ML20135D364
| ML20135D364 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 02/25/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20135D318 | List: |
| References | |
| 50-346-96-14, NUDOCS 9703050208 | |
| Download: ML20135D364 (26) | |
See also: IR 05000346/1996014
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U. S. NUCLEAR REGULATORY COMMISSION
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Docket No.:
50-346
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License No.:
Report No.:
50-346/96014
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Licensee:
Toledo Edison Company
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Facility:
Davis-Besse Nuclear Power Station
Location:
5503 N. State Route 2
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Oak Harbor, OH 43449
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Dates:
November 26,1996 through January 24,1997, and
February 13,1997
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inspectors:
S. Stasek, Senior Resident inspector
K. Zellers, Resident inspector
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K. Selburg, Radiation Protection Spccialist
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Approved by:
John M. 'Jacobson,' Ch'ief ~ '~
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Reactor Projects Branch 4
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9703050208 970225
ADOCK 05000346
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EXECUTIVE SUMMARY
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Davis-Besse Nuclear Power Station
NRC Inspection Report 50-346/96014
This inspection included aspects of licensee operations, maintenance, engineering, and
plant support. The report covers a seven-week period of resident inspection.
Ooerations
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In general, control room staff were aware of system / equipment status during the
inspection period (Section 01.1). One exception however, was noted and is
discussed further under Maintenance below.
Good teamwork and communications were exhibited by operations, maintenance
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and engineering personnel in response to a November 26 event involving the failure
of a blowout panel shear bolt. However, a decision to wait for a like-for-like
replacement bolt to be delivered frorn a Pennsylvania warehouse appeared non-
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conservative in that an engineering evaluation had authorized two options for easily
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implemented interim repairs (Section 01.2).
Operator response to a fire water underground linebreak was good overall.
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However, control room operators focused on the event to the detriment of normal
plant parameter monitoring for some period of time (Section 01.3).
Engineered safety features and important-to-safety systems reviewed this
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inspection period were found to be in excellent material condition (Section 02.1).
. Maintenance
Correlation of station vent stack radiation monitor operability on operability of the
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control room emergency ventilation system was not recognized during recent
radiation monitor filter changeout maintenance. Additional administrative controls
subsequently implemented (Section M1.2).
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The NRC identified that operators had failed to recognize a containment air cooler
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was inoperable during. surveillance testing on the cooler. . As such, entry into the.
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associated technical specification limiting condition for operation was not
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documented in the control room log. This is considered one example of a violation
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for failure to follow procedures (Section M1.4).
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The NRC identified that a licensed operator performed steps of a surveillance
procedure out of sequence. This is considered a second example of a violation for
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failure to follow procedures (Section M1.4).
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The NRC identified that technicians performed testing on a main turbme generator
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electrohydraulic controllow pressure switch outside of the scope of the surveillance
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procedure. This is considered a third example of a violation for failure to follow
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procedures (Section M1.6).
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The inspectors noted that an auxiliary feedwater flow control valve was potentially
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preconditioned prior to its being stroke timed (Section M1.5).
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Enoineerina
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' The licensee identified that original design calculations prepared to support
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operation of the control room emergency ventilation system were in error. The
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errors would have resulted in control room personnel experiencing a higher than
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anticipated radiation exposure during a loss of coolant accident. The licensee
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subsequently initiated appropriate corrective actions. Because the criteria for
. enforcement discretion specified in Section Vll.B.3 of NUREG-1600, " General
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. Statement of Policy and Procedures for NRC Enforcement Actions," were met, this
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matter is considered a noo<ited violation (Section E1.1).
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A violation of 10 CFR 70.24 was identified in that criticality monitors required by
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the regulations were not installed, nor was an exemption from the requirements
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requested since issuance of the plant operating license (Section E1.2).
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The inspectors questioned the appropriateness of an internal licensee technical
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specification (TS) interpretation involving the shield building and emergency
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ventilation TSs. The NRC Office of Nuclear Reactor Regulation has been requested
to provide a formalinterpretation (Section E2.1).
Plant Suonort
The inspectors observed that overall, plant personnel conducted their activities in
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conformance with the radiation protection and security program requirements
(Sections R2, S2).
An individual designated as a fire brigade captain appeared to not meet qualification
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requirements for that position (Section F1.1).
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Report Details
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Summarv of Plant Status
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The unit operated at nominally full powe throughout the inspection period.
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1. Ooerations
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Conduct of Operations
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01.1 General Comments (71707)
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Using Inspection Procedure 71707, the inspectors conducted ongoing
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reviews of plant operating activities. Performance of operators in the control
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room were observed, control room panels routinely walked down, logs
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reviewed, and discussions held with control room personnel. In general, the
inspectors noted good equipment material condition, and good operator
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awareness of plant operating status. Although good teamwork was
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exhibited during response to a November 26 blowout panel shear bolt failure,
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a decision to wait for a like-for-like replacement bolt to be delivered from the
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licensee's warehouse in Pennsylvania appeared to be non-conservative.
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01.2 Blowout Panel Shear Bolt Failure
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Inspection Scone (71707)
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On November 26,1996, during performance of maintenance support activities, .
plant personnelinadvertently failed a shear bolt associated with a blowout panelin
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the #4 mechanical penetration room (MPR). The panel was designed to release in
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the event of a high energy line break. The inspectors observed licensee followup
actions to the event.
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b.
Observations and Findinos
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The shear bolt was broken when it was apparently impacted by a ladder during
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maintenance activities. The worker who caused the break immediately contacted
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t% control room. When notified, control room operators invoked Technical
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Specification (TS) 3.6.5.2, " Shield Building Integrity," as directed by an internal TS
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interpretation (TSIR 89-0007 (revision 0)). The action statement associated with
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TS 3.6.5.2 allowed for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with an inoperable blowout panel prior to
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requiring a plant shutdown.
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The licensee contacted their offsite warehouse (in Pennsylvania) to immediately
ship a replacement shear bolt, as a like-for-like replacement was not available
onsite. Subsequently, engineer;ng determined by analysis that a standard bolt or a
larger capacity shear bolt, both available onsite, could serve as an acceptable
alternative. However, since operatior's had determined that a 24-hour action
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statement applied, an initial decision was made to wait several hours for the correct
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bolt to arrive. Eventually, following discussions with the NRC, a decision was made
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to install a standard bolt in the blowout panel while awaiting delivery of the
replacement shear bolt. During the installation of the standard bolt, the
replacement shear bolt arrived onsite. Installation activities for the standard bolt
were terminated, and the like-for-like replacement shear bolt was subsequently
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Installed.
