but then noted that the electrical ratings of the
' shunt device that was staged at the work site.were diffeent than what was called for by the procedure. The lead contract electrician obuned a shunt device with the correct ratings, but then noted that it did not have a cahoration sticker attached.
Intending to resolve the issue later, the lead contract electrician deferred connecting the shunt device and proceeded to connect the jumpers to the battery. At this point, the positive cable intermediate connection had been loosely made up-(fasteners not tightened, pending installation of the shunt device)., Further, it was not insulated and Nas in close physical proximity to the negative cable intermediate connection. With the jum;.sr connections to the battery completed, the lead
' contract electrician was determining what tools would be required to perform the r
shunt installation when she inadvertently moved the two intermediate connections into contact. Since neither connection was insulated, this resulted in a short circuit across the battery. The positive and negative cables became welded together at the point of contact, thereby maintaining the short circuit. Smoke productd by the initial contact was sensed by the fire detection system, which produce:n a fire alarm.
The contract electricians attempted to break the circuit by disconnecting the negative jumper from the battery. However, this produced a high resistance contact since the fastener was loosened, and the resultant heat caused the jumper ca' ale j
insulation to catch fire.
The VY electrician had become involved in other activities, and had not yet started
- attachment of the cables to the battery oharger. He was alerted to the problem by
.the. announcement of a fire, and proceeded to the battery room. When.he determined what was occurring, he put on a pair of high voltage insulated gloves L
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fire was put out with a hand-held fire extinguisher.
l In response to this event, VY generated a level-one event report (high significance
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i event, in accordance with their corrective action program procedure). A task force l
was formed to investigate the event and to perform the root cause evaluation.
Pending cause determination, VY stopped work on jobs that appeared to share q
similar characteristics (i.e., work on battery systems).
l The licensee's formal root cause evaluation, concluded on April 8, identified that the principal cause of the accident was procedural non-compliance for the " continuous use" procedure, OP-4215. While there was procedural non-compliance, in that steps were being performed in parallel rather than in series, the inspector noted that the procedure itself lacked some specificity, which could have prevented the event.
Specifically, step 5.5 of the procedure was written such that the sequence of jumper connections between the battery and the intermediate connections at the terminal box was not specified. However, the inspector concluded that this procedural concern did not contribute to the event, since the personnel who were
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involved in the incident stated that they understood the required sequence of connections.
Technical specification 4.10, " Auxiliary Electrical Power Systems Surveillance Requirements," section A.2, " Battery Systems," includes the requirement to perform test discharges of the main station batteries. Technical specification 6.5, " Plant Operating Procedures," section A, in part requires that detailed written procedures covering specified areas, including surveillance and testing requirements, shall be prepared and approved and that all procedures shall be adhered to. The failure to
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perform OP-4215 steps 5.3 through 5.5 in series was a violation of technical j
specification 6.5. (VIO 50-271/98-04-02)
l At the close of the inspection period, the VY task force was stillin the process of developing their final report concerning the cause of this event and any additional proposed corrective actions.
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Conclusions The short circuiting of the "A" main station battery that occurred on April 1 was the result of inadequate supervisory oversite of preparations to perform the battery l
charge. Connection of the battery charger and the cable installations were treated as production activities rather than as sequential steps in a procedurally-driven evolution. The personnel who were involved lost sight of the requirement that the steps of a continuous use procedure be performed in sequence. Similarly, personnel t
apparently lost sight of the fact that the discharged battery was still a substantial
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electrical power source. The procedural guidance on the installation sequence was weak, but did not appear to be a significant factor in the event. VY's immediate
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corrective actions were appropriate, and the initial root cause evaluation was adequate.
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This event was of minimal significance to the safe operation of the reactor plant, l"
because the battery had been removed from service prior to the event and was not l --
required to be operable by the existing plant conditions. However, the failure to perform OP-4215, steps 5.3 through 5.5, in series was a violation of technical specification 6.5.'
j M2.3, Obmarvations of Plant Material Conditions (62707)
Foreian Material Exclusion (FME) Protection on Eauinment Vents -- During a routine plant tour shortly after the start of the outage, the inspector noted that the casing vents for the high pressure coolant injection (HPCI) and reactor core isolation cooling-(RCIC) pumps were open. Because these are upward-pointing vents, the inspector was concerned that they represented an unprotected path for introduction of foreign material (i.e., falling material from overhead work) into these safety-class components. In response to this concern, VY installed flexible tubing on the vents, i.
[
with the open ends held pointing downward with tape, and labeled with reference to
'the FME controls procedure, AP-6024. In addition, the use of this type of FME protection was extended to open vents on safety-related systems piping. At the
{
close of the inspection period, VY was evaluating applicability criteria (such as
'
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minimum applicable pipe size) in preparation for proceduralizing the requirements.
'The inspector considered that these actions adequately sddressed the FME concern.-
- Duct Tana on Stainless Steel Pining -- During a tour of the drywell, the inspector a
noted duct tepe on stainless steel instrument piping associated with the main steam e
system. The tape had ueen in place for at least the previous operating cycle, as
!-
indicated by the discoloration caused by high temperature. The inspector was concerned because duct tape can contain high levels of chlorides, which can lead to i
stress corrosion cracking. Further inspection revealed that duct tape was in use on L
small-bore stainless steel piping, apparently as a cushioning material, at piping -
!.
. restraints on various systems, including the control rod drive system. From the
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I l
appearance and locations of the hangers, the inspector considered it likely that the duct tape had been put in place during original construction.
