ML20149L295
| ML20149L295 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 07/24/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20149L290 | List: |
| References | |
| 50-277-97-04, 50-277-97-4, 50-278-97-04, 50-278-97-4, NUDOCS 9707310261 | |
| Download: ML20149L295 (45) | |
See also: IR 05000277/1997004
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U. S. NUCLEAR REGUL.ATORY COMMISSION
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REGION I
Docket / Report No.
50-277/97-04
License Nos. DPR-44
50-278/97-04
Licensee:
PECO Energy Company
P. O. Box 195
Wayne, PA 19087-0195
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Facility Name:
Peach Bottom Atomic Power Statinn Units 2 and 3
Dates:
May 4 - June 7,1997
Inspectors:
W. L. Schrnidt, Senior Resident inspector
M. J. Buckley, Resident inspector
B. D. Welling, Resident inspector
J. W. Shea, NRR Project Manager
R. L. Fuhrmeister, Sr. Reactor Engineer
R. L. Nimitz, Sr. Radiation Specialist
Approved By:
P. D. Swetland, Acting Chief
Reactor Projects Branch 4
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Division of Reactor Projects
9707310261 970724
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ADOCK 05000277
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EXECUTIVE SUMMARY
Peach Bottom Atomic Power Station
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NRC Inspection Report 50-277/97-04, 50 278/97-04
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This integrated inspection report includes aspects of resident and region based inspection
of routine and reactive activities in: operations; surveillance and maintenance; engineering
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and technical support; and plant support areas.
Overall Assurance of Quality:
PECO Energy (PECO) operated both units safely over the period.
The PECO quality assurance department (QA) conducted surveillance in a broad range of
areas including operations, maintenance, security, and emergency planning. The written
record of the surveillance showed proper scope and good documentation of conclusions.
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Plant Operations:
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Operators performed routine activities wellincluding controls over the plant during on-line
control rod hydraulic control unit (HCU) maintenance and removal from service of the Unit
3 fifth stage feedwater heaters during Unit 3 coastdown operation. Operators also
responded well when they identified, during clearance application, that a control rod that
was not scheduled to be worked had been inserted, based on reactor engineering direction,
for HCU work. They also performed well when several abnormal conditions developed
during the HCU work, including: an individual Unit 2 control rod scrammed due to a
clearance application error by nuclear material department personnel, several scram air
header leaks and low pressure alarms, and a control rod that drifted from fully withdrawn
to fully inserted, due to scram inlet isolation valve damage and leakage.
Control room operators responded well to a faulty automatic high temperature isolation
switch for the high pressure coolant injection (HPCI) system at Unit 2.
However, during
jumpering of the faulty instrument, an operator lifted the wrong lead on the terminal strip,
causing a partial loss of logic power and inability of HPCI to automatically start. Operators
responded to the loss of logic power and restored the system to operable status within
several minutes. While the operator made a mistake, the procedural guidance on the
installation of jumpers did not provide specific information on where to install jumpere snd
possible problems that could result if not installed properly.
Plant housekeeping was generally excellent, however PECO needed to take actions to
address continued leaking of emergency diesel generator (EDG) lubricating and fuel oil, to
preclude an additional fire hazard.
The inspectors reviewed and closed three licensee event reports (LERs), finding that two
represented technical violations of the operating license core thermal power limit, which
the licensee identified, properly reported and corrected. These failures constituted
licensee-identified and corrected violations are being treated as Non-Cited Violations
consistent with Section Vll.B.1 of the NRC Enforcement Policy.
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Maintenance:
The inspectors found that PECO personnel conducted maintenance and surveillance
activities acceptably during the period.
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PECO planned and coordinated the HCU on-line maintenance generally well. Control room
staff exhibited very good control of post-maintenance testing. However, a maintenance
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technician caused the inadvertent scram of a single Unit 2 control rod due to an error while
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pulling fuses to establish electricalisolation. The inspectors noted that this was the first
time that PECO allowed maintenance technicians to establish their own isolation for work.
Further, this event revealed weaknesses in double verification techniques and supervisory
oversight.
The inspectors concluded that, overall, the maintenance outage on the E-3 and E-1
emergency diesel generators (EDGs) were effectively planned and implemented. PECO
corrected an out-of-specification crankshaft strain measurement on both machines,
implementing enhanced vendor information. Several minor problems, some involving
maintenance rework issues, caused delays in the post-maintenance testing and restoration
to an operable status. PECO contingency planning allowed the problems to be resolved in
a timely manner.
Operators adequately performed quarterly standby liquid control (SLC) system surveillance
testing in accordance with procedures. Operators identified some minor weaknesses and
opportunities for improvements in the recantly revised surveillance test procedure and in
the use of test equipment.
The inspectors reviewed an issue that occurred during the Unit 21996 refueling outage
where a maintenance work group pulled the wrong fuse during preparations for
containment isolation valve local leak rate testing (LLRT) work. This personnel error
resulted in'a loss of electrical power to the mechanical vacuum isolation valves, but no
actual valve motion resulted since the valves were closed. Although this activity resulted
in insignificant safety impact, it represents poor maintenance activity performance.
Unfamiliarity with fuse removal and self-checking methodologies by the personnel involved,
and an inadequate pre-job briefing contributed to the event.
In review of PECO's implementation of the improved technical specification (ITS), the
independent safety engineering group (ISEG) and the NRC noticed difference between the
ITS and the old custom specification, in the definition of an instrument channel functional
test (CFT). This difference caused ISEG to question whether PECO was conducting
adequate testing. The NRC Nuclear Reactor Regulation (NRR) staff to reviewed the issue
and concluded that PECO satisfied the objectives of the ITS CFT requirements, by verifying
the function of one contact in all relays supplying a signal to engineered safety function
system logic. The inspectors also concluded that the ISEG review had led to an improved
understanding of the regulatory requirements in this area.
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Enoineerino:
Engineering department management and system mangers provided very good support to
the El and E3 EDG maintenance outages, particularly in the review and dispositioning of
the crankshaft strain issues discussed in sections M1.2 and M1.3 above.
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The reactor engineers performed well during on-line HCU maintenance, properly reviewing
the needed initial conditions and the thermal limit effects prior to inserting control rods to
be worked, and conducting the post-work scram testing. In one instance, reactor
engineering directed the operators to insert the wrong control rod. There was no adverse
effect on thermal limits, and operators identified and corrected the mistake before work
began.
Plant Sucoort:
PECO implemented an effective fire protection program, maintaining fire fighting equipment
accessible and in good condition. One discrepancy was noted in the incorporation of a PEP
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issue lesson learned into a lesson plan for fire watch training. Housekeeping in the plants
was noted to be excellent. Evaluation of, and corrective actions for discrepancies
identified by audits and self-assessments, were comprehensive and well focused.
Overall effective radiological controls were implemented including planning and preparation
for the Unit 3 outage. The ALARA program was effectively implemented. The external
and internal exposure controls program were effective. Weaknesses were identified in the
evaluation and control of non-routine effluent / material release paths indicating a need for
enhanced attention to detailin this area. A violation for lack of proper controls over high
radiation area keys was identified by the licensee and corrected. Although it does not
appear that the keys were misused, this violation was cited because several keys were
uncontrolled for a number of years.
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TABLE OF CONTENTS
EX EC U TIV E S U M M ARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ii
SUMM ARY OF PLANT ACTIVITIES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1.
Operations
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01
Conduct of Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
04
Operator Knowledge and Performance . . . . . . . . . . . . . . . . . . . . . . . . . 1
04.1 High Pressure Coolant injection inoperable Due to Lifted Lead -
Unit 2...........................................
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Quality Assurance in Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
08
Miscellaneous Operations issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
08.1 (Closed) Licensee Event Reports 2-97-001, 2-97-002, and 3-
97-002..........................................
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II.
Maintenance................................................
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M1'
Conduct of Maintenance and Surveillance . . . . . . . . . . . . . . . . . . . . . . 4
M 1.1 Hydraulic Control Unit On-line Maintenance . . . . . . . . . . . . . . . . 4
M1.2 E-3 Emergency Diesel Generator Maintenance
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M1.3 E-1 Emergency Diesel Generator Maintenance
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M1.4 Surveillance Activities
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M4
Maintenance Staff Knowledge and Performance . . . . . . . . . . . . . . . . . . 9
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M4.1 incorrect Fuse Removal During Local Leak Rate Testing
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M8
Miscellaneous Maintenance issues (92902) . . . . . . . . . . . . . . . . . . . . 10
M8.1 Review of Instrument Channel Functional Test Practices . . . . . . 11
Ill.