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From the time that the bolt was broken, until the replacement was installed, the
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panel had the potential to release under a LOCA condition. Release of this panel
during a LOCA could have csmpromised the effectiveness of the emergency
ventilation system. This matter is further discussed in Section E2.1 of this report.
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Conclusions
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The inspectors noted that overall, the operations, maintenance and engineering
departments exhibited good teamwork and communications during response to the
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event. However, the decision to wait for the like-for-like replacement to be
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delivered appeared to be non-conservative. Engineering had determined that two
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other easily implemented options were available to temporarily repair the blowout
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panel in the interim but installation was delayed for several hours.
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01.3 Firewater Line Break
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Insoection Scone (71707)
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On December 18, a fire alarm was received in the control room concurrent with an
automatic start of the electric fire pump. The inspectors observed the response by
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control room operators.
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b.
Observations and Findinos
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The inspectors noted operator response to be both appropriate and timely. Good
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evaluation of control room indication and alarms was evidenced. Operators were
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able to ascertain quickly that the alarms / indications were a result of underground -
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boring operations that were being conducted in the owner controlled area.
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The boring operations were being conducted in support of a new water line
connection to the Carroll Township supply line that was being installed at the time,
it was subsequently determined that a Davis-Besse underground fire water line was
inadvertently hit by the boring machine outside of the Davis-Besse training center.
Appropriate portions of the fire water branch lines were isolated, thereby securing
the leak path.
The inspectors identified one apparent weakness during the event. During initial
followup actions, all control room personnel were primarily focused on the event.
' As a result, normal operator monitoring of plant parameters decreased for some
period of t'me. When this matter was subsequently discussed with the shift
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supervisor, he indicated that this situation was also recognized by control room
supervision during the event and actions had been taken to regain the expected
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level of monitoring. No other concems were noted during inspector review of this
matter.
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Conclusions
Overall operator response to the fire water line break was good. Good
interdepartmental communications and coordination ware noted. However one
weakness was noted wherein operations personnel became overly focused on the
event and normal monitoring of plant parameters were reduced for some period of
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time.
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Operational Status of Facilities and Equipment
02.1
Enaineered Safetv Features System Walkdowns (71707)
The inspectors used inspection procedure 71707 to walkdown the accessible
portions of the following engineering safety features (ESF) and important-to-safety
systems:
emergency diesel generators 1 and 2
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station blackout diesel generator
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auxiliary feedwater - trains 1 and 2
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high pressure injection system - trains 1 and 2
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low pressure injection system - trains 1 and 2
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containment spray system - trains 1 and 2
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motor driven feed pump
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emergency instrument air compressor
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Equipment material condition was found to be acceptable in all cases. Major
flowpaths were determined to be consistent with the updated safety analysis report
(USAR). Pump / motor fluid levels were within their specified bands. Local and
remote controllers were correctly positioned and required instrumentation appeared
functional. Auxiliary equipment necessary for system operability was also
functioning properly. Few oil and fluid leaks were noted during the walkdowns. In
general, housekeeping was found to be acceptable. The inspectors identified no
substantive operability concerns as a result of the walkdowns.
However, the inspectors noted two issues during an auxiliary building walkdown on
December 18.
A ladder was noted in the #2 emergency core cooling system (ECCS) pump
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room. The ladder had been erected on February 27,1996, so that operators
would have access to an overhead valve CC 174 during monthly position
verification checks. The licensee subsequently determined that the ladder
had been placed without appropriate engineering evaluation. Subsequent
evaluation at the time indicated that there was not a concern with the
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specifically noted installation. However because there was no clear
programmatic controls governing installation of the ladder, the licensee
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elected to remove the ladder.
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A one gallon can of thermoncement was stored in the auxiliary buiWog
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behind a cabinet adjacent to the #2 ECCS pump room. When notified,
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licensee management indicated that the can was not stored per management
expectations or station standards. The thermoncement was subsequently
removed, and the appropriate individuals counselled by maintenance
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management.
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Miscellaneous Operations issues (92901)
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-08.1 (Open) Unresolved item (50-346/96005-01(DRP)): Senior reactor operator (SRO)
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proficiency watchstanding requirements for shift managers. This matter addressed
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a concern that the plant guidelines for how shift managers performed SRO
proficiency watchstanding requirements were vague.
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The original concem was resolved following issuance of a memorandum from
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operations management that more specifically delineated proficiency watchstanding
requirements for shift managers. The information presented in the memorandum
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was then to be incorporated into appropriate training documentation.
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During this inspection period, the inspector noted that two individuals were
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simultaneously standing a proficiency watch as the assistant shift supervisor
(control room SRO). The inspector questioned whether both individuals could
sufficiently perform necessary actions during the shift to maintain proficiency
(including directing operator actions) to meet the intent of 10 CFR 55.53. The
inspector contacted NRC Region 111 operator Jcensing personnel who indicated that
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the licensee's placement of two operators onshift fulfilling the same proficiency
watch requirements appeared inappropriate.
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The licensee subsequently modified the proficiency watchstanding guidance to
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prohibit simultaneous standing of proficiency watch stations. The requirements of
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10 CFR 55.53 do not specify a means as to how to implement the subject
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requirements. The method as to how the licensee was meeting those requirements
now appear to be consistent with the inspector's understanding. However, this
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matter will remain open pending inspector review of the updated program guidance.
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08.2 (Closed) Unresolved item (50-346/94011-02(DRP)): Refueling related events. This
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' matter involved a number of events that related to the control of core alteration
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related activities. In response, the licensee strengthened associated procedural
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controls involved with core alterations, counselled the individuals involved in each
of the events, and thereafter conducted training sessions with other personnel who
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could be assigned future core alteration functions. No similar problems were noted
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during the most recent refuel outage.
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08.3 (Closed) Inspection Followuo item (50-346/94017-01(DRP)): Change in SRO shift
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work schedule. This matter involved the modification of SRO shift work schedules
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from an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> shift to a 12-hour shift. Shortly after the change was implemented,
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an issue arose involving short shift turnovers by reactor operators (ROs). This was
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documented in inspection reports 50-346/95008 and 50-346/95009 and resulted in
issuance of an NRC violation. Institution of the new 12-hour shifts had directly
contributed to SROs not providing adequate oversight of the RO turnover process.
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That weakness was subsequently corrected.
During the current inspection period, the inspectors evaluated the SRO 12-hour
work schedule and concluded that it appeared to be functioning sufficiently well to
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meet regulatory requirements. This item is considered closed. However, at the
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conclusion of the inspection period, the inspectors were informed that plant
management was currently in process of reevaluating the need for a 12-hour SRO
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shift work schedule.