- The inspector presented this concern to the licensee, and was informed that the
< issue had been addressed several years earlier. The inspector reviewed this
,'
evaluation (response to category A commitment MEC9405, dated November 4,
'
1994), and noted that it addressed recent uses of duct tape and instituted a requirement to use nuclear grade tape in the future. However, the inspector l
determined that it had not addressed the duct tape that had been installed during original construction. VY generated an event report to address this concern within their corrective action program. VY performed system walkdowns to determine the extent of the conditioni and also concluded that there had been wide spread use of duct tape during original construction. VY analyzed samples of the original
__
construction duct tape and determined that the chloride concentration was acceptably low. In addition, subject piping in 10 locations was examined by VY using a liquid penetrant test, and no cracking was identified. The inspector *
considered that the licensee's actions adequately addressed the concern.
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l M8 Miscellaneous Maintenance issues l
M8.1 Follow-un Licensee Event Reoort Retraction for Event Reoort 32833 (90712)
On February 5,1998, the licensee notified the NRC via the FNS that they were j
retracting the event notification for "being seriously degraded" per 10 CFR 50.72 (b)
'
(1) (ii). Event Number 32833 was reported to the NRC on August 28,1997 at about 4:00 p.m., when the licensee discovered during testing that two, in-series, primary containment isolation valves associated with the reactor recirculating water
sampling system were inoperable. The event report was based on two reporting requirements resulting from: (1) initiating a plant shutdown in accordance with technical specifications; and, (2) identifying a seriously degraded condition (operation outsida the technical specifications) during operation. This event and the licensee initial response to the degraded containment isolatioa valves was l
documented in NRC inspection report 50-271/97-06, paragraph M1.4.
j During this inspection period, an in-office review of the licensee's basis for the event retraction was conducted, as well as, a partial on-site review of selected corrective actions taken by the licensee due to the valve test failJre on August 28, 1997, including verification that: (1) the licensee had scheduled a modification to the valve (s) position indication system during this current refueling outage; and, (2)
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a revision to the test procedure had been implemented incorporating the new j
closure time acceptance criteria based on the modified position indicators.
The licensee retracted the part of the event report identifying the seriously degraded condition during operation. The basis for that retraction was that the licensee determined that operati ns were maintained within the technical specification allowed outage time. The failure of valve RV-39 to close occurred during the testing on August 28. The two containment isolation valves in the affected penetration are both normally closed and were opened on August 28 to support the " closure time" surveillance test. Given that the licensee was able to definitively determine when
- the valves became inoperable, i.e., at the time that they were opened to begin the testing on August 28, the licensee cuncluded that operations were conducted within the plant technical specifications at all times. This is because with both valves inoperable and the penetration not isolated, the plant technical specifications permitted continued operation such that an orderly shutdown be initiated and the reactor be placed in the cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The licensee had initiated an orderly shutdown, but restored valve RV-39 to an operable and closed condition prior to completing the plant shutdown. Therefore, the licensee determined that during this event, operations were always conducted within the plant technical specifications.
The inspector also observed that the licensee subsequently had maintained this
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containment penetration line closed since the August 28 event. As stated in inspection report 54271/97-06, the licensee was unable to demonstrate that valve, RV-40, was fully closed, making the valve inoperable. Technical specification 3.7.D.2 requires that in the event any specified isolation valve becomes inoperable,
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reactor power operation may continue provided at least one valve in the affected line is in the mode corresponding to the isolated condition. Subsequent to cycling
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valve RV-39 opened and closed, the licensee demonstrated by test that valve, RV-l 39, was in the isolated (closed) condition. The licensee removed power to the valve operator and routinely monitored the valve position indication to ensure that it remained closed throughout the ensuing operations prior to the plant shutdown during the current refueling outage. The inspector concluded that the licensee appropriately followed the plant technical specifications for this degraded containment penetration.
During the current refueling outage the licensee found that valve, RV-40, (thought to be inoperable by not closing properly during the August 1997 testing) was actually in the closed position. The valve was disassembled for inspection, and
,
some minor flaws in the seating surfaces were identified and corrected. In addition, I
as a result of engineering reevaluation, the valve closure spring was adjusted to i
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provide greater closing pressure. VY's root cause determination for the failure of l
valve, RV-40, was still in process at the close of the inspection period; however, l
their preliminary assessment was that the root cause of the event was inadequate
. instructions for adjustment and calibration of air operated valve actuators. As part I
of the corrective action, electrical design engineering was in the process of reviewing the calculations for operational settings of all air operated valves (projected to be complete by the end of the year).
!
Additionally, the licensee determined that the acceptance criteria that they were using for determining that valves RV-39 and RV-40 properly stroke closed during testing could have been improved. Previously, the licensee had baen using a flow isolation test as part of the stroke closed surveillance testing. Minor seat leakage during the testing when the valves closed masked the ability of the operators to l
determine if the valve had fully stroked closed. The as-found condition of valve RV-l 40, showed this. As a result, the licensee modifsd both valves during the current refueling outage, installing new valve position indicators on the valve stems to improve the operator's ability to verify valve position. Also, the inspector verified that Procedure No. OP 4110, Rev 28, " Reactor Recirc System Surveillance," was
!
issued on April 10,1998, incorporating new acceptance criteria to use the newly installed stem position indicators in verifying the closure timing of these valves.
The inspector concluded that the licensee fo!! owed the plant technical specifications for this degraded primary containment penetration. Also, the licensee provided reasonable basis for the event retraction of the " seriously degraded condition" event
,
l notification. Finally, the licensee's modification to the valve position indicators and to the test procedure acceptance criteria for verifying valve closure was deemed a reasonable corrective action to ensure that the test would demonstrate operability for the containment isolation function.
