ML20135B090
| ML20135B090 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 11/14/1996 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20135B071 | List: |
| References | |
| 50-289-96-06, 50-289-96-6, NUDOCS 9612040162 | |
| Download: ML20135B090 (18) | |
See also: IR 05000289/1996006
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U. S. NUCLEAR REGULATORY COMMISSION
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REGION I
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- Docket No.
50-289
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License No.
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Report No.
96-06
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Licensee:
GPU Nuclear Corporation
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Facility:
Three Mile Island Station, Unit 1
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Location:
P.O. Box 480
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Middletown, PA 17057
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Dates:
August 4,1996 - September 28,1996
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inspectors:
Samuel L. Hansell, Senior Resident inspector
Dan E. Billings, Resident inspector
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Approved by:
Peter W. Eselgroth, Chief
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Reactor Projects Section No. 7
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9612040162 961114
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ADOCK 05000289
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EXECUTIVE SUMMARY
Three Mile Island Nuclear Power Station
Report No. 50-289/96-06
This integrated inspection included aspects of licensee operations, engineering,
maintenance, and plant support. The report covers a 8 week period of resident inspection
for unit 1.
Plant Operations
An improvement was noted in the area of operability determinations. Senior reactor
operators (SROs) identified and properly dotamented a degradation of Auxiliary Building
and Fuel Handling Building Ventilation (ABFHV) system flow. The ABFHV flow dropped
below the Technical Specification minimum value during the routine purge of the Reactor
Building. The shift SRO entered the applicable TS Limiting Condition for Operation (LCO)
and initiated efforts to troubleshoot the problem (Section 01.3).
Plant operations performed and implemented multiple detailed on-line safety risk
assessments for planned safety related equipment outages. The applicable system
Technical Specification limiting conditions for operation were entered and exited correctly
for the equipment outage times (Section 01.3).
An area for improvement was noted for the senior reactor operator's lSROs) review and
understanding of a maintenance work activity and the associated impact on plant
operation. A Shift SRO authorized an emergency feedwater (EFW) controller pushbutton
module removal without a complete understanding of the module's impact on the controller
operation. The diligence of the system engineer ultimately resulted in the correct
understanding of the EFW module operation (Section 01.2).
Licensed operators were very knowledgeable about the decay heat pump generic minimum
flow concerns. The operators' had an excellent working knowledge of the small break loss
of coolant accident (SBLOCA) emergency operating procedure (EOP). The SBLOCA EOP
provided clear and concise written directions for the operators (Section E1.1).
Maintenance
The maintenance and surveillance test activities observed during this inspection were
performed satisfactorily and demonstrated that the associated systems could perform their
design safety functions (Section M1.1).
The weekly component vibration monitoring program, not required by the IST program,
was an example of an excellent maintenance initiative to detect and correct safety related
equipment problems before the component fails or becomes inoperable. The mechanical
maintenance workers' excellent attention tc, detail resulted in the replacement of the pump
without impacting the other safety related equipment in the river water screenhouse
(Section M1.1).
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Enaineerina
The TMl engineering response to the Crystal River decay heat pump minimum flow safety
issue was comprehensive, thorough, and demonstrated management's commitment and
perseverance to resolve the generic safety issue (Section E1.1).
Engineering management promptly addressed a potential design concern related to the
safety related decay river cooling water system. The immediate and long term system
operability concerns were addressed by a detailed plant review group evaluation and
subsequent engineering reviews (Section E1.2).
Plant Support
TMI identified a repeat problem related to the failure to control a posted high radiation
barrier. This issue is considered a violation of the Unit 1 Technical Specifications. The
immediate corrective actions were comprehensive (Section R2.1).
The site investigation team performed a detailed, thorough, and timely review of the high
radiation barrier incident. Plant Management support was meaningful and focused
significant resources on the incident to resolve the recent repeat work problems (Section
R 2.1 ).