08.4 (Closed) Inspection Followun item (50-346/96006-03(DRP)): Functional failure of
EDG lube oil checkvalve cap. The licensee subsequently performed an evaluation of
the event, and determined that the cap had been recently replaced during a
maintenance activity and may very well have failed as a result of that maintenance
activity.~ The cap was subsequently replaced with no further problems noted.
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11. Maintenance
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Conduct of Maintenance
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M1.1 Maintenance Activities (62707)
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The inspectors observed / reviewed all or portions of the following maintenance -
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activities:
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MWO 3-97-1248-01
Remove and replace inlet and outlet ventglass
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Joints on #1 main station exhaust fan C-19-1.
MWO-3-96 5425-01
Anticipatory reactor trip system (ARTS) main
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r.urbine pressure switch channel D snubber
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4eplacement.
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MWO 3 97-1995-01
EDG air compressor and air receiver pressure
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switch calibration
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The inspectors noted that maintenance personnel conducted these maintenance
activities in accordance with schedular requirements, that they received permission
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from operations shift management to proceed with the work, that maintenance
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documentation was at the work site and was found to be of good order. System
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clearance boundaries appeared to be properly established and workers appeared to
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perform their duties in a conscientious, effective manner. Post maintenance
housekeeping conditions were good.
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M1.2 Technical Soecifications LCO Inadvertentiv Entered Durina Routine Maintenance on
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Station Vent Stack Radiafion Monitor
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a.
Insoection Scooe (71707)
On December 12, the licenree identified a condition that was potentially outside the
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design basis of the plant in that station vent radiation monitors, RE4598AA and
RE4598BA, had both been simultaneously out of service. Both radiation detectors
(RE) being inoperable would adversely impact the operability of the control room
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emergency ventilation system (CREVS) as required by Technical Specification
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3.7.6.1. The inspectors verified that licensee followup actions following the event
were appropriate.
b.
Observations and Findinas
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One RE was returned to service following changeout of its filter media. The second
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RE was returned to service sometime later following correction of a check source
problem.
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Output from the subject REs was designed to initiate an isolation of normal control
room ventilation, although that function was not fully recognized at the time both
REs were inoperable. Tectaical Specification surveillance requirement 4.7.6.1.e.2,
in part, tests the specified function which would support operability of the CREVS.
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Once the correlation to CREVS operability was recognized, additional restrictions
were placed on future activities associated with these REs. Additional guidance
was provided to operations personnel as to operability implications associated with
the subject REs.
The licensee submitted licensee event report (LER) 96-010-00 to document this
event. Further NRC review of this matter will be conducted during the evaluation
of the subject LER.
M1.3 Surveillance Activities (61726)
The inspectors observed the following surveillances and noted that they were
generally performed in accordance with currently approved procedures, that
equipment functioned as described in the USAR, and that equipment operators
monitored running equipment for abnormal vibrations, fluid leaks, and lubrication
levels. Locked valve entries were made when required and independent
verifications were performed properly.
Except as noted in Section M1.4, limiting conditions for operation (LCO) wera
properly identified, logged, and tracked. LCO time limits were observed to be
followed on all occasions.
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' The documentation of the performance of surveillances was usually observed to be
without error, however, some lack of attention to detail was observed in the
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documentation of the performance of surveillance DB-SC-03023, "Off Site AC
Sources Lined Up and Available."
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DB-SP-03294 (Rev 02)
Containment Air Cooling Unit 1 Monthly Test and
Service Water Valve Testing (See Section M1.4).
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DB-SP-03161 (Rev 04)
AFW Train 2 Level Control, interlock, and Flow
Transmitter Test (See Section M1.5).
DB-MI 03353 (Rev 01)
Channel Functional Test of Main Feed Pump and
Turbine Hydraulic Oil Trip and Main Turbine Oil Trip
Arts Channel 3 (See Section M1.6).
DB-SC-03070 (Rev 03)
EDG #1 monthly (See Section M1.7).
DB-PF-04729 (Rev 03)
Containment Air Cooling Monitoring Test.
- M1.4 Administrative Ricor Lackina Durina Surveillance Testina
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Insoection Scone (61726)
The inspector observed portions of surveillance DB-SP-03294 (Revision 02),
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" Containment Air Cooling Unit 1 Monthly Test," which included portions of DB-PF-.
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03020 (Revision 01), " Service Water System Quarterly Train 1 Valve Testing," on
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December 30,1996.
b.
Observations and Findinas
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The inspector observed that steps in surveillance procedure DB-SP-03294 were
conducted out of sequence when steps 4.1.7 and 4.1.8 were completed prior to
the completion of steps 4.1.5 and 4.1.6. This observation was communicated to
the operator performing the procedure and to operations management. The
inspector determined that the improper sequencing did not invalidate the test and
did not cause any personnel or equipment safety concerns.
The performance of the aforementioned steps out of sequence was not in
conformance with administrative procedure DB-DP-00013 (Revision 04),
" Surveillance and Periodic Test Program." Section 6.3.7.h of the procedure stated
in part that, " Test prerequisites and procedure steps shall be performed in numerical
sequence ..." Consequently, this is an example of a violation (50-346/96014-
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01a(DRP)) of 10 CFR 50, Appendix B, Criterion V, " Instructions, Procedures, and
Drawings."
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Upon reviewing the unit log the following day, the inspector determined that entry
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into the action statement for Technical Specification 3.6.2.2 had not been
documented when the containment air cooler (CAC) service water manual isolation
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valves were shut during the test. Shutting these valves prevented the CAC from
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being automatically available, and consequently rendered the CAC inoperable. This
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observation was communicated to operations shift management, and the licensee
thereafter generated PCAQR 96-0010.
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Because the service water isolation valves were closed for less than the LCO
allowed outage time, no TS limits were exceeded. - However, the failure to log the
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entry was not in conformance with administrative procedure DB-OP-OOOO5
(Revision 05), " Operator Logs and Rounds." Section 6.2.2.d of the procedure
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stated in part that, "The following are entries which shall be recorded in the Unit
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Log:...d. Entering / Exiting a Technical Specification ' Action Sta;ement". Because this
unit log entry was an administrative requirement that should have been completed
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at the time the surveillance test was performed, this is a second example of a
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violation (50-346/96014-01b(DRP)) of 10 CFR 50, Appendix B, Criterion V,
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" Instructions, Procedures, and Drawings."