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1. Engineering E2 Engineering Support of Facilities and Equipment E2.1 Msin Station Batterv Operability Followina Fault a.
insoection Scoce (92903)
The inspectors reviewed the licensee's evaluation of the effect of the fault on the
"A" main station battery following the performance test discharge, to assess the i
licensee's technical review of the of the viability of the battery to support the safety related functions.
b.
Observations and Findinas The inspectors observed that the licensee had contacted the battery manufacturer to obtain the maximum estimated fault capability of a main station battery, consisting of 60 type LC-31 cells, following a full performance test discharge. The
. manufacturer estimated a potential fault current of 6000 amps at the terminals of -
the discharged battery. The manufacturer also indicated that a similar battery had successfully passed a short circuit test at 15,000 amps. Following the inspector's questioning of the time duration of the manufacturer's short circuit test compared to the VY fault duration of six to twelve minutes, the licensee calculated the current at the location of the fault at the remote end of the test cables. The licensee calculated a fault current of less than 3000 amps. The licensee's calculation neglected the additional resistance at the negative terminal of the battery when the electricians loosened the connection. Heat generated at this loosened connection because of the 12R energy was sufficient to start a fire at the terminated cable, indicating a relatively high value of resistance which would have additionally reduced the fault current. The inspectors confirmed that the manufacturer's published rating for the LC-31 cell five minute discharge was 250 amps per positive plate. The LC-31 cell has 15 positive plates (or 3750 amps).
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c.
Conclusions VY's initialinvestigation of the effect of the fault on the "A" main station battery was not of sufficient depth to prove operability.
Following discussions with the inspector, VY obtained additional information from the battery manufacturer and concluded that the battery had not been damaged as a result of the event. The inspector determined that this conclusion was adequately founded.
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e
13 2.2 (Undate) Insoector Fo!!ow-Un item 98-11-02: Main Steam Tunnel Blowout Paagl Actuation Setooint Variance a.
Insoection Scone (92903)
,
in December 1996, VY determined that the existing arrangemerit of blowout panels
)
in the main steam tunnel may cause the environmental qualification temperature limit for equipment in the reactor building to be exceeded during some high energy line break (HELB) scenarios. A basis for maintaining operation, BMO 96-18, " Main Steam Tunnel Blowout Panel," was developed to verify that continued operation was acceptable, and to specify corrective actions and completion dates. All corrective actions for this BMO were scheduled to be completed prior to startup from the 1998 refueling outage.
During this inspection, the inspector examined the configuration of two blowout panels in the main steam tunnel.
b.
Observations and Findinas
.The two panels are of the same design and are supposed to relieve to the turbine -
building at a positive pressure (+) of 0.25 psiin the steam tunnel. To achieve this set point, a weight is suspended from the panel through a pulley arrangement. The inspector noted that the construction of the weight and pulley arrangements did not -
appear to be consistent with that used for safety class components. Specifically,
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attachment of the pulleys was questionable; the eye of the pulley was passed through (what appeared to be) a flame-cut hole in a metal wall beam, and secured by a large nail that was tac-welded to the beam; and, the weights were coarsely cast concrete and did not appear to be a precision weight. In addition to these material concerns, the inspector questioned VY concerning how the set pressure
!
precision and tolerance of the blowout panels had been established and verified.
VY responded by generating an event report. When this report was reviewed by the management screening committee, it was recognized that changes to the design of the blowout panels would affect the resolution of BMO 96-18. As a result, increased management attention was focused on resolution of BMO 96-18 (and, consequently, on the blowout panel issues discussed above).
'
Subsequent examination of the reactor building pressure relieving devices identified
a problem with the reactor building blowout panels on the refueling floor. These l
panels were designed to relieve at +0.25 psi, based on the size of the panel and the i
. failure strength of the rivets that hold the panelin place. However, VY determined that variance in the failure strength of the rivets had not been accounted for in j
determining this value. When this variance was included, the actual blowout
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~ pressure was calculated to be +0.47 psi. Changing this.value, in tum, affected the reactor building temperature and pressure profiles associated with HELBs. On April 27, VY reported this determination to the NRC in accordance with applicable regulations.
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c.
Conclusions At the close of the inspection period, VY was developing a modification to upgrade the two blowout panels in the main steam tunnel, to be completed during the current outage. Engineering actions to resolve the remaining issues associated with BM3 96-18 and the reactor building HELB concerns were continuing. VY's resolution of these issues to support startup from the current refueling outage will be examined as part of the continuing inspector follow-up of this item.
E2.3 Inadeouate Installation Testino identified for Thermal Overpressure Relief Valves a.
Insoection Scoce (37551)
During a review of the procedure for non-Appendix J valve leak rate testing (OP-4028), VY determined that a problem existed with testing that had been performed on two valves that were installed during the 1996 refueling outage. The inspector reviewed a sample of the licensee actions regarding this problem.
b.
Observations and Findinas The valves at issue were the thermal relief check valves associated with containment penetrations for the low pressure coolant injection (LPCI) system shutdown cooling suction line (V1018A), and the recirculation system sample line (V2-39AA). These valves were installed for protection of piping within the containment penetrations, which could be subject to post-LOCA over-pressurization due to thermal expansion of water trapped between the inner and outer containment isolation valves. The problem that was identified was that the system alignment for the forward flow test, as established by the procedure, provided a flow path that bypassed the valve being tested. Therefore, the results were inconclusive in demonstrating that the check valves would actually open. For the valves at issue, this test had been performed as acceptance testing, as well as for in-service testing.