E n g i n e e ri n g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12
E1
General Engineering Comments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12
IV.
Plant Support
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F2
Status of Fire Protection Facilities and Equipment . . . . . . . . . . . . . . . . 13
F2.1
Material Condition Inspection and Equipment Inventories
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F3
Fire Protection Procedures and Documentation . . . . . . . . . . . . . . . . . . 15
F3.1
Fire Protection Procedure Reviews
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F5
Fire Protection Staff Training and Qualification . . . . . . . . . . . . . . . . . . 17
F5.1
Fire Brigade Training
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F.5.2 (OpenIFl 50-277/278/97-04-01) Fire Watch Training . . . . . . . . 18
F7
Quality Assurance in Fire Protection
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F7.1
Fire Protection Program Audits . . . . . . . . . . . . . . . . . . . . . . . . 19
F7.2 Review of August 10,1994 .......................... 20
F8
Miscellaneous Fire Protection issues . . . . . . . . . . . . . . . . . . . . . . . . . 21
F8.1
Conformance to Updated Final Safety Analysis Report
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D e s cription . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21
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Radiological Protection and Chemistry (RP&C) Controls . . . . . . . . . . . . 22
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R 1.1 Radiological Controls (Program Changes) . . . . . . . . . . . . . . . . . 22
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R1.2 ALARA Program and Unit 3 Refueling Outage
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Planning, Preparation, Emergent Work Control . . . . . . . . . . . . . 23
R1.3 Internal Exposure Controls . . . . . . . . . . . . . . . . . . . . . . . . . . . 24
R1.4 External Exposure Controls; (Open) Violation 97-04-01:
Inadequate Controls over Locked High Radiation Door Keys
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R1.5 Control of Radioactive Materials and Contamination . . . . . . . . . 28
R5
Staff Training and Qualification in Radiation Protection and Chemistry . 32
RS.1
Radiation Workers / Radiological Controls Personnel . . . . . . . . . . 32
R7
Quality Assurance in Radiological Protection and Chemistry Activities
(83750)
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R7.1 Radiological Event Reports . . . . . . . . . . . . . . . . . . . . . . . . . . . 32
R8
Miscellaneous RP&C Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33
R8.1
(Closed) Unresolved item 96-06-04: Review of Radioactive
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Material Storage Locations Versus Updated Final Safety
Analysis (UFSAR) Descriptions . . . . . . . . . . . . . . . . . . . . . . . . 33
R8.2 Housekeeping
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R8.3 Verification of Updated Final Safety Analysis Commitments
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V.
M a n a g em e nt Meeting s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 35
X1
Exit Meeting Su m m a ry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 35
X2
Review of UFS AR Commitments . . . . . . . . . . . . . . . . . . . . . . . . . . . . 35
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INSPECTION PROCEDURES USED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 36
ITEMS OPENED, CLOSED, AND DISCUSSED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 36
LIST O F ACRO N YM S U SED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 7
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SUMMARY OF PLANT ACTIVITIES
PECO operated both units safely over the period.
Unit 2 remained at essentially 100% power, until May 18, when operators reduced power
to approximately 70% to allow control rod (CR) hydraulic control unit (HCU) on-line
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maintenance. The nuclear maintenance division (NMD) completed the maintenance and
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operators began restoring the unit to 100% power on May 23.
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Unit 3 entered the period at 100 % power, operators reduced power to allow on-line HCU
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maintenance on May 4 and began returning the unit to 100% power on May 11, following
completion of the HCU work. The unit entered end-of-cycle coastdown, ending the period
at approximately 98 % power, af ter removal of the fifth stage feedwater heaters on June
1.
1. Operations
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Conduct of Operations'
Operators performed routine activities wellincluding control of the plant during on-
line HCU maintenance and removal of the Unit 3 fifth stage feedwater heaters from
service. Operators also responded well when they identified, during clearance
application, that a control rod that was not scheduled to be worked had been
inserted, based on reactor engineering direction, for HCU work. They also
performed well when several abnormal conditions developed during the HCU work,
including: a scrammed Unit 2 control rod due to a clearance application error by
NMD personnel, several scram air header leaks and low pressure alarms, and a
control rod that drifted from fully withdrawn to fully inserted, due to scram insert
valve damage and leakage.
Plant housekeeping was generally excellent, however PECO needed to take actions
to preclude an additional fire hazard from continued leaking of emergency diesel
generator (EDG) lubricating and fuel oil.
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Operator Knowledge and Performance
04.1 Hiah Pressure Coolant Iniection Inocerable Due to Lifted Lead - Unit 2
a.
Scope
The inspectors reviewed the circumstances leading to a short period of inoperability
of the Unit 2 high pressure coolant injection (HPCI) system on June 1, while
operators tried to install a jumper to place a failed high area temperature isolation
relay in the tripped condition.
InaiJll.'t d[o*rf.'".". "of".".Sa"ta .llanIn"o'd i".'?"IU"" """ '"* * "'"'*"'"' "'**** *""** * * "" "'""'
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b.
Observations and Findinos
During operator rounds it was noted that the instrument (TE 4944D) did not pass
the required daily channel check. As such, the operators properly decided to insert
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the trip from this instrument in accordance with technical specifications (TSs) and
general procedure (GP) 25.
GP 25 directed that the operator install a jumper between two terminals'that would
bypass the temperature switch contact and cause the high temperature isolation
relay to energize, placing the channel in the tripped condition per TSs.
When the operator went to install the jumper he saw that on one of the terminals
there was a test jack installed on the terminal strip external connection, with no test
connections on the internal side of the strip. The operator looked at the other
connection and decided to install the jumper on the internal side of the terminal
strip. After he lifted the lead on the internal side of the strip, he noted that there
was a flat metal "C" jumper installed between the terminal and the next terminal
below. Lifting the lead by removing the screw had caused a loss of power
continuity through the terminal, and resulted in a loss of power to this channel of
HPCIlogic, making the system inoperable for automatic actuation. The operator
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immediately identified the mistake and re-landed the lifted lead within two minutes
of lifting it.
In review of GP 25 and the operations manual (OM) Section 7.7 that covers the
installation of jumpers by operators, the inspector found that:
GP 25 did not reference OM 7.7
Neither procedure was specific as to which side of a terminal strip was the
external or the internal wiring point.
The procedures did not address the possibility of installation of jumpers
causing a loss of power to other components in the same circuit, since the
terminal screws need to be removed to install a round lugged jumper.
There was no discussion of possible flat "C" jumpers that may be hard to
see.
C.
Conclusions
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The operators responded well to the event and limited the duration of HPCI
inoperability. The operator made a mistake by lifting a lead on the internal
connection strip of the terminal strip. However, the procedure did not provide any
specific guidance on which side of terminal strips jumpers should be installed. The
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inspector will review the PECO corrective actions upon receipt of the licensee event
report.
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Quality Assurance in Operations
a.
Scoce
The inspectors reviewed the quality assurance department's surveillance completed
over the last several months.
b.
Conclusions
PECO QA conducted surveillance in a broad range of areas including operations,
maintenance, security, and emergency planning. The written record of the
surveillance showed proper scope and good documentation of conclusions.
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M!scellaneous Operations issues:
08.1 (Closed) Licensee Event Reports 2-97-001. 2-97-002. and 3-97-002 ~
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a.
Scooe
The inspector reviewed the issues documented in several licensee event reports
(LERs).
b.
Observations and Findinas
LER 2-97-001: Non-Conservative single loop average power range monitor (APRM)
flow biased scram setpoints.
PECO reactor engineering identified this issue as they reviewed previous core power
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and flow data from single loop operation in the 1992 time period. The error
involved was small(less than 1%) and was due to the method used to establish the
flow biased APRM setpoints required by TS during single loop operations. PECO
took appropriate actions to correct and report this issue.
LER 2-97-002: Reactor power slightly greater than licensed thermal power - due to
inaccurate accounting of recirculation pump power in the thermal heat balance.
PECO determined that due to inaccuracies in the instrumentation used to measure
the recirculation pump electrical energy core thermal power may have exceeded, by
1.5 megawatts, the Units 1 and 2 license limit of 3458 megawatts. PECO found
that the watt transducers for the recirculation pumps did not account for the power
factor of the electrical supply. Accounting for the power factor would lessen the
amount of energy actually being used to drive the pump and thus lessen the amount
of actual energy being added to the reactor coolant by the recirculation pump.
Since the calculation subtracts the recirculation pump energy from the core thermal
power (i.e., allowing reactor power to be increased by the amount being added by
the recirculation pumps), not accounting for the power factor allowed actual reactor
power to be above the license limit.