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TABLE OF CONTENTS
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EX EC UTI VE S U M M A RY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ii
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TABLE O F CO NT EN T S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iv
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l. Operations
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Conduct of Operations (71707)
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01.1 General Comments
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01.2 Operability of Emergency Feedwater Equipment / Systems
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01.3 Operability of Auxiliary Building Ventilation Equipment / Systems . . 2
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11. Maintenance
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M1
Conduct of Maintenance (62707,61726)
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M 1.1 General Comments
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Ill . Engine ering . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
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Conduct of Engineering (37551) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
E1.1
TMI Response to the Crystal River Decay Heat Pump Minimum
Flow issue (37 5 51 ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
E1.2 Operability of the Decay River Water System (37551)
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IV. Plant Support
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R2
Status of RP&C Facilities and Equipment . . . . . . . . . . . . . . . . . . . . . . . 6
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(Opened VIO, 50-289/96-06-01) Loss of Controls For a High
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Radiation Area (92904)
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P3.1
Emergency Plan Procedures and Documentation (71750)
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M a nage m e nt Me eting s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10
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Exit Meeting Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10
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PARTIAL LIST OF PERSONS CONTACTED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11
ITEMS OPENED, CLOSED, AND DISCUSSED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12
LIST OF ACRO NYMS U SED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13
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Report Details
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Summarv of Plant Status
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Unit 1 remained at 100% power throughout the inspection period.
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Conduct of Operations (71707)'
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01.1 General Comments
Using Inspection Procedure 71707, " Plant Operations," the inspectors conducted frequent
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reviews of ongoing plant operations. In general, the conduct of operations was
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professional and safety-conscious; specific events and noteworthy observations are
detailed in the section below. Improved Technical Specification (TS) operability
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determinations and documentation by the shift senior reactor operators (SROs), were noted
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for the emergency feedwater (EFW) and Auxiliary Building ventilation work activities.
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The operations department performed and implemented multiple detailed on-line safety risk
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assessments for planned safety related equipment outages. The applicable system
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Technical Specification limiting conditions for operation were entered and exited correctly
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for the equipment outage times.
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O 1.2 Operability of Emeraencv Feedwater Eauioment/ Systems
a.
Inspection Scope
An area for improvement was noted during the SRO review of the EFW control module, EF-
V-30A, work package. The flow control module was removed from the main control panel
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to allow the instrumentation and control (l&C) technicians to perform a routine cleaning
and inspection of the controller pushbutton and spring assembly, in the past two years,
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the controllers had experienced intermittent problems when the operators swapped the
controllers from the automatic to manual mode of operation.
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Observations and Findinos
The shift SRO authorized the work activity without a complete understanding of the
operational impact of the I&C work in relation to the EFW module. It was not clear to the
SRO if the controller would swap from the normal automatic mode to a manual mode of
operation when the l&C technician removed the pushbutton assembly. The SRO contacted
the system engineer to determine the correct controller operation with the pushbutton
assembly removed from the controller.
' 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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After a review of the EFW logic prints and associated technical documents, the system
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engineer informed the shift SRO that the EFW controller would remain in the automatic
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mode if the pushbutton assembly was removed. However, to verify the correct EFW
response the system engineer contacted Foxboro, the flow controller vendor. The vendor
informed the engineer that the EFW flow controller could switch from automatic to the
manual control mode when removing and installing the pushbutton assembly. The updated
information was provided to the control room personnel. The diligence of the system
engineer ultimately resulted in the correct understanding of the EFW flow control module
operation,
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Conclusions
An area for improvement was noted for the senior reactor operator's review and
understanding of a maintenance work activity and the associated impact on plant
operation. One example was noted when a shift SRO authorized an emergency feedwater
controller pushbutton module removal without a complete understanding of the module's
impact on the controller operation. The diligence of the system engineer ultimately
resulted in the correct understanding of the EFW module operation
01.3 Operability of Auxiliary Buildina Ventilation Eauioment/ Systems
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Insoection Scoce
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An improvement was noted in the area of the SROs' Technical Specification equipment
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operability determinations. The SROs identified and properly documented a degradation of
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Auxiliary Building and Fuel Handling Building Ventilation (ABFHV) system flow. The
ABFHV flow dropped below the TS minimum value during the routine purge of the Reactor
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Building.