The inspector noted that once operations management became aware of the'
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inspector's findings, they took prompt measures to prevent recurrence. This
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included issuance of Operations Standing Order 97-002, " Operability Review of
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Sur.reillance Tests." This standing order reiterated some of the guidance of Generic
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i.etter 91-18, which discussed the resolution of degraded and non-conform;ng
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conditions and operability determinations. The standing order also required that
operations personnel review surveillance test procedures specificalle for opembility
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considerations prior to their performance, and to generate a PCAOR to obtain a
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formal engineering evaluation for any operability question that arose. As of the end
of the inspection period,13 FCAORs had been initiated as a result of this initiative.
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Additionally, plant engineering had initiated actions to review all appropriate
surveillance procedures to assess whether operability information should be
included in those procedures.
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Additionally, the inspector observed during performance of the surveillance, that the
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test leader relayed the start of an action step to field test personnel through a RO
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who did not have the procedure available. He relayed that he was closing SW
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1356 by using other than precise communications jargon. This did not cause the
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test procedure to be compromised and did not prevent the testing activity from
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proceeding, but was observed to be communications that did not meet inspectoi
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and licensee management expectations.
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M1.5 Potential Preconditionina of the Auxiliary Feedwater (AFW) Flow Control Valve Prior
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to Stroke Time Testina.
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Insoection Scooe (61726)
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The inspector observed the performance of DB-SP-03161 (Revision 04), "AFW
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Train 2 Level Control, Interlock and Flow Transmitter Test," on December 19,
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1996.
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Observations and Findinos
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The inspector noted during the performance of this surveillance test, that AF 6451,
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the solenoid operated flow control valve for auxiliary feedpump (AFP) #2, had been
cycled shut and then opened prior to stroking the valve to obtain stroke time data.
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A preliminary review of the status of AF 6451 in the station configuration database
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indicated that it was a quality class (Q), Appendix R, and American Society of
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Mechanical Engineers (ASME) valve. Additionally, the surveillance procedure itself
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referenced TS 4.0.5 which referenced the station ASME testing program.
The AFW plant engineer's initial . 1ponse to the question of possible
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preconditioning of the valve was that the valve was timed for the fail open direction
only and that its standby position was the open position. Therefore, if the valve
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failed to function during an accident condition, feedwater could still be delivered to
the steam generators with the flow rate being controlled by varying the AFP speed.
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Additionally, he added that ASME code testing of control valves was not required
.
j.
and that the timing data was simply being gathered to trend the performance of the
'
valve. However, the test procedure itself listed the timing of the valve as an
'
acceptance criteria to be met in order to pass the surveillance test.
,
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' -
The plant engineer indicated that if the timing of the valve had exceeded the
e
j
acceptance limit, he would have considered the valve inoperable, which would have
[
rendered the AFW train inoperable.
'
i
Since the inspector has not completed review of this matter to determine if a
j
regulatory requirement was violated, and since the licensee was going to determine
'
if the surveillance test could be changed to do the stroke timing without potentially
!
preconditioning the valve, this is an unresolved item (50-346/96014 02(DRP)).
i
.
M1.6 Testino Conducted Outside Scope of Procedure
Ii
a.
Insoection Scone (61726)
3
The inspector observed portions of DB-MI-03353 (Revision 01);" Channel
Functional Testing of Anticipatory Reactor Trip System (ARTS) Channel 3," on
,
!
December 3,1996.
.
b.
Observations and Findinos
,
,
This surveillance tested the trip setpoint of a channel of the main turbine
electronydraulic control (EHC) low pressure trip setpoint input to the ARTS. The
i
test methodology was to isolate the pressure switch, install a manual test pump
i-
fitted with a calibrated test gauge, and cycle the switch using the pump to ensure
!
that the switch tripped within an established band.
i
Step 8.2.3.e of the test procedure stated, " Increase test INPUT Pressure Source to
[
PSL-4535C until Test Gauge reads approximately 375 PSIG." The inspector noted
i
12
a
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_
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1
that the maintenance person in the field was capable of complying with this
1
'
t
specification to within plus or minus ten pounds per square inch gauge (psig). The
.
test leader observed that the switch had not reset at approximately 375 psig, and
j
then directed the field person to increase the pressure to about 450 psig.
t
The switch did not reset, the test was terminated, and the control room was
.
notified that the test could not be completed. Operations shift personnel declared
i
i-
the instrument inoperable and placed it into a trip condition per TS requirements.
The instrument was subsequently repaired, tested, and placed back into service.
j
The inspectors determined, that when the maintenance person in the field
j
pressurized the switch to 375 psig, step 8.2.3.e of the test procedure had been
completed. Consequently, the subsequent raising of the pressure to 450 psig was
an action that was performed outside of the scope of the procedure. This is a third
'
example of a violation (50-346/96014-01c(DRP)) of 10 CFR 50, Appendix B,
{
Criterion V, " Instructions, Procedures, and Drawings".
I
,
'
M1.7 Control Room Ooerator Caused Voltaae Transient Durina EDG Run
!
.
i
i
a.
Insoection Scone (61726)
I
The inspector observed pmtions of surveillance test DB-SC-03070 (Revision 03),
i
" Emergency Diesel Generator Monthly Test," and conducted followup questioning
of control room personnel on December 12,1996.
1
b.
Observations and Findinas
.
i
During the surveillance run of the EDG, significant voltage fluctuations were noted
at the output of the EDG. When the unit main generator voltage regulator control
circuit was switched from automatic to manual voltage control, the EDG voltage
'.
fluctuations were eliminated. Operations personnel had no prior experience with
operating a loaded EDG while the unit voltage regulator was in manual.
,
,'
The load dispatcher thereafter requested that MVAR output be changed
i
incrementally. A RO then made the required change to the voltage regulation
~
j
circuit, which caused a significant downstream voltage transient within the station
'
electrical distribution system. This caused a major oscil!ation to the voltage output
of the still running EDG. The voltage transient was observed by the equipment
i
operator and the plant engineer who was observing the EDG run.
They then called the control room to determine the cause of the transient. It was
only then that control room personnel realized that a small adjustment to the unit
generator voltage regulation circuit would cause a major voltage perturbation to an
EDG synchronized to the grid.
PCAOR 96-1558 was generated to document the occurrence as a potential plant
problem. Subsequently, operations management issued standing order 96-007 to
operations personnel. This standing order specified precautions for manipulating
13
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_._
. . _
_ . . _ . - . . _ _ . _ _ . _ _ _
. . . .
,
,
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i
.the unit main generator voltage regulator circuit with a diesel generator tied to the
grid.
.-
c.
Conclusions Related to Surveillance Testina Activities
4
Overall, equipment functioned according to descriptions provided in the USAR.