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LER 3-97-002: Manual Reactor Scram due to Natural Circulation Operation
On March 9,1997, while in single loop operation to investigate a low lube oil
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condition indicated on the idle recirculation pump, Unit 3 experienced a loss cf the
operating recirculation pump due to a faulty interlock between the main generator
and the recirculation pump power supply. The resultant manual reactor scram and
recovery were discussed in NRC Inspection Report 50-270/97-02. After a manual
power transfer from the main generator to the offsite power supply, in preparation
for a turbine trip, an auxiliary breaker position contact failed to change position
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indicating that the main generator was still supplying the recirculation pump
switchgear. When the turbine was tripped, the logic for the recirculation pump
thought that the recirculation pump was still powered from the main generator and
tripped the recirculation pump motor generator supply breaker open in anticipation
of the main generator output breaker opening. With the resultant natural circulation
reactor condition, procedures called for a manual reactor scram. PECO repaired the
breaker auxiliary contact and after other plant repairs restarted the unit. PECO
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committed to review the other 13 KV breakers installed at the site to determine if
there were any other problems related to these auxiliary switches. The corrective
actions for this event, and the application of the NRC maintenance rule to the 13
KV breaker switches will be followed during a subsequent inspection. (IFl 50-
278/97-04-01)
c.
Conclusions
PECO adequately documented the conditions described in the LERs reviewed above.
With respect to LERs 2-97-001 and 2-97-002, the inspectors found that these
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events represented technical violations of the operating license that the licensee
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identified, properly reported and corrected. These failures constituted licensee
identified and corrected violations and are being treated as Non-Cited Violations
consistent with Section Vll.B.1 of the NRC Enforcement Policy.
II. Maintenance and Surveillance
M1
Conduct of Maintenance and Surveillance
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The inspectors found that PECO personnel conducted activities acceptably during
the period. There were several performance and equipment issues that developed
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during HCU and EDG maintenance, as discussed below.
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M1.1 Hydraulic Control Unit On-line Maintenance
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a.
Scone (62707)
During the period PECO conducted HCU on-line maintenance at both units. These
activities included scram pilot valve (SSPV) replacement and general HCU
preventive maintenance. The inspector reviewed
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The Unit 3 activities conducted between May 4 and May 11,1997.
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The Unit 2 activities performed during the week of May 19,1997.
b.
Observations and Findinas
Overall, PECO NMD personnel conducted the HCU work well. Supervisor
involvement was generally apparent, and procedures and work controls were
adequately used. The inspector noted that the work was generally well-planned and
controlled. Control room staff displayed good command and control, and conducted
effective briefings on several occasions to ensure that the numerous control rod
movements to support the maintenance and post-maintenance testing, were
performed without error. Good coordination and communications were evident
between maintenance technicians, operators, and reactor engineers.
Maintenance technician performance was generally good. However, during the
activities several issued developed:
Inadvertent scram of one control rod - Unit 2
On May 22,1997, technicians pulled fuses for the wrong control rod, resulting in
an inadvertent single control rod scram. The technicians were directed by the shift
supervisor to re-install the fuses. Operators correctly entered the applicable off-
normal procedure and determined that no thermallimits were exceeded as a result
of the event. The rod was returned to its original position.
PECO determined that the technicians had pulled fuses in the wrong fuse panel
during the application of a clearance. Specifically, they were to pull fuses for
control rod 26-11, which corresponded to the 14th fuse row down in panel
2HC068. Instead, they pulled the fuses for control rod 18-59, which was the 14th
row in panel 2CC068. Although the task involved double verification (one
technician observing another), the second technician's verification focused on the
14th row and missed the check of the correct panel.
PECO found that this event revealed weaknesses in double verification techniques,
attention to detail, and supervisory oversight. PECO's immediate corrective actions
were to relieve maintenance personnel from performing HCU clearance work.
Operators have applied all subsequent HCU clearances pending formal assessment
under the performance enhancement process.
The inspector's review of this issue determined that poor labeling inside of the fuse
cabinets was also a contributor in this event. Inside the panel was a single label
that consisted of the control rod numbers cross-referenced to fuse row numbers
written with a grease pencil. Thus, the task of counting the row numbers added
difficulty to the job, and apparently allowed the technicians to lose focus on other
verification steps. Furthermore, the inspector noted that the event occurred late in
the workers' shift, when errors can be more likely to occur.
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There were several minor issues dealing with scram air header leaks during work on
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the SSPVs. Specifically, while PECO worked on these valves at Unit 3, the scram
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air header supply tubing upstream of the closed manual isolation valve was
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unsupported. During the maintenance, some of the upstream compression fittings
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came loose. This leak caused low scram air header pressure problems during the
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work. In one case, the operators quickly placed the backup air supply into service
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to prevent a low pressure condition. The pressure surge caused one on the control
rod scram inlet isolation valves to deform because of an improperly set over travel
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stop. The deformed isolation valve seat allowed air leakage, which resulted in a
fully withdrawn control rod drifting to the fully inserted position. PECO took
adequate actions to preclude the air leaks from the scram air tubing by designing a
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clamp that provided extra support to this section of tubing during the maintenance
activity, at Unit 2.
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insertion of a Wrong Control Rod
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-The reactor engineers had generally good control over the HCU activities and
coordinated with the operators and NMD crew to properly position the control rods
to be worked. The required post-maintenance testing, which included control rod
movement and scram testing was conducted very well. However, in one instance
the reactor engineer directed that the operators insert a control rod that, while it
was in the groups of rods to be worked, was not in the sequence for work at that
time. In this event, the reactor engineers had conducted the proper pre-insertion
core thermal limits analysis, which predicted no negative result from the insertion.
After the rod was inserted, the operators began to apply the maintenance clearance
and noted that the wrong rod had been inserted. The operators and reactor
engineer then completed a review of the issue and returned the control rod to the .
fully withdrawn position,
c.
Conclusions
The inspector concluded that the planning and coordination associated with the Unit
2 HCU on-line maintenance was generally good. However, leakage from
unsupported, disconnected air lines caused some operational problems. Also, a
work sequencing error by the reactor engineer resulted in inserting the wrong
control rod. Control room staff exhibited very good control of post-maintenance
testing. Although maintenance technician performance was generally good, an error
resulted in pulling the fuses for an incorrect control rod. This event revealed
weaknesses in double verification techniques and supervisory oversight.
M1.2 E-3 Emeraency Diesel Generator Maintenance
a.
Scope (62707)
The inspectors reviewed the planned maintenance and post-maintenance testing
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performed on the E-3 emergency diesel generator (EDG) during the period May 12-
23,1997.
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b.
Observations and Findinas
The inspectors observed several planned preventive and corrective maintenance
actions completed on the E-3 EDG. Overall, the work was effectively planned and
implemented. The inspectors found that maintenance was performed according to
approved procedures and technicians were knowledgeable of their work
assignments.
1
One of the significant activities involved the discovery of an out-of-specification
crankshaft strain measurement. During this outage, PECO began using a new
evaluation criteria for crankshaft strain measurements. The out-of-specification
measurement indicated that the generator could be set too high for alignment with
j
the diesel engine. Measurements on the diesel engine #14 bearing provided
additional data that indicated a small alignment problem. After evaluation and
consultation with the vendor, PECO determined that the generator was set too high
l
for the engine. PECO lowered the generator by removing shims on the bedplate.
I
Technicians also replaced the #14 diesel engine bearing because of out-of-
specification measurements, although no sign of bearing damage was evident.
These actions returned the crankshaft strain measurement to within specifications.
PECO confirmed that the small misalignment had not affected EDG operability.
Because of the additional maintenance associated with the bearing replacement, an
extended post-maintenance test of the EDG was required.
A number of minor problems caused delays in the post-maintenance testing. First,
on May 20, operators observed a high crankcase vacuum reading and shut down
the EDG. Technicians found that the exhaust eductor restricting orifice area had
opened up. Several Icaking fuel injectors were also identified, including some that
had been replaced during the maintenance period. PECO noted that there was some
confusion over the leakage acceptance critorion, and was pursuing the cause of the
leaking injectors with the vendor. Also, an intermittent ground caused some delays
and was traced to a fuel rack reset mechanism limit switch. Finally, the EDG lube
oil standby circulating pump, which helps keep the lube oil warm, tripped repeatedly
between EDG test runs. PECO considered this to be a repeat maintenance / rework
problem because the pump was worked and reworked during the maintenance
period. PECO was evaluating the cause of the repeat maintenance problem at the
end of this inspection period. Despite the problems during the post-maintenance
testing, PECO contingency planning was adequate to allow the problems to be
resolved in a generally timely manner. The problems were entered into the
corrective action process for tracking and causal analysis,
c.