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Observations and Findinas
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On August 7,1996, control room operators' responded to a plant computer ABFHBV low
flow alarm. The ventilation low flow alarm was received after the initiation of the Reactor
Building (RB) purge system in preparation for a routine tour and maintenance activities.
The ABFHBV exhaust low flow alarm was received when total flow dropped to 99,160
cubic feet per minute (CFM). The shift SROs declared the system inoperable and entered
TS # 3.15.3.3b. An event or near miss capture form (ENMCF) was initiated to document
the problem and initiate a root cause evaluation for the event.
The inspectors reviewed the ABFHBV control room flow recorder, SRO log, RB purge
procedure and associated TSs. The shift SRO entered the applicable TS Limiting Condition
for Operation (LCO) when flow dropped below 100,580 CFM. The SRO log included the
same ABFHBV out of service time when compared to the flow recorder strip chart
readings. The SRO promptly documented the abnormal occurrence with an ENMCF to
ensure the problem was evaluated to determine the reason for the flow change. The
ventilation flow returned to normal after the RB purge system was secured.
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Conclusion
An improvement was noted in the area of operability determinations. SROs identified and
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properly documented a degradation of the ABFHBV system flow. The flow dropped below
the Technical Specification minimum value during the routine purge of the Reactor Building.
The shift SRO entered the applicable TS Limiting Condition for Operation (LCO) and
initiated efforts to troubleshoot the problem.
11. Maintenanc_gt
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Conduct of Maintenance (62707,61726)
M 1.1 General Comments
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Insoection Scoce
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The inspectors observed all or portions of the following maintenance and surveillance work
activities:
Job Order No. 121455, " Clean and Inspect EF-V-30A Hand Auto Controller."
Job Order No. 126097, " Troubleshoot the Main Condenser Offgas Radiation
Monitor RM-A 15."
Job Order No. 123250, " Lubricate the Feedwater Control Valves FW-V-
17A/B."
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Job Order Nos. 126111 and 126112, " Nuclear River Water Pump NR-P-1C
Pump High Vibration and Overhaul."
Electrical Maintenance Procedure E-1, " Vibration Monitoring for the River
Water Pumps."
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Surveillance Procedure 1303-3.1, " Control Rod Movement."
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Surveillance Procedure 1303-5.1, " Reactor Building Emergency Cooling and
Isolation System Logic Channel and Component Test."
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Observations and Findinos
The electrical maintenance routine vibration monitoring of safety related equipment resulted
in the early detection of a problem related to the nuclear river (NR) water pump, NR-P-1C.
The component vibration readings, not required by the inservice test (IST) program, were
an example of an excellent initiative by the maintenance department to detect and correct
equipment problems before the component fails or becomes inoperable.
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The inspector observed a significant portion of the mechanical maintenance pump repair
activities. The pump shaft and impeller were removed and replaced with new parts. The
location of the pump resulted in very limited work space for the maintenance mechanics to
remove the long pump column and shaft sections. The workers excellent attention to
detail resulted in the replacement of the pump without impacting the other safety related
equipment in the river water screenhone.
The emergent pump work was prioritized appropriately based on the significance of the
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safety related cooling water pump with respect to the previously planned work activities.
The plant risk associated with the out of service pump was low due to the fact that the
degraded pump was not required to be operable by Technical Specifications.
c.
Conclusions
The maintenance and surveillance test activities observed during this inspection were
performed satisfactorily and demonstrated that the associated systems could perform their
design safety functions.
The extra component vibration readings, not required by the IST program, were an example
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of an excellent initiative by the maintenance department to detect and correct equipment
problems before the component fails or becomes inoperable. The mechanical maintenance
workers' excellent attention to detail resulted in the replacement of the pump without
impacting the other safety related equipment in the river water screenhouse.
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Ill. Enaineerina
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Conduct of Engineering (37551)
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TMI Response to the Crvstal River Decay Heat Pumo Minimum Flow issue (37551)
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a.