,
Except for the failure of the instrument noted in section M1.4, no noteworthy
ll
degradation of related systems, structures or components was observed
s
j
(Section M1.3).
l
.
!
i
"
A reactor operator performed test steps out of sequence. This is an example of a
violation relating to failure to follow procedures (Section M1.4).
,
j '
Operators should have recognized that isolating the service water supply to
{
i
containment air cooler No.1 rendered the containment air cooler inoperable.
I
Consequently, operations personnel failed to log that a TS limiting condition for
operation had been entered. This is a second example of a violation relating to-
3
failure to follow procedures (Section M1.4).
,
An auxiliary feedwater solenoid operated flow control valve may have been subject
i
l
to preconditioning because it was cycled prior to stroke timing. Since the inspector
has not completed review of this matter to determine if a regulatory requirement
-
j
was violated, this is an unresolved item (Section M1.5).
,
[
Technicians performed testing on the main turbine generator EHC low pressure
switch outside of the scope of the surveillance procedure. The inspectors were
i
concerned that conducting surveillance testing outside of the scope of approved
i
procedures may result in not identifying degrading equipment performance. This is
I
j
a third example of a violation relating to failure to follow procedures
(Section M1.6).
4
1
!
Control room operators failed to recognize the effect of adjusting the voltage output
i
of the main generator on the performance of an EDG that was running loaded to the
i
grid. (Section M1.7).
I
i
M4
Maintenance Staff Knowledge and Performance
![
M4.1 Good Questionina Attitude Exhibited by Maintenance Personnel
!
a.
Insoection Scone (37551)
,
4
The inspector conducted inspection activities relating to licensee efforts to
i
determine the cause of the divergence of steam generator #2 startup level
j
transmitter, LTSP9A4, as documented in PCAOR 97-0005.
i
i
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I
14
- .
_ . _ . . . - . . -
-
_ . - ,
,.
_.
- . _ _
-
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-
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4
.
b.
Observations and Findinas
.
PCAOR 97-0005 documented that the output of LTSP9A4, #2 steam generator
startup level transmitter, was exhibiting drift from the other #2 steam generator
startup level transmitters and was 3.8 inches divergent at the time.
The licensee's initial assessment of the root cause of the divergence, based on
similar occurrences in the past, was a partial loss of reference leg level due to
leakage past the equalizing valve. However, subsequent troubleshooting and
information gathering determined this not to be the cause; and the station focused
its resources on the level transmitter as the root cause, which was in containment
in a high dose field.
A good questioning attitude was exhibited by instrumentation and controls
personnel when they questioned the approach the station was taking. This
questioning attitude resulted in instrumentation and controls personnel performing
diagnostic maintenanco activities on portions of the instrument string that were
outside of containment prior to proceeding with further troub:eshooting efforts in
containment. The diagnostic activity in itself resolved the indication divergence
when a high resistance electrical connection in the auxiliary shutdown panel was
manipulated, causing the high resistance condition to clear, and returned the
subject level instrumentation indication to expected levels.
The discovery that the root cause of the divergence problem was outside
containment kept the station from proceeding with troubleshooting efforts inside
containment, preventing additional radiation dose to personnel.
The inspector independently verified by review of station drawings, USAR
descriptions, and by questioning plant engineering personnel, that the station
appropriately determined the safety system and control equipment affected. The
level divergence only affected the automatic flow control circuit for the #1 AFW
pump when it provided flow to the #2 steam generator under #1 steam generator
faulted conditions. No effect on the trip setpoints of any protection circuitry
resulted. The inspector also determined that the station had acted conservatively in
determining that if the level divergence exceeded 10 inches, that the AFW flow
control circuit, and consequently, the #1 AFW train, would have been declared
c.
Conclusions
instrumentation and controls personnel exhibited a good questioning attitude when
they requested authorization to perform a diagnostic check on the portion of the
steam generator startup levelinstrument string that was outside of containment
prior to proceeding with troubleshooting efforts inside containment.
Station personnel properly determined the safety consequence of the divergent
.
steam generator startup level transmitter, and committed to a conservative level
divergence value to declare the AFW train inoperable.
15
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MS
Miscellaneous Maintenance issues (92902)
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M8.1. (Closed) Violation (50-346/94006-01(DRP)): Inadequate foreign material exclusion
[
(FME) controls, inappropriate scaffolding erection, and use'of standing orders in lieu
,
l
of a procedure change in response to the aforementioned items, the licensee
reinforced procedural and implementation controls over the FME program and
3
"
'
scaffolding erection activities, and standing order usage was clarified. No further
i
similar problems have been noted to date.
1
1
!
M8.2 (Closed) Violation (50-346/94008-01(DRP)): Unauthorized manipulation of
l .
emergency diesel generator (EDG) room ventilation. In response, the licensee
1
!
counselled the culpable individuals, and reinforced to the maintenance shops the
'
requirement to have appropriate operations approval prior to manipulation of inplant
!
equipment. No additional similar events have been noted to date.
i
I'
'
M8.3 (Closed) Violation (50-346/95007-03(DRP)): Limit switch housing on motor
1
operated valve (MOV) SW 1399 not properly secured upon work completion. In
'
response, applicable plant maintenance personnel as well as maintenance
supervision were cc.unselled on procedural requirements to complete each work
step prior to signing the associated procedure step off. In addition, maintenance
supervision wa; also reminded as to their responsibility to ensure that equipment
and work sp::ces were retumed to an appropriate post work configuration. No
additional problems have been noted.
M8.4 (Closed) Violation (50-346/95008-04(DRP)): Computer point T413 not included in
the measuring and test equipment (M&TE) progmm. The instrument was
subsequently included in the M&TE program, oppropriately calibrated, and had been
maintained under that program since the date of the event. In addition, the licensee
also performed an extent of condition review to identify other instruments that
potentially could have been similarly affected. A minor issue associated with a
control room temperature sensor was also identified and subsequently corrected.
This matter is considered closed,
3j
1
M8.5 (Closed) Unresolved item (50-346/96002-03(DRP)): Licensee discovery of
scaffolding erected in front of mechanical penetration room blowout panels.
Subsequent engineering evaluation determined that the blowout panels would have
functioned as designed with the scaffolding placed in the as-found condition. As
such, no violation of NRC requirements was identified. The licensee subsequently
moved the scaffolding and reiterated to engineering and maintenance services
personnel the need for clear communication to identify where scaffolding has been
erected to ensure proper evaluations are performed.
16
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lit. Enoineerina
E1
Conduct of Engineering
E1.1
Inadeauste Control Room Emeroency Ventilation System Calculated inleakaae
Limits (37551)
'
a.