Conclusions
The inspectors concluded that, overall, the maintenance outage on the E-3 EDG was
effectively planned and implemented. PECO corrected an out-of-specification
crankshaft strain measurement by lowering the generator. Several minor problems,
some of which were maintenance rework issues, caused delays in the post-
maintenance testing and restoration to an operable status. PECO contingency
planning allowed the problems to be resolved in a timely manner.
!*
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8
M1.3 E-1 Emeraency Diesel Generator Maintenance
a.
Scooe (62707)
A review by the inspectors, of the planned on-line maintenance and testing, for the
E-1 emergency diesel generator (EDG) overhaul started on June 1 and continued
i
through the end of the inspection period to verify that maintenance activities
performed provided for assurance of reliable and safe operation.
b.
Observations and Findinas
The inspectors observed several effectively planned and executed E-1 EDG
1
preventive and corrective maintenance actions. Maintenance personnel performed
work according to approved procedures, and technicians were aware and
knowledgeable of their work assignments. Maintenance supervision coordinated
well with system managers.
The new crankshaft strain measurement criterion resulted in significant work
activity for PECO during the overhaul. Another generator move would be required
during the overhaul. After verification of the out-of-specification crankshaft stain
measurement and supporting measurement from the #14 bearing, maintenance
moved the generator in the downward direction by removing shims. PECO lowered
the generator in the same method used during the successful overhaul of the E-3
EDG. Also a lateral generator move to bring the crankshaft stain and bearing
measurements within specification was required after consultation with the vendor.
The licensee chose to calculate the required move, but to move the generator
incrementally, thereby minimizing the flexing on the generator and diesel shafts.
During post-maintenance testing to return the E1 EDG to service, PECO identified a
field flash circuit relay (K1) reset problem that needed to be addressed before
operability of the El could be assured. Specifically, the K1 relay wouldn't reset
following an EDG run. K1 must operate to reset the field flash after an engine
shutdown to ensure proper operation of the emergency generator on a subsequent
automatic start. PECO appropriately concluded that this could affect EDG
operability and proceeded with a relay replacement. A satisfactory reset verification
of the K1 relay on the other EDGs, performed by PECO, addressed the possibility of
a generic failure. This unexpected relay replacement caused a delay of the return of
the E1 EDG beyond the initial schedule.
c.
Conclusions
The inspectors concluded that PECO had effectively planned and implemented the
maintenance outage activities on the E-1 EDG. PECO competently corrected the
crankshaft strain measurement and field flash relay problems.
!
M 1.4 Surveillance Activities
e
Standby Liquid Control System Surveillance Test - Unit 2
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a.
Scoce (61726)
4
':
The inspector observed the performance of surveillance test ST-O-011301-2,
?
" Standby Liquid Control Pump Functional Test for IST," on May 27,1997.
)
b.
Observations and Findinos
4
'
The inspector found that the Unit 2 quarterly standby liquid control (SBLC) system
'
surveillance, which verifies the operability and performance of the SBLC pumps and
check valves, was conducted adequately according to the surveillance test
.
procedure. The inspector observed operators performing preparations, valve
,
i
positioning, and other actions specified by the test procedure. Operators
appropriately held a pre-job briefing to ensure that test participants were aware of
the actions necessary to restore the system should it be needed to perform its
'
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safety function.
!
"
The operators identified a number of minor weaknesses in the recently revised (April
1997) surveillance test procedure. For example, some revised steps requiring
operators to take pump oil samples caused considerable delays in the conduct of
the test. The steps specified that oil samples be taken immediately following the
running of the pumps. However, at that time, some of the oil had not yet returned
to the reservoir, and the oil level was about the same level as the sample port. This
resulted in considerable difficulty in drawing the oil sample, in addition, operators
identified confusing directions as to whether to mark as "not applicable" procedural
sub steps or complete the steps of the procedure.
The operators also recognized some opportunities for improvement in their
familiarity with the test procedures and equipment. The inspector observed that
when problem areas arose, the operators appropriately consulted with shift
supervision for clarification and guidance.
The inspector noted that a review of the SBLC pump standby oillevel band or
consideration of additional procedural guidance to the operators may be warranted.
This is based on the fact that following the pump runs and oil samples, the
operators needed to add oil to bring the level up to the low end of the band.
However, by the next day, the oil level had risen to above the maximum level mark.
This discrepancy was brought to the attention of the operators.
c.
Conclusions
The inspector concluded that the quarterly SBLC system surveillance test was
performed adequately in accordance with test procedures. Operators identified
some minor weaknesses in the recently revised surveillance test procedure and
identified some opportunities for improvement in their femiliarity with the test
procedure and equipment. The inspector observed a weakness associated with the
methodology or procedural guidance for taking pump oil samples.
M4
Maintenance Staff Knowledge and Performance
.
.
10
M4.1 incorrect Fuse Removal Durina Local Leak Rate Testina
a.
Scoce (62707)
The inspector reviewed an issue that occurred during thr. Unit 2 refueling outage in
September 1996 as it related to human performance dur;ng localleak rate testing
(LLRT).
b.
Observations and Findinas
On September 20,1996, with Unit 3 shutdown, the maintenance personnel
mistakenly pulled the primary containment isolation system (PCIS) inboard and
outboard mechanical vacuum pump trip logic fuses (F6 fuses in panels 20C041 and
20C042) while working on a localleak rate test activity. This would have caused a
these valves to close if they had not been closed at the time. These valves were
not primary containment isolation valves, but do receive a PCIS signal to close in
response to a main steam line isolation, to limit release of radioactive contamination
from the condenser.
The inspector discussed this activity with a maintenance supervisor with respect to
the expectations related to fuse manipulations by maintenance personnel. Also, the
inspector verified the persons involved had received some training in LLRT and the
associated fuse removal. During LLRTs and post-maintenance testing on the main
steam line sample valves, fuses were to be removed and reinstalled several times.
The lead maintenance technician doing the tast had installed the fuses, but he had
other maintenance personnel remove the fuseo. During this process they removed
the wrong fuses.
PECO initiated a PEP (10006104) to investigate the, cause of the wrong fuse
manipulation. During this evaluation, PECO identified several causes including less
than adequate use of self-check methods, the workers selected for the ta.sk may not
have been familiar with fuse labeling and wire labelir:g conventions used at PBAPS,
and an inadequate pre-job briefing,
c.
Conclusions
The maintenance work group personnel error of pulling the wrong fuses during
.
preparations for containment isolation valve LLRT work resulted in a loss of
electrical feed to the mechanical vacuum isolation valve's, but no actual valve
motion resulted. These valves are not primary containment isolation valves and the
plant was shut down at the time. Therefore, this problem had no safety
consequences. However, it represents pocr maintenance activity performance.
Unfamiliarity with fuse removal and self-checking methodologies by the personnel
involved, and an inadequate pre-job briefing contributed to the event.
M8
Miscellaneous Maintenance issues (92902)
.
.
11
M8.1 Review of Instrument Channel Functional Test Practices
a. Scope
During the period, the inspectors completed a review of how PECO implemented the
improved standard technical specification (ITS) definition of an instrument channel
functional test (CFT). The NRC staff became aware of this issue when inspectors
and PECO ISEG identified some possible interpretation problems with the ITS
definition. ISEG Report 96-29, " Review of Channel Functional Tests for Steam Leak
Detection Instrumentation at Peach Bottom Units 2 and 3," questioned whether
channel functional test practices for the HPCI and RCIC system complied with
Peach Bottom ITS requirements. The inspectors reviewed the issue with assistance
from the Office of Nuclear Reactor Regulation (NRR).
b. Obsgrvations and Findinas
ISEG and the inspectors found that Peach Bottom did not test all channel relay
contacts that input to the logic circuits to verify logic circuit contact operability as
part of the CFT. ISEG questioned if by excluding some of the contacts believed to
be within the scope of the channel, PECO was complying with the ITS requirements
for channel system functional tests.
The NRC found that the issue revolved around the differences in the wording of the
ITS and the old custom TS definitions of a CFT.
Old Custom TS CFT definition:
Prior to implementing the ITS, Peach Bottom operated with custom TS. The
previous TS contained a definition for " Instrument or Channel Functional Test,"
which stated:
"An instrument or channel functional test means the injection of a simulated signal
into the channel or instrument as close to the primary sensor as practicable to verify
the proper instrument channel response, alarm and/or initiating action."