The valves had been satisfactorily tested in the closed (containment isolation)
direction.
In response to this finding, VY declared valves V10-18A and V2-39AA inoperable for flow in the forward direction, but operable for the purpose of containment isolation. The associated containment penetrations were assessed to be operable, as had been the conclusion of the BMO that addressed the original issue of thermally induced post-LOCA over-pressurization of piping in containment penetrations (prior to installation of these valves). The valves were subsequently tested satisfactorily during the current refueling outage.
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c.
Conclusions Thorough licensee review of an in-service test procedure identified that two thermal relief check valves had not been adequately tested in the forward direction during the 1996 refueling outage. Immediate operability determinations were prompt and
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i adequately founded. Although the inadequate test procedure constituted a violation
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' of NRC requirements, this non-repetitive, licensee-identified and corrected violation i
is being treated as a non-cited violation, consistent with section Vll.B.1 of the NRC
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Enforcement Policy. (NCV 50-271/98-04-03)
E2.4 (Closed) Insoector Follow-Un item 96-11-01: Emeroency Diesel Generator Tornado Protection a.
Inspection Scone (92903)
This inspection was to confirm that the permanent design modifications and hardware installation had been completed to address the issue of building pressure relief design for the emergency diesel generator (EDG) and fuel oil day tank rooms.
Also, this inspection was to confirm that the related issue of the turbine building high energy line break (HELB) pressure relieving concern had been resolved. The inspector reviewed the design basis documentation for the damper and blowout panel fabrication and installation. The documentation consisted of Engineering Design Change Request (EDCR)97-409, " Diesel Generator and Daytank Rooms,.
!
-Tornado Modifications," and EDCR 97-419, " Turbine Building Blowout Panels"
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(which included the applicable safety evaluations 97005 and 97-33).
j b.
Observations and Findinas j
i The inspector performed a w'alkdown of the EDG fuel oil / day tank rooms and-l verified that the modifications were in place at the specified locations. ~ Structural modifications of the block walls between the fuel oil / day tank rooms and the EDG j
rooms were completed by the addition of steel reinforcement where the block ' walls j
intersected the EDG rooms. An exhaust damper was permanently installed in the j
ceiling of each of the EDG rooms. A damper functional test had been performed at the vendor manufacturing facility and was witnessed by a VY quality assurance l-representative. After installation, a free movement test was performed with satisfactory results. The inspector verified that the door sills at the entrances to the
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two fuel oil / day tank rooms had been removed providing a gap of 3/4" maximum.
This dimension was'specified in the design document for venting purposes.
The inspector performed a walkdown of the turbine building to determine if the l '
specified number of blowout panels (three) had been installed in the designated location. The inspector observed that three blowout panels had been installed on the east wall, north end of the turbine building, at a level which was well above the operating floor.
No deficiencies were noted during the document review and the verification
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walkdown for installed hardware.
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Conclusion
.The inspector verified that the specified hardware had been permanently installed in the designated locations in the turbine building, EDG and fuel oil / day tank rooms.
Modifications to the entrances of the fuel oil /daytank rooms had been made and were verified to be as specified in the design change documents. Therefore, inspector follow-up item IFl 98-11-01 is closed.
E8 Miscellaneous Engineering issues E8.1 (Closed) Insoector Follow-Un item 97-02-06-Use of main station batterv standbv charaer " CAB" (92903)
'This licensee-identified issue concerned the use of spare battery charger, CAB, with the Division ll emergency electrical system 125VDC battery. Each of the two-(Divisions I and 11) 125VDC main station batteries is supplied by a dedicated battery charger which receives power from the AC portion of its associated division.
Battery charger. CAB receives AC power from a Division I source, but can supply
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-either of the main station batteries. If the system were being operated with CAB
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supplying the Division ll battery, loss of Division i AC power would result in eventual loss of both division DC systems due to the inability to recharge either battery.
However, technical specification 3.10.A.2.b identified this configuration as being acceptable for continuous use.
in response to this finding, VY instituted administrative controls to restrict the use of battery charger CAB with the Division il battery. VY initiatJ a technical specification change request to eliminate use of CAB as an alternative to use of.
either of the dedicated battery chargers. This change was approved and issued as Amendment No.153 to the facility operating license, dated March 5,1998.
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~Accordingly, inspector follow-up item IFl 97-02-06 is closed.
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l IV. Plant Support
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R1 Radiological Protection and Chemistry (RP&C) Controls R1.1 Refuelino Outane Radioloalcal Controls (Proaram Changes)
a.
Insoection Scone (83750)
The inspector reviewed selected radiological controls program changes since the previous inspection in this area. Areas reviewed included organization and staffing, facilities and equipment,' and procedure changes.
' b.
Observations and Findinas
The inspector observed that a senior radiation protection technician was providing
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essentially all of the coordination and control for activities involving initiation' of
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L cutting of torus downcomers on March 31,1998. The inspector noted that the L
- cutting of torus downcomers was a major task involving significant challenge to radiological controls. Notwithstanding, the technician was not provided with
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defined responsibilities or authorities for performing this function. -The inspector observed weakness in radiological controls briefings, inter and intra-departmental communications, and the recognition and resolution of industrial safety issues associated with the work. It was also apparent that, with respect to the amount of l
radiological challenging work in progress to support the torus modification, there was insufficient radiation protection personnel available to provide sufficient job t
coverage at the work location.