Inspection Scope
A generic issue was raised at the Crystal River (CR) nuclear power plant related to the
decay heat (DH) pump design minimum flow capabilities. Based on the most recent
concerns TMI and CR have initiated a DHP minimum flow test with at an independent test
facility. The spare TMI DH pump was selected to perform a simulated minimum flow test
that would be representative of the post small break loss of coolant accident (SBLOCA)
conditions. As of October 21,1996, the pump had run for twelve days without any
unexpected problems.
b.
Observations and Findinas
TMl's two low pressure injection pumps are manufactured by Worthington and are similar
to the CR design. A difference is the minimum flow: at TMI the flow is 125 gallons per
minute (GPM) compared to the 80 GPM at CR. The pumps are single stage centrifugal
pumps rated at 3000 GPM at 151 pounds per square inch (psig). According to the TMI
updated final safety analysis (UFSAR) Chapter 6.1-4, the pumps can operate indefinitely at
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shutoff head with the 125 GPM minimum flowrate. Also, in response to Bulletin No.
88-04, TMl noted that the operation of the pumps in the low flow mode would last no
more than six hours under actual ECCS operation. TMI engineering is currently involved
with this issue through the Babcock and Wilcox owners group and have attended the
recent meetings in NRC Headquarters about the Bulletin No. 88-04 issue.
The inspectors reviewed the decay heat system section of the UFSAR, emergency
operating procedures (EOPs), and the plant review group (PRG) evaluation of the issue.
The TMl DH pumps start on_ low reactor coolant system (RCS) pressure at 1600 psig and
again at 500 psig, and a RB pressure of 4 psig. The TMl EOPs have the operators secure
the DH pumps if they started at 1600 psig RCS pressure OR 4 psig RB pressure and the
event is a SBLOCA that results in sufficient high pressure injection (HPI) flow and a slow
drop in RCS pressure. The pumps would receive an automatic signal to start when RCS
pressure dropped to 500 psig and begin injecting at approximately 250 psig.
All control room operators were familiar with the CR issue and understood how the TMI DH
pumps were designed. The operators' were very knowledgeable about the DH pump
automatic start signals and the expected system response to a SBLOCA. They had an
excellent working knowledge of the SBLOCA EOPs and understood the operational
limitations for the DH pumps. The EOPs provided clear direction for the operators about
the DH pump starting and stopping requirements.
The PRG evaluation of the fr:,ue concluded that the DH pumps were operable. The
determination was based upon the past operating history of the pumps and the fact that
the TMI pumps have been maintained in very good condition. The TMI pumps do have an
intermittent vibration condition that Engineering is trying to resolve. The PRG also
proposed four alternative flowpaths to maximize the DH system flow if the plant
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experienced a SBLOCA and the RCS suction path was unavailable for long term decay heat
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removal operation. Management's decision to perform the pump minimum flow test
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showed their willingness to completely resolve this potential safety issue.
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Conclusions
The TMI engineering response to the Crystal River decay heat pump minimum flow concern
was comprehensive, thorough, and demonstrated management's commitment and
perseverance to resolve the generic safety issue.
Licensed operators were very knowledgeable about the decay heat pump generic minimum
flow concerns. The operators' had an excellent working knowledge of the small break loss
of coolant accident emergency operating procedure. The SBLOCA EOP provided clear and
concise written directions for the operators.
E1.2 Operability of the Decav River Water System (37551)
A potential safety concern was raised by engineering related to the decay river (DR)
cooling water flow measurement. The decay heat river water flow data could be
unconservative due to the location of the annubar flow instrument near a pipe elbow, if
the DR system flow indicated higher than actual flow, then the measured flow could be
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less than the Technical Specification requirement. Increasing flow above current levels
raises issues with emergency diesel loading. Engineering management recognized the
potential safety significance of the concern and convened a plant review group (PRG)
meeting to evaluate the system operability and if the issue met the reportability criteria in
administrative procedure AP-1044. A material nonconformance report (MNCR) was
submitted to document the potential safety concern and track the recommended corrective
actions.