NRC review of this definition found that the intent was to demonstrate channel
operability by verifying that at least one contact has changed state. If the design of
the channel was such that additional contacts associated with channel relays can
be verified operable, then it is desirable to do so as part of the CFT. However, if
the design is such that jumpering or lifting of leads is necessary for verifying
contact operability, then the CFT need not include these additional contacts. These
contacts would be included in the ITS required logic system functional test (LSFT).
,
.
12
A CFT is defined in the ITS as follows:
"A CHANNEL FUNCTIONAL TEST shall be the injection of a simulated or actual
signal into the channel as close to the sensor as practicable to verify OPERABILITY,
including required alarm, interlock, display, and trip functions, and channel failure
,
trips. The CHANNEL FUNCTIONAL TEST may be performed by means of any series
of sequential, overlapping, or total channel steps, so that the entire channel is
tested."
In this ITS definition, the "and/or" language from the previous TS was removed and
the term " required" was added to characterize the intent of "and/or." However,
that language change may cause the interpretation that all relay contacts associated
with a channel are required to be verified operable during a CFT.
The purpose of CFT was not intended to test the change of state of all channel
relay contacts, in the ITS conversion, the licensee considered the definition change
as administrative and the staff accepted the new definition because this change in
definition was not intended to increase the scope of CFT. Testing of all contacts-
during CFT is considered unnecessary since change of state of one contact could
infer change of state of all other contacts associated with a relay. The operability
of all contacts is assured during LSFT. The probability of failure of individual-
contacts, based on past history, is low.
c. Conclusion
The staff has not required licensees to monitor all relay contacts during the more
frequent CFT. Thus, the staff concludes that PECO is satisfying the objectives of
,
the Peach Bottom ITS CFT requirements. The inspectors also concluded that ISEG's
review had led to an improved understanding of regulatory requirements in this area.
In order to avoid further misinterpretation of the TS requirements for CFT, the NRC
staff has undertaken an initiative, with industry representatives, to develop a more
clearly worded CFT definition in the ITS.
111. Enaineerina
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E1
General Engineering Comments
a.
Scone
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The inspectors reviewed the general support provided by engineering to day-to-day
operations of both units.
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b.
Conclusions
Engineering department management and system mangers provided very good
support to the E1 and E3 maintenance outage, particularly in the review and
dispositioning of the crankshaft strain issues discussed in Sections M1.2 and M1.3
above.
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IV. Plant Support
F2
Status of Fire Protection Facilities and Equipment
F2.1
Material Condition Insoection and Eauioment Inventories
a.
Scone
The inspector conducted a walkthrough of the facility with PECO's fire protection
,
staff for PBAPS. During the walkthrough, the inspector noted the material condition
of the fire protection equipment, the material condition and housekeeping of the
facility, the readiness of the fire protection equipment for use, and discussed the
upgrades of the equipment, which had been completed. The inspector conducted
an inventory check of the fire brigade locker and the hose cart house at the south-
east corner of the facility. In addition, the inspector reviewed the unified log to
,
evaluate the out-of-service times for the Cardox systems during the current year.
The inspector also reviewed the records of the following fire protection equipment
functional tests:
-e
ST-O-37C-360-2, Rev. 4, " Motor Driven Fire Pump Operability Test,"
conducted April 2,1997.
ST-O-37D-370-2, Rev. 6, " Diesel Driven Fire Pump Operability Test,"
conducted March 27,1997.
RT-F-37G-392-2, Rev. 2, "E-2 Diesel Generator Cardox System Simulated
Actuation and Air Flow Test," conducted June 5,1996.
RT-0-100-990-2, Rev.1, " Participants Record Docuruentatbn for STs, RTs,
Clearances, and Check Off Lists (COLs)."
b.
Observations and Findinas
The material condition and housekeeping in the facility were excellent. No build up
of trash or combustible material was noted in the plant during the tours. The fire
fighting equipment in the plant, both manual and automatic, was observed to be in
good repair and in an excellent state of readiness. Portable fire extinguishers were
in place where required, and of a type appropriate for the hazards in the area.
The fire protection alarm panels in the control room have been replaced with newer
models. The newer models allow for acknowledging and resetting alarms
individually. This improvement will allow control room operators to acknowledge an
alarm without masking any additional alarms which might occur.
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Some oil and carbon blowout was noted in'the vicinity of the exhaust system and
underneath the turbochargers of the emergency diesel generators (EDGs). The
inspector identified this condition.to PECO personnel, who initiated actions to
d
commence cleanup. Due to the constricted access to the area below the
turbocharger end of the engine, the CO, flooding system for the room is required to
{
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be disabled while work is in progress to limit the safety hazard to personnel. The
j
inspector verified that appropriate impairments and fire watches were poeted.
$
,
The testing of the fire protection equipment was controlled by approved procedures,
which implement the testing requirements of the Technical Requirements Manual.
The fire pumps are tested monthly, on a staggered basis. This results in a fire pump
being tested approximately every two weeks. During the review of the completed
^
functional tests on the fire pumps, the inspector noted that Section 10.0,
" Participants Record," had not been completed. This section provides the initials
and printed names of the individuals who participated in the test. When the
j.
inspector questioned this, PECO personnel provided a copy of " Participants Record
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Documentation for STs, RTs, Clearances and Check Off Lists," RT-0-100-990 2,
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Rev.1. This procedure is performed annually, remaining open for the entire year, to
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permit operations personnel to omit their names and initials in STs, RTs, tagging
orders and check lists, which they conduct on a daily basis.
.
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The testino of the diesel generator CO, flooding sysmn was conducted once every
eighteer
ionths. The test verifies the system activation time and discharge
'
s
duration. F.w test also verifies that the flow path is unobstructed by performing an
air blow from the CO, discharge header to the discharge nozzles. During the
conduct of the test, the appropriate impairments and fire watches are implemented
by the procedure.
The review of the out-of service times of the Cardox systems, based upon the out-
of-service and return-to service times recorded in the unified log, indicated that of
the approximately 2,760 hours0.0088 days <br />0.211 hours <br />0.00126 weeks <br />2.8918e-4 months <br /> in the year-to-date (at the time of the inspection),
Cardox systems had been out-of-service for approximately 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> due to system
problems. :n contrast, Cardox systems had been removed from service for
personnel p.otection approximately 490 hours0.00567 days <br />0.136 hours <br />8.101852e-4 weeks <br />1.86445e-4 months <br />. The inspector considered this to be
indicative of highly reliable fire suppression systems.
c.
Conclusions
The inspector concluded, based upon the observed conditions, the results of the
reviews of the operability tests, and evaluation of Cardox system outage times that
the fire protection equipment at the PBAPS is in good repair and is ready to perform
its intended function.
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1
F3
Fire Protection Procedures and Documentation
F3,1
Fire Protection Procedure Reviews
)
,
.
The inspector reviewed the procedures controlling fire protection activities at the
l
facility to determine what management controls had been developed to prevent fires
and rapidly suppress any fire which might occur. The following specific procedures
i
were reviewed:
I
e
Nuclear Generation Group Policy No. NP-FP-1, Rev. O, Fire Protection
e
FF-01, Rev. 3, Fire Brigade
4
2
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A-C-920, Rev. O, Nuclear Generation Group Fire Protection Program
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AG-CG-012.02, Rev.1, Control of Combustible and Flammable Materials
!
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AG-CG-012.01, Rev. 2, Actions for Fire Protection impairments
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AG-CG-012, Rev. O, Hot Work Guideline
,
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NE-C-250, Rev.1, Fire Protection Review
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NE-C-250-2, Rev. O, Fire Protection Review Checklist
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HZ C-5-4, Rev.1, Safety Storage Equipment
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HZ-C-5-1, Rev.-1, Request for Permanent Chemical Storage
b.
Observations and Findinas
The fire protection policy statement and administrative procedures have been
revised and reissued, with new nurnbers, since the last fire protection inspection.
This effort was conducted, in part, to provide for standardization between PECO's
nuciear stai!ons at Limerick and Peach Bottom. This will aid in ensuring compliance
with requirements as personnel move between the sites in response to outage and
specialty maintenance requirements. The exception to this is FF-01, " Fire Brigade,"
which remains site specific for Peach Bottom.
A C-920, Nuclear Generation Group Fire Protection Program
This procedure is a common procedure between the stations and provides the broad
administrative guidelines for fire protection and extends the concept of defense-in-
depth to fire protection activities. This procedure provides the overall goals to be
achieved by the fire protection program, and specifies lower tier procedures for
carrying out specific functions within the program. This procedure also provides
general guidance for responding to fires, control of hot work, and review of all work
activities to ensure that protection against fires has been included in the work
planning.