.Upon conveying concerns to the licensee's management representative, the licensee suspended the wort activities in the area pending the implementation of improvements in the management oversight and radiological and industrial safety controls for this activity. Section R8.3 of this report also pertains.
t c.
Conclusion l'
- -The licensee was ineffective in establishing sufficient and positive radiological control technician coverage of significant work involving torus modifications as evidenced by deficient briefing of affected personnel, and insufficient technician e
resources to cover significant radiological work in-progress. Additionally, industrial safety issues involved in the work were not immediately recognized and addressed
,
~ by the licensee until brought to management's attention by the inspector. Upon notification, the licensee suspended work activities pending improvements in radiological control coverage, and took action to address industrial safety. No violations were identified.
I R1.2 Refuelino Outaan Radiological Controls (ALARA Plannina and Preparation)
a.
Insoection Scone (83750)
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The inspector selectively reviewed the ALARA planning and preparation for the refueling outage. Additionally, the inspector reviewed radiological control records, interviewed licensee representatives relative to outage planning, and observed activities to determined the effectiveness of planning, preparations, and management oversight for radiologically challenging work activities. The inspector reviewed certain work activities that had the potential for creating radiological
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. hazards (e.g., refueling activities, reactor water cleanup valve work, and control rod drive removal).
b.
Observatbna and Findinas
.The licensees planning and control was such that there was significant delay in initiating torus work. The contractor assigned to perform the major work activities in the torus scheduled and person-loaded the workdays prior to the outage and i
during the start of the outage in anticipation of starting the planned torus work. As
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a consequence of the delay, the inspector determined that workers were placed in non-radiation work activities preparing for the major work activities (e.g., shop preparation and mock-up training) or used in limited radiation exposure work activities (e.g., unloading trailers within the turbine building, and wrapping and packing equipment for transport into the reactor building). The resulting accrued dose of 1.075 rem was sustained to support about 2191 hours0.0254 days <br />0.609 hours <br />0.00362 weeks <br />8.336755e-4 months <br /> of work, principally within the turbine building. The accrued person-rem was approximately twice that originelly estimated and the man-hours expended were approximately four times that originally planned for preparation work.
The licensee's ALARA Engineer informed the inspector that the workers were working in an area that did not require a specific RWP and that the ALARA group had not yet received final man-hour estimates for the planned torus tasks, including pre-work mobilization. The ALARA Engineer elected to issue a specific RWP for the mobilization to specifically track accrued exposure for the activities. The individual
indicated that the ALARA group had made a rough estimate of personnel exposure and man-hours and were closely monitoring accrued personnel exposure.
cThe ALARA Engineer reviewed the accumulated exposure for the work on a daily
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basis and periodically toured the turbine work areas. The ALARA Engineer did not -
identify any personnel receiving unnecessary radiation exposure.
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.The inspector's review of ALARA planning indicated generally good efforts on tasks.
reviewed based on the following:
The licsnsee supplemented the ALARA staff (ALARA Engineer) with six
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contractor ALARA personnel to preform pre-outage ALARA reviews. The '
inspector also observed that ALARA personnel were performing in field i
ALARA audits to identify areas for exposure reduction. The inspector noted
. that a comprehensive review of areas for exposure reduction was performed at the start of the torus modification work.
The inspector's review of completed work activities (e.g., control rod drive j-
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. work) indicated the work was completed within expected ALARA goals.
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Extensive training was accomplished with mock-up to training and qualify
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~ welders in properation for the torus strainer modification.
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L Notwithstanding, some areas for improvement, relative to ALARA, were noted as:
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The shielding for the drywell was not optimized. The inspector observed
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open, unshielded locations where ongoing work was to occur. The licensee indicated permanent shielding for these areas was under consideration.
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ALARA recor mandations from pervious outages were not tracked or incorporated the most recent outage planning work packages. The licensee staff interviewed were uncertain if any previous recommendations were u_:---_____-__-____
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implemented. The inspector observed that during the previous refueling outage in 1996, a 69 percent increase in work orders was generated after the start of the outage (i.e.,266 orders were in place at the start of the outage, but 451 orders had been accomplished by the end of the outage).
Consequently, many tasks were accomplished with limited time for ALARA
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evaluation and no analysis was accomplished relative to accrued occupational exposure associated with the increase of unplanned or unscheduled work.
)
The inspector observed that the ALARA planning checklist used for ALARA
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task review was limited in scope.
The ALARA group was provided with an outage man-hour estimate only
about one month prior to the start of the outage, leaving little time to determine the expected cost of the outage relative to personnel exposure, or re-assess exposure reduction initiatives.
The following additional observations were made:
The licensee completed the design change package and approval for torus
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work several weeks before the start of the outage but did not complete the installation and test procedures until about March 27,1998 (i.e., about one week after the start of the outage).
The inspector observed that workers had been directed to suit up in
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protective clothing to wait for entry into the torus. Tne inspector noted this
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potentially created heat stress problems for the workers. Subsequently, the licensee determined that this practice was inconsistent with the lessons-l learned from previous heat stress events. Subsequently, the licensee took actions to improve this condition. Section R8.3 of this report also pertains.
c.
Conclusions i
Overall, the licensee implemented an ALARA program that met the requirements of L
the regulations. Notw!!hsianding, ALARA planning and preparation activities were limited in scope and areas for improvement were noted.
i R1.3 Refuelino Outaan Radiological Controls (Internal and External Exoosure Controls)
.
- a.