Insoection Scooe
s
While performing a detailed review of the Updated Safety Analysis Report (USAR),
l
licensee personnel identified a condition that potentially was outside of the plant's
1
design basis. It appeared that original design calculations prepared to support
design and operation of the control room emergency ventilation system (CREVS)
I
were in error. During the inspection period, the inspectors reviewed the licensee's
followup actions and independently evaluated additional design calculations
generated as a result of this issue.
b.
Observations and Findinas
On October 3,1996, licensee review of the calculations for USAR Figure 9.4-13,
" Relationship of Measured Pressure vs. Leakage Area in the Control Room,"
1
revealed that the total CREVS flowrate of 3,300 cubic feet per minute (cfm) was
erroneously used to determine the positive pressure mode inleakage rate through
>
the control room boundary. The correct value should have been 300 cfm. This
was the amount of makeup flow that would be drawn into the control room from
the outside during system operation in the positive pressure mode with the balance
(3000 cfm) being recirculated within the envelope.
A preliminary calculation subsequently prepared indicated that to achieve the target
1/8-inch of water column (WC) positive pressure in the control room, with an
assumed 300 cfm makeup flow, a maximum aggregate opening in the control room
boundary could not exceed 40 square inches. Previous calculations had indicated
that the maximum allowed aggregate opening in the control room boundary could
be as much as 216 square inches (1.5 square feet). Aggregate opening sizes
greater than 40 square inches had occurred in the past.
A second related issue was subsequently identified on October 22, relating to
operation of CREVS in the recirculation mode, which by original design, was to
occur for approximately the first four days following a loss of coolant accident
(LOCA). Since no outside air was to be drawn into the control room while CREVS
was operated in recirculation mode, there would be no pressurization of the control
room envelope during this time period. Therefore, an early design calculation (and
USAR description) specified an assumed unfiltered inleakage limit of 25 cfm during
this timeframe. In October, the licensee determined that the unfiltered leakrate into
the control room during these first four days could be much greater than 25 cfm.
Specifically, the adjacent ventilation equipment room could be at a positive pressure
with respect to the control ram because the equipment room ventilation supply fan
17
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.
.
"
was not turned off as a result of accident initiation (unless a loss of offsite power
were to occur concurrent with a LOCA).
-
-
Based on the calculated pressure in the equipment room, the resultant maximum
allowable aggregate hole size between the equipment room and the control room
l
,
was less than about three square inches. The total opening between the control
room and the ventilation equipment room had been greater than three square inches
l
during past operation.
i
During the current inspection period, the licensee subsequently reconstructed the
original Bechtel computer code used to determine control room operator post-LOCA
radiation exposures in an attempt to validate the original design calculations. Once
!
,
the code was reconstructed, the appropriate variables were modified to the new
inteakage assumptions. The licensee removed as many conservative assumptions
.
'
as they could to make the resultant calculation as " realistic" as possible.
-
The inspectors independently reviewed the calculations and verified that
assumptions used appeared appropriate, that the calculational methodology used
was in conformance with regulatory requirements and published guidance, and that
the results obtained were translated to necessary CREVS system operational
restrictions. The inspectors noted that the results indicated a radiation exposure to
control room operators of greater than 30 rem thyroid would have occurred within
the 30 days following a postulated LOCA.
The apparent cause for the non-conservative maximum allowable aggregate hole
size for the positive pressure mode was an error in original Bechtel Calculation
"
26.003, Control Room Pressurization System, which was performed in 1973. The
calculation erroneously concluded that CREVS should be sized for a total system
4
makeup capacity of 3300 cim, even though CREVS was limited to 300 cfm of
outside air intake during positive pressure mode.
j
4
in addition, the at parent cause of error in the recirculation mode inleakage limit was
that the size of the allowable opening between the ventilation equipment room and
,
'
the control room had not been previously evaluated for the conditions assumed
when CREVS was operating in recirculation mode.
As corrective action, the licensee revised the maximum allowable aggregate
opening size to reflect both new limits. A review of the current integrity of the
i
,
centrol room was conducted to assure thet the aggregate opening size was withm
the new specifications. During the inspection period, the inspectors performed a
-
walkdown of CREVS with plant engineering personnel. The inspectors noted
several potential leakage points in the CREVS ductwork located in the adjacent
ventilation equipment room. Some small amount of airflow was noted from two
{
humidifier drains and from a portion of duct tape that was on ventilation ductwork.
A PCAOR was initiated and the outleakage was corrected. The inspectors also
reviewed the plant records of charcoal testing on various ventilation systems for the
past several years. The licensee had routinely been within the TS requirements; no
substantive problems were noted.
18
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-.
i
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<
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l
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j
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,
,
-The licensee initiated actions to revise appropriate sections of the USAR. In
'
.
addition, Licensee Event Report (LER) 96-007-00 was submitted on this matter.
o
i
Subsequently, the station elected to not operate the CREVS in recirculation mode
l
'
following a LOCA, but, rather to manually initiate the system in positive pressure
[
mode immediately. This will result in not ex% Jing the 30 rem thyroid limit.
!
i
Associated procedure changes fiacluding emergency operating procedures) were in
l
process at the end of the inspection penod.
'
5
i
i
c.
Conclusions
!
i
Due to errors in original engineering calculations, following a LOCA, CREVS design
would not have adequately maintained radiation dose to operators less than 30 rem
.
.
!
. thyroid (accepted assumed equivalent to GDC 19 limit of 5 rem whole body). In
~!
the recirculation mode, a non-conservative inleakage flowrate was assumed. In the
i
,
!
positive pressure mode, a non-conservative maximum aggregate opening size was
j
assumed, one large enough to adveisely affect the CREVS ability to maintain the
requisite pork
sure in the control room. Appropriate corrective actions were
!
[
subsequent
ar were in process at the end of the inspection period.
j
,
i
.
I
10 CFR 50, Appendix A, General Design Criteria (GDC) 19, specifies that, "A
control room shall be provided from which actions can be taken to operate the
- {
i
nuclear power unit safely under normal conditions'and to maintain it in a safe -
1
>
l
condition under accident conditions, including loss-of-coolant accidents. Adequate
j
radiation protection shall be provided to permit access and occupancy of the control
!
4
room under accident conditions without personnel receiving radiation exposures in
.
excess of 5 rem whole body, or its equivalent to any part of the body, for the
j
duration of the accident." The accepted whole body equivalent dose to the thyroid
is 30 rem.
i
i
10 CFR 50, Appendix B, Criterion lil, specifies, in part, that " Measures shall be -
!