AG-CG-012, Hot Work Guideline
This procedure minimizes potential fire hazards by specifying a hot work permit
system to control work involving ignition hazards. The procedure specifies those
areas wherein a hot work permit is required, lists examples of typical ignition
sources, and provides instructions for generating hot work permits. This procedure
also assigns responsibilities to the personnel involved, such as the hot work
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opf rator, fire watch, work supervisor, and industrial risk management (IRM)
penonnel. The procedure also requires that communications equipment, such as
j
7t page, plant phone, or portable radio, is available in the immediate vicinity of
r
the hot work area. This last requirement was added as part of the corrective
actions for the fire, which occurred at the site on August 10,1994 (see Section F
7.2).
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AG-CG-012.01, Actions for Fire Protection impairments
This procedure providos guidance for reporting, tracking, and ensuring restoration of
fire protection systems and features required by the NRC or American Nuclear
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insurers (ANI). The requirements of this procedure include designation and
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scheduling of fire watch personnel by the work supervisor who is responsible for
i
the work which will cause the impairment (security force provides fire watches for
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impairments caused by broken equipment). Shift management is responsible to
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perform a review of compensatory actions prior to issuing the impairment, and the
work group is responsible for implementing the compensatory actions before taking
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the equipment out of service. For those situations where there are questions
i
regarding the effect on fire protection systems or features, IRM provides resolution.
l
AG-CG-012.02, Control of Combustible and Flammable Materials
1
,
This procedure establishes controls to minimize the risk of damage or loss due to
the use of flammable or combustible materials in the plants. The procedure
provides for the reporting of fire hazards, control and storage of combustible
materials, designation of combustible free zones to prevent the spread of fire,
l
controls on smoking in the faci!ity, and disposal guidelines for combustible
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materials.
<
NE-C 250, Fire Protection Review
This procedure provides guidelines for conducting fire protection reviews of
modification packages, and for completing the fire protection checklist. The
procedure also establishes threshold levels of changes in combustible loadings,
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which require prior review by the fire protection / safe shutdown (FP/SS) branch prior
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to implementation of the modification. Changes less than these threshold values
are reported to the FP/SS branch for incorporation into controlled documents.
.
FF-01, Fire Brigade
4
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This procedure provides guidelines for responding to a fire at the station, both
within and without the protected area. Responsibilities for specific personnel are
.
given, and requirements for brigade response are listed. The procedure is
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considered to be general guidance, and the brigade leader is permitted to adjust the
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actions of the brigade to suit the conditions encountered during each individual
response.
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c.
Conclusions
Based upon the results of the procedure reviews, the inspector concluded that the
fire protection procedures provide good guidance and controls to prevent fires,
maintain the fire fighting capabilities of the organization, and respond to any fires
which might occur.
F5
Fire Protection Staff Training and Qualification
F5.1
Fire Briaade Trainina
a.
Scope
The inspector observed a fire brigade training session for health physics personnel;
reviewed lesson plan PHPCT-94-02C, Rev. No. 001, "HP Introduction to Fire
Brigade Response," dated April 19,1995; reviewed the documentation of the first
quarter 1997 fire brigade meeting; computerized records of fire brigade training for
the past two years; and Issue Evaluation Report 10002682, " Unit 2 Vent Stack Fire
and Unusual Event Declaration," dated December 19,1994.
b.
Observations and Findinas
As discussed in Section F7.2, during the vent stack fire of August 1994, some
confusion existed regarding the chain of command due to the presence of senior
operations management personnel on the refueling floor. To alleviate the problem,
PECO training for fire brigade responders emphasizes that the fire brigade leader is
the on-scene commander at the fire. All other personnel at the scene are under the
direction of the brigade leader.
The training focuses on fire field activities and fire response command structure.
The training indicates that the duties of the health physics (HP) technician
responding to the scene of the fire are the same as those normally performed,
namely evaluating the radiological hazards and proposing protective measures for
the personnel involved.
i
The training records for the fire brigade members and fire brigade leaders are
maintained in the plant information management system (PIMS). The training status
is available for viewing to all system users, but can only be changed by the
authorized person. Having the status available on PIMS to all users makes it easy
for the shift personnel to determine training and qualification status when making
fire brigade assignments.
c.
Conclusions
Based upon the review of lesso7 plans, observation of training in progress, and
review of the computer printout of training and qualification records, the inspector
4
concluded that the training is sufficient to ensure that an adequate pool of trained
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18
and qualified personnel are available to staff the fire brigade. In addition, having the
training and qualification status available on PIMS aids in making assignments to fire
brigade positions and determining the training needs.
F.5.2 (OcenIFl 50-277/278/97-04-Oli Fire Watch Trainina Revision to Fire Protection
Trainina Lesson Plans
a.
Scope
The inspector reviewed computerized records of fire watch training, issue evaluation
report 10002682, " Unit 2 Vent Stack Fire and Unusual Event Declaration"; lesson
plan MCTR-1075, Rev. 2, " Hot Work (Ignition Source) Firewatch Training"; and
lesson plan MCTJ-0035R, " Portable Fire Extinguisher Usage Requal."
b.
Observations and Findinas
The inspector's review of the firewatch training lesson plan determined that the
training covers the responsibilities of the ignition source fire watch. The training
also stresses that the firewatch should be actively engaged in setting up the hot
work area to ensure that all sparks, slag, and molten metal are contained in the hot
work area.
Issue evaluation report 10002682 reviewed the fire which occurred in the Unit 2
plant vent stack on August 10,1994, for lessons learned. One of the identified
problems was the lack of communications equipment on the building roofs. The
nearest plant communications equipment available to the firewatch at the scene of
the fire was approximately 75 feet down the scaffolding and across the turbine
floor. The future corrective actions for this event included revising the hot work
firewatch lesson plan to include lessons learned. The revision of the lesson plan
should have included the need for communications equipment in the immediate
vicinity of the hot work area, including the use of a portable radio if plant page or
phone are not available. Instead, the lesson plan still reads that the firewatch is
responsible to know "... the location of the nearest page or phone, and fire alarm
pullstation." The issue evaluation report is also not listed in the references section
of the lesson plan. When these deficiencies were identified to the fire protection
staff, actions were initiated to make appropriate changes to the lesson plan. This
will require action by the Limerick generating station fire protection staff, since they
are responsible for this lesson plan. This revision will be verified during the next fire
protection program review (IFl 50-277, 278/97-04-02).
'
c.
Conclusions
Based on the review of the lesson plans and the review of the computer printout of
training records, the inspector concluded that the training is sufficient to ensure that
qualified and knowledgeable personnel are available for performing hot work fire
watch duties. In addition, having the training records available on PIMS ensures
that work supervisors can easily determine the qualification status and training
needs of personnel in their group.
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F7
Quality Assurance in Fire Protection
F7.1
fire Protection Proaram Audits
a.
Scope
The inspector reviewed the reports of aud.ts of the fire protection program, which
have been conducted by PECO since the last inspection to determine whether the
audits have been effective in identifying deficiencies and initiating corrective
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actions. The review included the following audits:
!
Assessment Report No. A0808620, " Assessment of Fire Protection Program,
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MAP Area D02," dated December 15,1994.
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Assessment Report No. A0900835, "PBAPS Triennial Fire Protection
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Program, MAP Area D02, Rev. 8," dated September 25,1995.
4
Assessment Report No. A0967111, "PBAPS Annual Fire Protection Program
e
j
Assessment, MAP Area D02, Rev. 9," dated July 17,1996.
i
e
.PEPlssue 10005802, " Deterioration of CO, Extinguisher Hoses," initiated
i
j'
June 20,1996.
b.
Observations and Findinas
!
l_
The audits covered the entire scope of the fire protection program, including
i
administrative controls, the status and condition of fire protection equipment, fire
l
brigade training, corrective actions and self-assessment activities. Outside expertise
'
was used in the performance of the triennial program assessment. The assessment
j
plans were reviewed and concurred in by the Nuclear Review Board (NRB). In
3
several cases, members of the NRB requested the assessment team to look into
l
specific issues, such as ALARA considerations in the placement of continuous fire
j
watch stations and priority for clearing fire system impairments.
l
The audits generally concluded that the program was being conducted in a safe and
j
effective manner. The audits determined that the number of outstanding
i
j
maintenance items against fire protection equipment had decreased over the three
L
year period.