Inanection Scone (83750)
The inspector selectively oxamined the internal and external exposure control program. The inspector reviewed records, interviewed cognizant licensee personnel, l
and observed exposure control practices during work activities and tours of the
> RCA.~The inspector reviewed high radiation area controls, general radiological posting, implementation of the radiation work permit program, implementation of the dosimetry program, and exposure control practices. The inspector toured the drywell, torus, refueling floor, and reactor building and observed ongoing activities
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and radiological conditions. The inspector selectively made independent radiation measurements to verify licensee results.
b.
Observations and Findinas Widespread use of lapel air sampling to sample air in workers' breathing zones was being effectively used. Engineering controls were used to reduce ambient airborne radioactivity concentrations. Radiation protection personnel closely monitored activities with the potential to generate airborne radioactivity. The licensee used respiratory protective equipment, principally to limit potential intab of lead. The
observed individuals were adequately trained and qualified in the use of the respiratory protective equipment.
The inspector reviewed airborne radioactivity sampling in the torus during tours, downcomer cutting, and during turbine blade cleaning activities. The inspector noted that the licensee provided for use of breathing zone air sampling and air sampling using both low volume and high volume airborne radioactivity sampling equipment. No significant airborne radioactivity was identified and no individuals sustained any significant airborne radioactivity intake.
The inspector noted that in general, radiological areas (e.g., high radiation areas, radiation areas) were properly posted and locked. In general workers were provided
. briefings as required by applicable radiation work permits and 10 CFR 19.12, and were observed to be generally wearing dosimetry as prescribed. During in-field observations of ongoing work, the inspector noted workers to be sensitive to unplanned radiation exposure.
The following observations were made:
The licensee implemented use of a single air sampling head to sample j
workers' breathing zones for airborne radioactivity, lead, and total metals.
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However, the licensee was not able to provide information as to the l
adequacy and effectiveness (e.g., collection efficiency) of the various i
contaminants being measured, including airborne radioactivity. The licensee indicated the sampler was being used in conjunction with other conventional air sampling equipment. The licensee suspended use of the single sample j
head and initiated use of the previously used contaminant sampling head j
pending confirmation of the adequacy of the sampling head. The inspector will follow-up on the licensee's use and evaluation of the sample head in a subsequent inspection. (IFl 50-271/98-04-04)
j The inspector verified by checking radiation work permit (RWP) computer l
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access information that workers were properly signed in on their assigned RWP. The inspector did however identify one worker, who signed in on RWP
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No. 98-00063 on March 23,1998, for work in the turbine building, who entered the reactor building for approximately one-half hour. The RWP did
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not specify work within the reactor building. The licensee reviewed this matter and did not identify any procedure adherence conc 6.ns.
The fo!!owing licensee-identified violation was determined relative to high radiation area boundary controls:
The licensee identified four instances were high radiatiors are boundaries
were compromised between February 25,1998, and March 20,1998. The boundary problems principally invc,1ved a special design swing arm gate used by the licensee to provide ease of access and closure of the gate upon entry to non-locked high radiation area. Each instance was identified by radiation protection personnel, immediately corrected, and an Event Report was issued. The licensee took short term corrective actions for the first three events which occurred in close proximity in time to each other. The actions included informing workers of the events and concerns via the station newspaper, conduct of radiation protection discussions at worker safety i
l meetings, and enhancement of training for incoming contractor personnel supporting the outage. When the fourth event was discovered on March 30, 1998, increased actiun was taken. These included a requirement, via issuance of a radiation protection standing order, requiring radiation protection personnel to escort workers to high radiation area boundaries and discuss boundary controls for entry and exit, including review of as found boundary conditions. In addition, radiation protection initiated reviews and enhancement of swing boundary gates, including replacement of gates as appropriate.
The inspector noted that Technical Specification 6.5 B.1. requires that high radiation areas (i.e., areas exhibiting radiation dose rates preater than l
100 mrem /hr at 30 cm) be barricaded and conspicuously posted as a high
!.
radiation area, and that the four instances of high radiation areas boundaries l
not being barricaded and conspicuously posted between February 25,1998, l
and March 20,1998, was a violation of Technical Specification 6.5.B.1.
l Notwithstanding this, the conditions were identified by the licensee all within a short time; reasonable corrective actions were taken in a prompt mannor; i
l the as found conditions did not appear to be willful; and, the conditions were l
not due to a lack of management oversight. This non-repetitive, licensee l
identified and correctsd violation is oeing treated as a Non-Cited Violation l
consistent with Section Vll.B.1 of the NRC Enforcement Policy. (NCV 50-271/98-04-05)
!
c.
Conclusions Applied radiological controls for ongoing work activities were generally well
. implemented. The licensee implemented generally effective external and internal exposure control programs.
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j.
A non-cited violation was determined based on a licensee-identified condition involving high radiation areas that were left with barricades and postings not fully engaged in accordance with the license requirements (i.e., Technical Specification
!
6.5.B.1). Corrective actions were initiated that appeared appropriate for the circumstance.
R1.4 Refuelino Outana Radialaaical Controls (Control of Radioactive Materials and Contamination)
.a.
Inanection Scone 183750)
The inspector selectively reviewed radioactive material and contamination control practices, including: the adequacy of supply, maintenance, calibration and performance checks of survey and monitoring instruments; the use of personal contamination monitors and friskers; and application of hot particle contamination
.
monitoring.
b.
Observations and Findinas The inspector's tours of the station, including the drywell and torus, indicated that the licensee implemented generally effective radioactive material and contamination l
control practices. Radioactive material was properly labeled and stored.
p Contaminated areas were posted and barricaded. Contamination monitoring
- equipment was properly calibrated, field check, and used correctly and effectively l
for monitoring contamination and radiation. Radioactive materials'were properly monitored and controlled.