The PRG evaluated the issue on August 20,1996, and determined that a concern may
exist under the worst case conditions assumed in the updated final safety analysis report
(UFSAR). The "as found" DR system data that was presented to the PRG showed that the
heat exchanger fouling, strainer differential pressure (DP), river water level and river water
temperature did not exceed the design limits. Considering the above, the PRG determined
that the decay river system was operable and the issue was not reportable. The Shift
Engineer (SE) was instructed to monitor and log the DR plant conditions to assure that the
system does not approach the safety analysis assumptions.
The inspectors verified that the SEs understood the DR safety concern and that the DR
data was properly monitored and logged. In addition, the associated DR parameter
computer alarm setpoints were adjusted to alert the control room personnel before the
system deigri limits were exceeded. The PRG meeting conducted a detailed review of the
potential safety concern and recommended appropriate actions to ensure the DR system
operability would be considered if a key design parameter exceeds a value used for the
safety analysis calculations. The PRG also assigned engineering an action item to
determine the proper flow alignment during the routine system flow test.
In summary, engineering management promptly addressed a potential design concern
related to the safety related decay river cooling water system. The immediate and long
term system operability concerns were addressed by a detailed plant review group and
engineering evaluation.
IV. Plant Support
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Status of RP&C Facilities and Equipment
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R2.1
(Opened VIO. 50-289/96-06-01) Loss of Controls For a Hiqh Radiation Area
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(92904)
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Inspection Scoce
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The inspectors reviewed a licensee-identified incident that occurred on August 7,1996,
involving a barrier to a posted high radiation area (HRA) in the Auxiliary Building (AB) to
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determine the effectiveness of the licensee's root cause investigation and proposed
corrective actions. Postings on the barrier were appropriate and cautioned workers that
the barrier was required by TS. Actual dose rates in the area varied from 300 to 500
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millirem per hour on contact and 90 mrem at 30 centimeters. The barrier was estimated
out of position for approximately three hours. The highest radiation dose received by a
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plant worker was approximately 6 millirem on the day of the moved barrier.
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b.
Observations and Findinos
The HRA and contamination postings around the AB 'B' emergency safeguards vault area
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were changed on dayshift to support the plant preservation activities. The activities
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included the floor scabbling and painting work in the 'B' decay heat removal and building
spray vault areas. The radiological controls technician requested that the contaminated
area and high radiation area postings be extended to the boundary of the entire area to be
scabbled. The expanded boundaries provided a more convenient working area for the
preservation activities. The group radiological control supervisor (GRCS) approved the
posting change in the 'B' emergency safeguards vault area. After completion of the floor
scabbling work activities at approximately 2:30 p.m., the preservation crew left for the day
with the high radiation area gate in the proper position.
At approximately 5:00 p.m. a second dayshift GRCS inspected the expanded postings
surrounding the 'B' emergency safeguards vault area. The GRCS specifically remembered
looking at the two high radiation area swing gates around the work area access points sr.d
determined that the posting and gates were properly positioned. At approximately 8:00
p.m., the high radiation area swing gate at the west access point step off pad was found
to be moved 180 degrees and did not provide a barrier to the area, as required by the TMl
Technical Specification 6.12.1.a. An Auxiliary Operator (AO) noticed the gate open and
inforrned the in plant shift foreman (SF). The SF immediately notified the duty GRCS of the
situation. The GRCS and SF photographed the HRA gate in the as found condition, the
gate was returned to its proper position, and a second photograph was taken to capture
the HRA barrier condition. TMI management was informed of the incident, a radiation
survey was performed of the AB area, and an event near miss and capture form was
initiated to document the problem.
On August 8,1996, TMl initiated an investigation to determine the cause of the incident.
The investigative team consisted of a representative from nuclear safety assessment
(NSA), security, operations, maintenance, and radiological controls departments. Based on
the GRCS verification and AO recognition of the HRA barrier, the Team determined that the
high radiation barrier was prGpped open between 5:00 p.m. and 8:00 p.m. on August 7th.