$
On the two occasions that equipment deficiencies were identified during the audit
'
team's facility walkdowns, appropriate corrective actions were initiated. One issue
was related to six CO, fire extinguishers being identified as having cracked hoses-
during the walkdown. Plant staff took immediate actions to determine the extent of
j
the condition, and all 21 deteriorated hoses were replaced. Subsequent review,
'
conducted under PEP issue 10005802, initiated June 20,1996, determined that the
hoses were removed from the CO, extinguishers when the extinguishers were sent
3
out for hydrostatic testing. The hoses were placed onto extinguishers, which were
J
being returned from testing without the complete inspection required by the
{
procedure. The use of procedures by the particular workgroup involved was
i
reviewed to ensure there were no other procedural compliance problems. The other
j
issue was related to the outside fire brigade equipment cage having two sets of
1
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.
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_ - - , - - . _
-.
_
.. -
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_
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20
turnout gear missing. An action request was issued to replace the missing
equipment. In addition, PECO established a tracking system to determine if an
adverse trend developed regarding missing fire protection equipment.
c.
Conclusions
Based upon the audits finding only minor deficiencies, and the plant staff taking
corrective actions to address the findings, the inspector concluded that the audits
were effective in identifying problems and causing corrective action to be taken.
F7.2 Review of Auaust 10.1994 Vent Stack Fire Corrective Actions - Unit 2
a.
Scone
The inspector reviewed the circumstances surrounding the August 10,1994, fire
on-site to determine what conclusions PECO had drawn with regard to the cause of
the fire, and what actions PECO initiated to improve the performance of the fire
fighting organization as a result. This was identified as the only fire in the facility
since the last inspection.
The inspector reviewed the following documents during the course of the review:
e
NRC Combined Inspection Report 50-277/94-13 and 50-278/94-13, dated
September 19,1994.
e
Licensee Event Report (LER) 50-277/94-07," Secondary Containment
Breached to Fight Fire," dated September 8,1994.
e
Performance Enhancement Program (PEP) Issue 10002682, " Fire in U/2 Rx
Building Roof Vent Stack Caused by Weld Spark," initiated August 10,1994.
e
issue Evaluation Report 10002682, " Unit 2 Vent Stack Fire and Unusual
Event Declaration," dated December 19,1994.
b.
Observations and Findinas
The circumstances which led to the fire are briefly described in NRC Inspection
Report 50-277&278/94-13. During modification work to upgrade the Unit 2
ventilation stack radiation monitor, new pipe hangars were required to be attached
to the structural steel for the vent stack. During welding, sparks escaped the
materialintended to contain them and ignited insulation and/or bird nests within the
stack's double wall. Attempts by the fire watch to extinguish the fire with a dry
chemical extinguisher were not successful, and the fire brigade was activated. The
fire brigade opened a hatch in the roof of the reactor building to bring a hose to the
scene. The fire was rapidly extinguirhed using a hose stream.
During their review of the event, PECO identified two contributing causes for the
fire and twelve extraneous conditions adverse to quality (ECAQ). The contributing
causes were determined to be lack of understanding of the requirements for staging
the area on the part of the work supervisor, and the fire watch being provided with
an incorrect type of extinguisher. The ECAQs included confusion regarding who
,
.
21
was in control at the scene of the fire due to the presence of senior operations
management personnel, lack of plant communications equipment at the hotwork job
site, misunderstanding on the part of the work supervisor regarding what
constituted the " work area," and the fire area not being covered by prefire strategy
plan, among others. Corrective actions were planned and carried out for all of the
identified problems. These actions are discussed further under Section FS, " Fire
Protection Staff Training and Qualification."
c.
Conclusions
Based on the corrective actions carried out for the observed deficiencies during the
program audits and fire fighting activities on August 10,1994, the inspector
concluded that PECO is effectively identifying and correcting problems with fire
protection activities.
F8
Miscellaneous Fire Protection issues
F8.1
Conformance to Uodated Final Safety Analysis Report Description
a.
Scoce
The inspector reviewed the Facility Operating Licenses DPR-44 and DPR-56; the
Peach Bottom Fire Protection Plan; the Peach Bottom Updated Final Safety Analysis
Report; and NRC Safety Evaluation Reports dated August 24,1994, September 16,
1993, November 24,1980, October 10,1980, September 15,1980, August 14,
1980, and May 23,1979.
b.
Observations and Findinas
The fire protection requirements were transferred from the PBAPS technicai
specifications to the UFSAR by Amendment No. 210 to License DPR-44, and
Amendment 214 to License DPR-56. The description is contained in the fire
protection plan, which is incorporated into Section 10.12 of the UFSAR by
reference. The inspector determined that the fire protection program conforms to
the description in the UFSAR.
During the inspector's walkthroughs of the facility, particular attention was paid to
fixed suppression systems and fire protection features described in the fire
protection plan and NRC safety evaluation reports. The fixed suppression systems
and other features described in the fire protection plan and safety evaluation reports
have been maintained in effect.
c.
Conclusions
Based on the inspector's observation of the fixed suppression systems and other
features described in the fire protection plan and the safety evaluation reports, the
inspector determiner.1 that the fire protection systems conform to the descriptien in
the UFSAR.
, _.
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_ _ _ _
_..
_
._
_ _
_ _ _ _ _
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_
0
22
R1
Radiological Protection and Chemistry (RP&C) Controls
R1.1 Radioloaical Controls (Proaram Chances)
a.
Scope (80750)
The inspector reviewed selected radiological controls program changes. Areas
reviewed included organization and staffing, facilities and equipment, and procedure
changes.
b.
Observations and Findinas
e
Organization, Staffing, Training and Qualification
PECO implemented a radiological controls organization change in late 1996
involving temporary assignment of the radiological engineering manager to the
radiation protection manager position. The acting radiation protection manager
(RPM) met applicable qualification guidance of Regulatory Guide 1.8. and was later
selected as RPM.
A new radiological engineering manager was selected. The individual met
applicable experience requirements, At the time of the inspection, the individual
was not familiar with applicable program procedures and current industry guidance
in the areas to be managed. Consequently, an experienced individual within the
radiological engineering group was temporarily promoted to acting manager pending
the completion, by the newly selected manager, of a familiarization program for
department procedures and industry standards.
An experienced individual was acting in the capacity as manager, technical support
in the absence of the incumbent due to the incumbent's temporary assignment at
another licensee's facility.
During a previous inspection, a new individual was selected to provide training of
station personnel in the area of radioactive material shipping. This individual
appeared to have limited experience and training in the area. The licensee provided
training to compensate for the individual's limited experience and knowledge in this
area. An experienced individual was used to provide training in the interim.
e
Review of Dosimetry Equipment
During a previous inspection, the inspector noted that the licensee was
encountering difficulty with the low energy beta response of its vendor supplied
dosimetry. Although the test results of the dosimetry met applicable national
testing criteria, the error associated with the test results was higher than the
licensee wished to accept. Consequently, the licensee recently changed its
dosimetry vendor. The inspector reviewed applicable information and determined
that the new dosimetry system met the requirements of 10 CFR 20.1501 relative to
accreditation by national testing standards. The new dosimetry was noted to
w
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23
. exhibit improved dosimetry performance (i.e., lower error) and was accredited in all
radiation test categories of the national standard.
o
Programs and Procedures
The licensee implemented hydrogen water chemistry and injection of depleted zinc
into the reactor coolant system. (See Section R1.2.b of this report.)
c.
Conclusions
No program changes were identified that reduced the effectiveness of the
'
radiological controls program.
1
!
No safety concerns or violations were identified.
,
I
R1.2 ALARA Proaram and Unit 3 Refuelina Outaae Plannina, Preparation. Emeraent Work
"
Control
a.
Scope (83750)
The inspector selectively reviewed various ALARA program elements and reviewed
the planning and preparation for the Unit 3 refueling outage, including control and
review of emergent work. The inspector reviewed records, discussed outage
planning, and observed activities to verify necessary planning, preparations, and
management support for the implementation of radiological controls. The inspector
reviewed lessons learned from previous outages to determine if they were
incorporated into planning and preparations for future outages.
b.
Observations and Findinas
The licensee continued to implement initiatives to reduce overall occupational
radiation exposure. The licensee recently implemented injection of " depleted zinc"
in order to reduce the dose rates attributable to production of Zinc-65. The licensee
had started injection of natural zine into the coolant in 1991 (Unit 2) to reduce
piping degradation. However, use of natural zine results in production of Zinc-65
which increases drywell radiation dose rates. The licensee implemented use of the
depleted zinc at both units on October 28,1996, and expects about a 25%
reduction in drywell radiation dose rates (recirculation piping) in 3 to 4 fuel cycles.