I
established to assure that applicable regulatory requirements and the design basis,
i
l
are correctly translated into specifications, drawings, procedures, and instructions."
l
.
'
Since the design and operation of CREVS would not have limited control room
,
operator radiation dose to less than 30 rem thyroid following a postulated LOCA,
i
this is considered a violation of 10 CFR 50, Appendix A, GDC 19; and 10 CFR 50,
t
Appendix B, Criterion ill. However, the violation was identified by plant personnel,
j
was not likely to have been discovered as the result of routine surveillance or
!
quality assurance activities, and appropriate corrective actions were taken. As
3
such, the provisions of Section Vll.B.3 of NUREG-1600, " General Statement of
Policy and Procedures for NRC Enforcement Actions" were satisfied and
enforcement discretion deemed appropriate. Therefore, this matter is considered a
!
non-cited violation (50-346/96014-03(DRP)).
f
.
4
1
19
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j
E1.2 10 CFR 70 Criticality Monitors Not Installed
.
.
a.
Insoection Scope (37551)
,
Recently the NRC Office of Nuclear Reactor Regulation (NRR) conducted a review
of the licensee's conformance to 10 CFR 70.24 requirements. The subject
l
regulation specified,in part, that criticality monitoring be established for fuel
handling areas at the plant.
,
b.
Observations and Findinos
NRR review identified that Davis-Besse had not installed the specified criticality
monitors. However, prior to the issuance of the current operating license, the plant
i
3
had been granted approval to possess special nuclear material (SNM) under a Part
70 license, and under that license, had been granted an exemption from the
requirements of 10 CFR 70.24(a). Subsequently though, when the Part 50
operating license was granted on April 22,1977, transfer of the exemption under
the new license provisions was not made.
c.
Conclusions
As of January 24,1997, the licensee was found to be in violation (50-346/96014-
04(DRP)) of 10 CFR 70.24,in that criticality monitors required by regulation had
not been installed, nor was the licensee granted an exemption from the
requirements under their current operating license.
Shortly after the end of the inspection period, on January 30,1997, the licensee
submitted an exemption request from 10 CFR 70.24 (a) to the NRC on this matter.
E2
Engineering Support of Facilities and Equipment
E2.1 Technical Soecification Interoretation For Shield Buildino Inteority
a.
Insoection Scone (37551)
During review of a November 26 event where a shear bolt associated with a
blowout panelin the #4 mechanical penetration room was inadvertently broken
]
(reference Section 01.2 of this report), the inspectors questioned whether the
licensee had invoked all appropriate TS requirements during the event.
l
b.
Observations and Findinos
During the event, control room operators invoked Technical Specification (TS) 3.6.5.2, " Shield Building integrity", as directed by an internal Technical
Specification Interpretation (TSIR) 89-0007 (revision 0). The action statement
associated with TS 3.6.5.2 allowed for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with an inoperable blowout
panel or door prior to the TS requiring a plant shutdown.
20
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- - . - . - - . ~ .
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4 4
1
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Upon review of the technical specifications, the inspectors questioned whether
.
l
TS 3.6.5.1, " Emergency Ventilation System (EVS)," was also applicable. This was
!
. -
due to the fact that with a blowout panelinoperable and blown out during a LOCA
!
.
,
condition, the opening size in the shield building envelope was sufficiently large to
}
prevent the EVS system from drawing the requisite 1/8" negative pressure per its
!
i
intended design.
!
!
!
Discussions were held with operations management personnel who indicated that
i
i
they did not feel the EVS TS was applicable based upon the TSIR However, they
I
agreed to pursue installation of a larger capacity shear bolt while awaiting delivery
}
j
of a replacement bolt. During that changeout the like-for-like replacement bolt
j
-j
arrived onsite and was subsequently installed.
{
The inspectors remained concerned that the licensee's technical specification
i
,
i
interpretation may allow for a longer than intended limiting condition for operation.
l
}
As such the inspectors requested a formal position on this matter from NRC's
i
Office of Nuclear Reactor Regulation (NRR). This matter will remain an unresolvH
j
j
item (50-346/96014-05(DRP)) pending NRR's response.
t
,
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E8
Miscellaneous Engineering issues (92902)
[
f
f
E8.1
(Onen) Unresolved item (50-346/96010-04(DRP)): Portable air purifiers placed in
'
the control room without appropriate engineering evaluation. During the current
!
'
inspection period another PCAOR was initiated by licensee personnel whereby
j
evaluation of additional equipment recently placed in the control room was
requested for evaluation. The items included computer consoles, printers, and
!
additional support equipment. Based upon the additional examples, it appeared that
i
I
the control of equipment to be placed in the control room was weali. This matter
j
will remain open pending completion of inspector review.
j
i
.4-
E8.2 (Ocen) Insoection Followuo item (50-346/94004-03(DRP)h Nuclear
!
,
j
instrumentation divergence events. During the tenth operating cycle the number
and magnitude of divergence events involving nuclear instrumentation (Nis) 7 and 8
9
had declined. No specific root cause had been determined by the licensee as to the
reason for this. However, they postulated that a combination of fuel types and the
.
i~
core reload scheme both contributed to a reduction in the severity of the
!
divergences observed. This matter will remain open through the current operating
,
i-
cycle to ascertain whether the divergence events continue to decline in magnitude
'
and duration.
-
1i
E8.3 (Closed) Insoection Followun item (50-346/94013-03(DRP)): Qualification of fire
!
extinguishers mounted in containment. The inspector reviewed a licensee
'
engineering evaluation that addressed the qualification of fire extinguishers left in
j
containment during the operating cycle. The evaluation appeared adequate to
ensure that the extinguishers were appropriately qualified for both normal operating
conditions as well as during a postulated LOCA.
21
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I .-
E8.4 (Closed) Insoection Followuo item (50-346/95007-02(DRPil: Unit transient
j
1,
occurred due to electrohydraulic control (EHC) panel push button problem. This
!
o-
matter involved a 200 MWe power transient that occurred when an operator
j
inadvertently pushed a pushbutton that was supposed to have been abandoned in
l
,
{
place during a lamp changeout on the EHC panel. The transient was terminated via
4 -
the operator pushing a second button on the EHC panel that was also supposed _to
.
i
!
have been abandoned in place. The modification to abandon the panel in place had
- ;
i
occurred in the early 1980's. The modification had not completely abandoned the
!
l
control circuitry. Since the event occurred, the licensee completed abandonment of
the associated control circuitry, installed additional label placards on the panel
i
. identifying the pushbuttons as abandoned, and instituted an operational philosophy
- '
to not change light bulbs on equipment abandoned in place.