Computer printouts were obtained that listed all of the personnel signed on an RWP (18
workers) and all of the personnel that key carded through the AB vital area doors (43
workers) between the hours of 5:00 p.m. and 8:00 p.m. The investigation team
conducted individual interviews with the personnel that accessed the AB. Ir* addition,
plant walkdowns of the 'B' emergency safeguards vault area were performed with the
GRCS that inspected the area at 5:00 p.m. and the AO and SF that found the gate propped
open at 8:00 p.m. Nobody that was in the AB on August 7th admitted to moving or
knowing who moved the HRA gate during the three hour time period.
The licensee conducted an investigation and critique of this event and determined that one
root cause that had contributed to this event was a lack of attention to detail by personnel
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working in the room. Since the time of this event, the licensee has further investigated
and analyzed the failures and implemented some corrective actions. These actions
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included providing a review of the event to all operations personnel, reviewing the
corrective actiors taken as a result of previous similar incidents, evaluating all HRA
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postings and barriers for " user friendliness", and coordination of feedback from employees
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on the condition and ease of use for HRA barriers.
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The inspectors interviewed plant workers independently to determine the facts about the
incident. The inspectors determined that the employees understood their responsibilities
for ensuring that radiological controls and barriers were in place when they entered and
exited a radiological area. Similar to the GPUN investigation, nobody admitted that they
moved or knew who moved the HRA gate on August 7,1996. It appears that there were
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three probable reasons for the moved barrier: 1) the barrier made it difficult to remove
protective clothing at the contaminated and HRA boundary; 2) a plant worker moved the
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barrier to transport equipmant into or out of the contaminated /HRA: and 3) an unknown
person intentionally opened the HRA swing gate.
The inspectors also reviewed the maintenance activities performcd in the AB on August
7th. The maintenance tasks that were performed in the AB did not require the workers to
pass through the HRA boundary on the 281 foot elevation. A review of the personnel
exposure on the day of the moved the barrier revealed that the maximum individual dose
was 6 millirem.
The site investigation team performed a detailed and thorough review of the incident. The
Team was assembled promptly after the facts about the issue were known Management
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recognized the significance of the repeat HRA barrier problems and assigned the most
knowledgeab!e personnel onsite to perform the root cause investigation. The investigation
was initiated immediately after the Team was formed in an attempt to increase the
probability that the person and reason for the moved barrier would be determined. As part
of the interview process, personnel were asked for suggestions and recommendations to
improve the TMl radiological controls program with respect to high radiation area posting
and control. The following is a summary of the comments:
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The vast majority of suggestions by both radiological controls and non-
radiological controls personnel were to eliminate where possible, or reduce to
a minimum, overly conservative high radiation area postings within TMI. The
incident described in the event capture form involved a high radiation area
posting that was in a 2 mr/hr radiation field.
Use of " turnstiles", instead of swing gates, for high radiation area barriers
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would ensure that a physical barricade is in place at all times.
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Installation of devices similar to limit switches on the swing gates that would
cause a light or buzzer to energize when the gate was open.
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Placement of video cameras at high radiation area gate locations.
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Increase TMl worker sensitivity to company and regulatory concerns with
repeated high radiation area barrier violations, through short training
discussions, videotapes, or shop meetings.
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c.
Conclusions
The immediate corrective actions were comprehensive. Because of the repeat problem,s
related to the control of high radiation barriers, this latest example is considered a violation
of the TMI Unit 1 TS 6.12.1.a. (VIO 50-289/96-06-01).
The site investigation team performed a detai!ed and thorough review of the incident. The
team was assembled promptly after the facts about the issue were known. Plant
Management support was meaningful and focused significant resources on the incident to
resolve the recent repeat work problems.
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P3.1
Emeraency Plan Procedures and Documentation (71750)
The Region i emergency preparedness inspector performed an in-office review of the
revisions to the TMI emergency plan irnplementing procedures. Based on your
determinations that the changes did not decrease the overall effectiveness of your
emergency plan, and that it continues to meet the standards of 10 CFR 50.47(b) and the
requirements of Appendix E to Part 50, NRC approval is not required. Our initial review of
these changes indicates them to be in accordance with 10 CFR 50.54(q). Implementation
of these changes will be subject to inspection to confirm that they have not decreased the
overall effectiveness of your emergency plan. A list of the specific revisions that were
reviewed is included below. The inspector concluded that the revisions did not reduce the
effectiveness of the E-Plan and were acceptable.