L c.
Conclusions
!
The licensee implemented an effective radioactive material and contamination control program with respect to observations made during the outage.
!
.
R5 Staff Training and Qualification in Radiation Protection and Chemistry i.
R5.1 Radiolanieml Controls Trainina and qualification L
l a.
Inspection Scone 183750)
The inspector reviewed the training and qualifications of contractor personnel used
>
L to provided radiological control oversight of significant radiological work activities.
Training and qualification of these individuals was evaluated with regard to applicable regulatory, and the licensee's procedural requirements. The inspector reviewed training records, personnel resumes, and interviewed cognizant licensee personnel. Additionally, the verified radiation workers training relative to the requirements of 10 CFR 19.12.
f l
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I
The inspector observed briefing practices and questioned workers in the radiological controlled area as to their knowledge of ambient radiological conditions in their work sareas...The inspector also selectively reviewed training of radiation professional and
<a m
supervisor rtaff.
b.
Observations and Findinas l
Radiological controls personnel were qualified in accordance with applicable l
requirements. The licensee implemented a defined training and qualification program for contracted radiological controls personnel who were responsible for providing radiological oversight during the outage.
u The inspector observed a radiological controls briefing for workers who were to enter the torus and perform initial cutting of downcomers. The briefing was held in l
a very noisy area that was not conducive to effective communication of important l~
radiological control information.
,
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l In response to this observation, the licensee reconvened the briefing in a less noisy
'
_ location. The briefing was determined to be effective in the communication of
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p important radiation protection information to the affected personnel. Section R8.3 l
of this report also pertains.
- c.
Conclusions i.
L The licensee established, implemented, and maintained an effective program fer the i
training and qualifications of contractor personnel responsible for the implementation l-of the licensee radiation protection program.
L R7 Quality Assurance in RP&C Activities a.
Insnaction Scone (831501 The inspector selectively reviewed quality assurance activities including, as
,
'
appropnate, audits, surveillance, and self-assessment activities.
I
!
b.
Observations and Findinas
!
<
l The licensee implemented an active audit, surveillance, and self-assessment
'
I program. A specific audit plan was developed and implemented for outage radiological controls activities. A departmental in-field review program was implemented. Corrective actiom were taken as appropriate, h
The inspector selectively reviewed the audit plan and checklist for the current p
~ ongoing radiation protection audit. The check list did not contain any specific references to regulatory requirements or national standards relative to evaluation of-program adequacy. The licensee acknowledge the observation and indicated that
-
the audit program would be reviewed to determine if the observation would result in
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...
- increased effectiveness. This observation does not constitute a violation of NRC regulatory requirements c.
Conclusions
' The licensee implemented an effective audit of ongoing radiological controls activities.
s R8 Miscellaneous issues R8.1 Plant Tour Observations During the inspection, the inspector made various tours of the radiological controlled '
area. The inspector's review indicated generally good housekeeping.
i R8.2 Meeting With Station Management As a result of weakness identified in coordination and control of torus down comer p,
c cutting and apparent industrial safety concerns (as noted in various sections of this -
report), the inspectors met with station management. In response, the management representative took the following actions:
The licensee initiated additional reviews of'the affected work activity,-
concluded management expectations were not being met, and initiated a work stoppage of torus activities on March 31,1998. The licensee also -
identified that corrective actions for previous heat stress concerns had not been' implemented and effected corrective measures.
The licensee met with senior management of the contractor performing torus
work to discuss' work expectations relative to radiological controls and industrial safety.
The licensee took action to improve direct oversight of the contractor by
- establishing more aggressive quality assurance oversight, i
'
I
'
The licensee constituted a special oversight team to increase _and maintain i
oversight and control of the work activity. The team was charged with the
responsibility to provide single points of contact for radiation protection,
'
L operations, and industrial safety issues.
'
j
!
I The licensee took action to scale the pace and number of work activities to a
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scope that was within the ability of available radiation protection resources l
to effectively control.
j
' Additional industrial safety reviews were conducted of tasks within the
torus; and the safety department initiated close monitoring of torus work activities. A requirement was implemented to require a specific job safety
,
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analysis for each major work activity performed in the torus by the contractor.
R8.3 - (closedl Unresolved item 96-11-03: Followup on change to Updated Final Safety Analysis Report (UFSAR) involving control room staff radiation monitoring. This i
unresolved item tracked the licensee's resolution of an outdated description in the
!
l VY UFSAR which stated that the control room staff would be trained in the use of (
portable radiation monitors, and that these monitors would be available in the L
control room. This UFSAR statement pre-dated radiation protection technicians being assigned to shift duties, which entail routine and incident response radiation I
monitoring of the control room and other occupied work spaces. The inspector verified that the RP manager has submitted a UFSAR change request through the l
licensing department to reflect current practices. A documentation review was first
!
conducted to ensure the original UFSAR statements did not reflect any previously l
made licensing commitments. None were found. Therefore, this unresolved item l-did not constitute a violation of regulatory requirements, and is closed.
j S2 Status of Security Facilities and Equipment
- S2.1 Potential Pathway for Unmonitored Vital Area Accans a.
Inanaction Scoon (717501 During a routine inspection on March 31, the inspector noted a potential pathway i
for unmonitored access from the turbine building into the reactor building vital area.
The inspector reviewed VY's immediate and long term corrective actions to address
this finding, b.