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l
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i
IV. Plant Suncort
l
l.
R2
Status of Radiation Protection and Chemistry (RP&C) Facilities and Equipment
j
(71707)
'
During the inspection period, the inspectors conducted frequent walkdowns of the
radiological restricted area (RRA) of the auxiliary building. The inspectors observed
radiation, high radiation, and contaminated area controls and postings. Recent area
surveys were reviewed and determined to be reflective of actual radiological -
,
conditions. A sample of survey instruments were reviewed to _ verify they were
!
properly calibrated and functional. Personnel conducting work in the RRA were
.
observed to be doing so in accordance with radiation protection program
'{
requirements. No substantive concems were noted by the inspectors in this area.
1
S2
Status of Security Facilities and Equipment (71707)
,
!
Security facilities and equipment were observed as part of the routine inspection
i
program. Ingress explosive monitors and metal detectors were noted to be in
3
proper operating condition. Perimeter monitoring equipment also appeared to be
functional. Guard force members appeared to perform their duties in a professional,
orderly manner. Plant personnel in general conducted themselves in accordance
with security program requirements. No substantive concerns were noted in this'
area,
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F1
Control of Fire Protection Activities
F1.1
Fire Briande Cactain Potentially Not Qualified
a.
. Insoection Scoos (71707)
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During the inspection period, the inspectors performed a review relating to an
observation that one fire brigade captain designated by way of the control room
placard system was an individual that did not have current fire brigade captain
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qualifications.
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b.
Observations and Findinas
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The inspectors noted during a control room walkdown that the "outside SRO" was
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designated as the fire brigade captain for a particular shift. However, the
inspectors recognized that the individual so designated ~was not currently qualified
for that position. Discussions with operations management revealed that he was
.e
placed in that position because the other SRO onshift who could be designated the
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fire brigade captain was completing a required SRO proficiency watch as the control
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room SRO. An agrecrnent had been made on the shift that in the event of a fire,
3
the unqualified individual would turn over to the qualified individual and the two
would then exchange job positions for the fire condition.
4
Once the concern was identified to the licensee, operations management committed
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to not utilize the unqualified individual in the subject manner until resolution was
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reached and/or he obtained final fire brigade captain qualification. Pending
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completion of inspector review of this matter, this is considered an unresolved item
(50-346/96014-06(DRP)).
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V. Manaaernant Mantings
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Exit Meeting Summary-
1
. The inspectors presented the preliminary inspection results to members of licensee
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management at the conclusion of the inspection on January 24,1997. A followup
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meeting was held on February 13,' 1997 where the finalinspection results were discussed
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~ with licensee management representatives. The licensee acknowledged the findings
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presented.
The inspectors asked the licensee whether any materials examined during the inspection
should be considered proprietary. No proprietary information was identified.
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PARTIAL LIST OF PERSONS CONTACTED
Licensee
J. K. Wood, Vice President, Nuclear
J. H. Lash, Plant Manager
R. E. Donnellon, Director, Engineering & Services
T. J. Myers, Director, Nuclear Assurar.ce
L. M. Dohrmann, Manager, Quality Services
R. C. Zyduck, Manager, Design Basis Engineering
J. L. Michaelis, Manager, Maintenance
J. L. Freels, Manager, Regulatory Affairs
M. C. Beier, Manager, Quality Assessment
W. J. Molpus, Manager, Nuclear Training
L. D. Hughes, Manager, Supply
D. L. Echelman, Manager, Operations
R. J. Scott, Manager, Radiation Protection
G. A. Skeel, Manager, Security
D. M. Imlay, Superintendent, Operations
D. H. Lockwood, Supervisor, Regulatory Affairs
J. E. Reddington, Superintendent, Maintenance
R. B. Coad, Superintendent, Radiation Protection
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INSPECTION PROCEDURES USED
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$
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IP 37551:
Onsite Engineering
>
IP 61726:
Surveillance
IP 62707:
Maintenance
IP 71707:
Plant Operations
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IP 92901:
Followup - Operations
IP 92902:
Followup - Engineering
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IP 92903:
Followup - Maintenance
<
$
ITEMS OPENED, CLOSED, AND DISCUSSED
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Ooened
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50-346/96014-01
Failure to Follow Procedures.(3 examples)
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50-346/96014-02 _ UNR Potential Preconditioning of AFW Flow Valve
50-346/96014-03
NCV Inadequate CREVS Design Calculations
50-346/96014-04
Failure to Meet 10 CFR 70.24 Requirements
{
50-346/96014-05 ~ UNR Technical Specification Interpretation for EVS
50-346/96014-06
UNR Fire Brigade Captain Not Qualified
Citaed
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50-346/94011-02
UNR Refueling Related Events
50-346/94017-01
IFl
SRO Shift Schedule
,
50-346/96006-03
IFl
EDG Lube Oil Check Valve Failure
50-346/94006-01
Inadequate FME, Scaffolding, and Standing Order
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50-346/94008-01
Unauthorized Manipulation of EDG Room Ventilation
50-346/95007-03
SW1399 Not Properly Restored After VOTES Test
t
50-346/95008-04 VIO
T413 Not included in M&TE Program
50-346/96002-03
UNR Scaffolding Erected Adjacent to Blowout Panel
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50-346/94013-03
IFl
Qualification of Containment Fire Extinguishers
50-346/95007-02
IFl
EHC Circuit Not Properly Abandoned in Place
4
Discussed
50-346/96005-01
UNR SRO Proficiency Watch Guidelines
50-346/96010-04
UNR Portable Air Purifiers Placed in Control Room
,
50-346/94004-03
IFl
NI Divergence
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LIST OF ACRONYMS USED
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AFP
auxiliary feedpump
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as low as reasonably achievable
.
American Society of Mechanical Engineers
containment air cooler
CFR
Code of Federal Regulations
CNRB
Company Nuclear Review Board
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engineered safety feature
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EVS
emergency ventilation system
HPl
high pressure injection
l&C
instrumentation and controls
IFl
inspection followup item
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IR
inspection report
inservice inspection
licensee amendment request
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LCO
limiting condition for operation
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LER
licensee event report
motor control center
motor operated valve
MWO
maintenance work order
non-cited violation
NI
nuclear instrumentation
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NRC
Nuclear Regulatory Commission
Nuclear Reactor Regulation
PCAQR
potential condition adverse tn quality report
quality assurance
OC
quality control
RE
radiation element
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reactor operator
[
radiation protection
SRB
Station Review Board
senior reactor operator
TS
technical specification
TSIR
technical specification interpretation request
unresolved item
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violation
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