EMERGENCY PLAN AND IMPLEMENTING PROCEDURES REVIEWED
Document
Document Title
Revision
EPIP-TMI .01 Emergency Classification and Basis
4
EPIP-TMf .07 Activation of the RAC
2
EPIP-TMI .16 Contaminated injuries
4
EPIP-TMI .27 Emergency Operations Facility
7
EPIP-TMI .28 Activation of the TSC
6
TEP-ADM-1300.02 Emergency Preparedness Training
1
TEP-ADM-1300.04 Administration of the TMl Initial
Response and Emergency Support
Organization Duty Roster
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10
Manaaement Meetinas
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Exit Meeting Summary
At the conclusion of the reporting period, the resident inspector staff conducted an exit
meeting with TMl management on October 15,1996, summarizing Unit 1 inspection
activities and findings for this report period. TMl staff comments concerning the issues in
this report were documented in the applicable report section. No proprietary information
was identified as being included in the report.
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PARTIAL LIST OF PERSONS CONTACTED
Licensee
J. Knubel, Vice President, TMI
- M. Ross, Director, Operations and Maintenance
L. Noll, Plant Operations Director
R. Maag, Plant Maintenance Director
D. Etheridge, Radiological Controls / Occupational Safety Director
J. Schork, Regulatory Affairs
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J. Wetmore, Manager, Regulatory Affairs
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D. Hosking, NSA Manager
G. Skillman, Technical Functions Site Director
P. Walsh, Engineering Director
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R. Hess, Training Manager
- senior licensee manager present at exit meeting on October 15,1996.
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NRC
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J. Norris, TMl Project Manager, NRR
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INSPECTION PROCEDURES USED
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IP 62707:
Maintenance Observation
IP 71707:
Plant Operations
IP 37551:
Onsite Engineering
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lP 71750:
Plant Support Activities
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IP 92904:
Followup - Plant Support
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ITEMS OPENED, CLOSED, AND DISCUSSED
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Ooened
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50-289/96-06-01, " Loss of Controls For a High Radiation Area" (VIO).
Closed
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None
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Updated
None
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LIST OF ACRONYMS USED
Auxiliary Building
Annex to the Emergency Operations Facility
As low As Reasonably Achievable
American Society of Mechanical Engineers
Core Damage Frequency
Committed Effective Dose Equivalent
CR
Control Room
CFR
Code of Federal Regulations
Design Basis Documents
Emergency Director
Emergency Feedwater
ENMCF
Event or Near Miss Capture Form
Emergency Plan and implementing Procedure
Engineered Safety Feature
High Efficiency Particulate
IFl
Inspection Followup Item
Individual Plant Evaluation
IR
inspection Report
Inservice Testing Program
JO
Job Order
4
LCO
Limiting Condition of Operation
LER
Licensee Event Report
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MNCR
Material Nonconformance Report
Mine Safety Appliance
Non-Cited Violation
NI
Nuclear Instrument
NRC
Nuclear Regulatory Commission
Nuclear Safety Assessment
Nuclear Voluntary Laboratory Accreditation Program
Offsite Dose Calculation Manua!
' Operations Support Center
PAS
Post Accident Sample
Procedure Change Request-
PPB
Part per Billion
Part per Million
Plant Review Group
QV
Quality Verification
RAC
Radiological Assessment Coordinator
Radiological Control Area
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Radiation Protection
Remote Shutdown Panel
Radiation Work Permits
Systematic Assessment of Licensee Performance
SF
Shift Foreman
Senior Reactor Operator
Shift Supervisor
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Temporary Instruction
Thermoluminescent Dosimeter
TS
Technical Specification
Updated Final Safety Analysis Report
Unresolved item
Violation
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