Observations and Findings I
The inspector discussed the observation with VY security management. VY indicated that the subject pathway had been previously considered, and had been l
evaluated to be a tortuous pathway, based on 1) the radiological condition during plant operations, 2) the physical configuration on the reactor building side of the pathway, and 3) the large amount of physical force that would be required to remove the existing impediment to access. While the radiological controls on turbine building access provide some assurance against unauthorized access during plant operations (the subject pathway is in a locked high radiation area), this control is relaxed after the plant is shut down. The inspector was concerned that the physical configuration on the reactor building side did not constitute a significant challenge, and that the existing impediment to access could be cleared with far less
,
force than its weight would imply.
j On April 1; the inspector discussed this issue with region-based NRC security inspectors. Following discussions between these inspectors and VY, the potential access pathway was appropriately compensated. As long term corrective action, l
,
,
p
!
l
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'
VY will install a permanent barrier around the subject pathway. Planning for this modification was in progress at the close of the inspection period.
l The VY Physical Security Plan, section 4.3, " Vital Area Barriers," paragraph k, states that, "All entrances to Vital Areas... are locked or guarded..." Therefore, the subject pathway was an apparent violation of the VY Physical Security Plan.
c.
Conclusions A pathway was identified which provided unmonitored access from the turbine building into the reactor building vital area. Following discussion of the issue with region-based NRC security inspectors, appropriate compensatory measures were established, and long term corrective action is under development. The pathway for unmonitored vital area access was a violation of the VY Physical Security Plan.
(VIO 98-04-06)
V. Management Meetings X1 Exit Meeting Summary The resident inspectors met with licensee representatives periodically throughout the inspection and following the conclusion of the inspection on June 3,1998. At that time, the purpose and scope of the inspection were reviewed, and the preliminary findings were presented. The licensee acknowledged the preliminary inspection findings.
X3 Review of Updated Final Safety Analysis Report (UFSAR)
A recent discovery of a licensee operating their facility in a manner contrary to the Updated Final Safety Analysis Report (UFSAR) description highlighted the need for a special focused review that compares plant practices, procedures and/or parameters to the UFSAR description. While performing the inspections discussed in this report, the inspectors reviewed the applicable portions of the UFSAR that related to the area inspected. The inspectors verified that the UFSAR wording was consistent with the observed practices and procedures and/or parameters.
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ITEMS OPENED, CLOSED, AND DISCUSSED
OPEN NCV 98-04-01 Incorrect orientation of a peripheral fuel bundle in the reactor core.
!
VIO 98-04-02 Short circuiting of the "A" main station battery due to procedural non-compliance.
NCV 98-04-03 Inadegaate installation testing of thermal overpressure relief valves.
IFl 98-04-04 Review air sampler collection efficiency.
NCV 98-04-05 Four instances of high radiation areas not being barricaded and conspicuously posted between February 25,1996, and April 1,1998.
VIO f)8-04-06 Unmonitored access pathway into reactor building protected area.
CLOSED IFl 96-11-01 Emergency diesel generator tornado protection.
- IFl 97-02-06 Use of main station battery standby charger " CAB".
NCV 98-04-01 incorrect orientation of a peripheral fuel bundle in the reactor core.
NCV 98-04-03 Inadequate installation testing of thermal overpressure relief valves.
NCV 98-04-05 Four instances of high radiation areas not being barricaded and conspicuously posted between February 25,1998, and April 1,1998.
URI 96-11-03 Follow-up on change to UFSAR to ensure it does not conflict with
,
other licensee commitments.
l DISCUSSED
1 IFl 96-11-02 Main steam tunnel blewout panel actuation setpoint variance.
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i
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PARTIAL LIST OF PERSONS CONTACTED D. Reid, Senior Vice President, Operations G. Maret, Plant Manager F. Helin, Tech. Services Superintendent E. Lindamood, Director of Engineering
,
l K. Bronson, Operations Manager l'
M. Watson, Maintenance Superintendent M. Desilets, Radiation Protection Manager R. Gerdus, Chemistry Manager
' G. Morgan, Security Manager S. Jefferson, Scheduling Manager-Operations J. Laughney, Quality Assurance Supervisor-Duke Engineering J. McCarthy, Radwaste Supervisor l,
' M.- Pietcher, Technical Instructor-Training G. Weyman, Technical Assistant-Chen.istry R. Morrissette, ALARA Engineer
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LIST OF ACRONYMS USED l
. As Low As Reasonably Achievable BMO Basis for Maintair.:ng Operation
"
CFR Code of Federal Regulation CR
' control room -
DAW Dry Radioactive Waste l
DOT Department of Transportation -
EDG emergency diesel generator j.
ER Event Report L
GL Generic Letter HIC High Integrity Container l
HP Health Physics y
HPCI high pressure coolant injection i
IFl Inspector follow-up item
'
IN.
Information Notice LCO Limiting Condition for Operation e
LER-'
Licensee Event Report LPCI low pressure coolant injection l
LLRW Low Level Radioactive Waste l..
' LSA'
Low Specific Activity
-
p MCC motor control center
!
NRC Nuclear Regulatory Commission
!
NNS Non-nuclear safety PCP Process Control Program PORC Plant Operations Review Committee QA Quality Assurance j-RHR residual heat removal RCA
. Radiological Controlled Area RP&C Radiation Protection and Chemistry
.
RW Radioactive Waste i
' Radiation Work Permit TS Technical Specifications
[
l UFSAR Updated Final Safety Analysis Report URI unresolved item
l.
-VY Vermont Yankee I-e i
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