The licensee continued to implement other activities to reduce unnecessary
occupational exposure including use of robotics, hot spot reductions, permanent
shielding (e.g., scram discharge headers), station modifications, enhanced use of
video cameras, and improved timeliness of drywell shielding installation. The
licensee also installed station and component pictures on the station's local area
network for viewing during work planning. The licensee also enhanced its
benchmarking of individual job tasks and has obtained a " gamma camera" for use in
identifying elevated dose rate areas. The licensee develops reasonable occupational
exposure goals and meets those goals.
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24
The licensee also implemented ute of hydrogen water chemistry to limit system
degradation. Unit 2 was in a hydrogen water chemistry test mode at the time of
the inspection. Because this activity has the potential to increase ambient radiation
levels at various locations at the station, the licensee initiated a campaign in late
1996 to train all plant personnel on the activity and potential radiation dose rate
increases. 'The licensee developed applicable radiation dose rate limits at various
locations at the station (e.g., site boundary, controlled area, restricted area) to
ensure conformance with applicable regulatory requirements. The licensee plans to
j
perform a TLD study to evaluate dose rate increases.
,
c.
Conclusions
The licensee implemented an overall effective ALARA program. The inspector noted
very good ALARA plans for significant radiological work activities. PECO
implemented overall effective ALARA planning for the Unit 3 refueling outage
.
including emergent work.
!
No safety concerns or violations were noted.
R1.3 Internal Exoosure Controls
a.
Scooe (83750)
The inspector selectively examined the internal exposure control program. The
inspector reviewed records, discussed the program with cognizant personnel and
observed exposure control practices during tours of the RCA.
b.
Observations and Findinas
There were no recorded internal exposures during the past two years. The
inspector noted that the licensee's effective control of airborne radioactivity has
resulted in a substantial reduction in use of respiratory protection equipment.
Licensee data indicated respirator usage has declined from about 17,000 respirators
worn in 1991 to approximately 54 respirators worn in 1996, which included a
refueling outage. The licensee performed appropriate internal dose calculations and
DAC-hours were calculated and tracked, as necessary.
The inspector observed an individual being fit-tested for a respirator and checked
applicable fit test machine calibrations. The inspector noted that the printout from
the fit test machine referenced the incorrect national standard for calculation of
respirator fit factor. The licensee initiated a review of this matter.
The following area for improvement was identified:
The inspector selectively reviewed the program for estimating internal exposure to
transuranics (e.g., alpha emitters) which are not readily detectable by invivo
bioast,ay (e.g., whole body counting). The inspector did not identify any procedural
guidance for personnel to use to readily ascertain that if an intake of transuranics
,
O
25
may have occurred, properly estimate the intake by evaluation of breathing zone air
sample results, and confirm the intakes, as appropriate, by use of supplemental
invitro bioassays (e.g., fecal analysis) of samples collected in a timely manner.
c.
Conclusions
PECO implemented an effective internal exposure control program. However,
although the licensee did not have any current problem with transuranics, there was
no clearly defined program to perform internal dose assessment for these
radionuclides in the event of their appearance (e.g., following fuel failures).
No violations were noted.
R1.4 External Exoosure Controls: (Ocen) Violation 97-04-01: Inadeauste Controls over
Locked Hiah Radiation Door Kevs
a.
Scope (83750)
The inspector selectively examined the external exposure control program. The
inspector reviewed records, discussed the program with cognizant personnel and
observed exposure control practices during tours of the RCA and observation of
work activities. The inspector reviewed high radiation area controls and general
radiological posting, implementation of the radiation work permit program, and
implementation of the dosimetry program.
b.
Observations and Findinas
PECO continued to implement and maintain effective real time personnel exposure
control by use of an electronic dosimetry (ELD)/ access control system. The system
was set-up to preclude unauthorized individuals from signing onto invalid radiation
work permits (e.g., new permits or revised permits). Workers were observed to be
appropriately wearing dosimetry on their heads, as directed by radiation protection
personnel, when working in radiation dose rate gradients emanating from overhead.
The inspector noted that areas (e.g., high radiation areas, radiation areas) were
properly posted and locked (as appropriate). The inspector inventoried high
radiation area keys and noted all to be present and properly signed out if applicable.
The inspector verified workers performing work activities in high radiation areas
were properly signed-in on applicable radiation work permits. The inspector noted
that the licensee was providing neutron monitoring, in accordance with guidance in
NRC Regulatory Guide 8.14, of personnel working in neutron areas in preparation
,
for the Unit 3 outage. The licensee used a calibrated neutron survey meter, source
i
checked for the range of use,
j
During a previous inspection, the inspector noted that the licensee established
" standing" radiation work permits (RWPs) for areas as well as other types of RWPs
(e.g., special). The RWPs typically permitted certain defined work and also included
a statement (as work description) that other " approved work" was authorized. The
~
.
26
inspector questioned licensee personnel, including radiation protection control point
personnel, as to what constituted " approved work." The licensee subsequently
revised procedure HP-C-310 to include a definition as to what constituted
" approved work." However, the statement indicated the following caveat:
"For work which is not controlled by one of the work control documents specified
above, the verbal approval of a qualified RP technician is required."
The inspector noted that this provision did not discuss obtaining applicable
supervisory approval (e.g., work group supervisor, radiation protection supervisor,
or operations supervisor). The licensee initiated a review of this matter.
The following observations were made:
The licensee performed periodic calibration of each electronic dosimeter, but
e
there was no verification that the dose rate alarm would alarm at the pre-set
dose rate. The integrated dose alarm feature was, however, tested. The
licensee relied, in part, on the dose rate alarm feature to alert workers to
changing conditions, but relied on the integrating dose alarm feature to
conform with Technical Specification high radiation area monitoring. The
licensee initiated a review of this matter.
The inspector observed workers building scaffolding in the Unit 3
moisture separator area on May 5,1997. Workers were continuously
monitored by use of teledosimetry systems. The inspector noted,
however, that one worker's teledosimeter lost contact with the
remote monitoring station. However, no action was taken by the
radiation protection technician at the remote monitoring location to
attempt to understand the basis for the loss of contact (e.g., the
worker exited the approved work location). Further, the worker later
A;ted tir : area, walked past the technician, and despite the
teledosimetry still failing to make contact with the remote monitor,
the technician did not challenge the worker. The inspector considered
this a weakness in oversight of activities. The licensee initiated a
review of this matter.
Radiological signs / maps posted at the main radiological controlled area
access were difficult to read. The licensee initiated a review of this matter.
The inspector reviewed the control of high radiation area access door keys. The
following observations were made:
e
The keys for the cabinets where Level 1 and Level 2 high radiation area door
keys are kept were not on the key inventory list. The licensee added the
keys to the key listing.
There was no guidance regarding timeliness of updating of the high radiation
area key inventory list when a change to the list occurs.
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.
27
On January 30,1997, the RPM became aware that master keys that could
be used to open locked high radiation area doors at the Limerick and Peach
Bottom stations, were improperly controlled and in the possession of
unauthorized personnel between mid-1993 and November 1996. The keys
had been improperly made and distributed to fire protection personnel by the
licensee's corporate locksmith. The locksmith did not know the keys opened
high radiation area doors, in addition, the licensee's radiation protection
manager was unaware of the existence of the master keys maintained by the
corporate locksmith. Consequently, the inspector concluded that the
licensee's administrative key control program for locked high radiation areas
was not effeotive.
The inspector noted that the keys possessed by the fire protection personnel were
unauthorized and were not under the administrative control of radiation protection
personnel. As a result of the identification of the unauthorized keys, the licensee
took the following actions:
The security access authorization was removed for the individuals known to
possess the keys and other individuals within the work group who may have
received the keys.
The locksmith who provided the master keys was relieved of his duties.
The licensee initiated tours every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of locked high radiation area doors
to review for unauthorized entries.
The licensee initiated reviews for unplanned / unexplained radiation exposures
for the affected areas using a combination of key card data and knowledge
of the work areas of individuals known to possess the unauthorized keys.
No unplanned or unusual exposures were noted.
The licensee made a 1-hour report to the NRC on this rnatter.
The licensee's and an NRC security inspector's review identified that the
,
individual who made the keys did not realize the keys could be used at the
station and did not knowingly make the keys for unauthorized access
purposes at the station.
The licensee initiated an event report.
The licensee evaluated the occupational radiation exposures of the
individuals determined to have copies of the keys. The licensee determined
that the individuals did i.ot have unescorted access privileges to the
radiological controlled area, no unexplained or unplanned occupational
exposures occurred, and no uncontrolled high radiation area access by those
individuals occurred.
~
.
28
The licensee changed-out the lock cores of the affected doors (Level 1
doors-greater than 1,000 Mr/hr and Level 2 doors- greater than
10,000 Mr/hr) with special lock cores whose keys were only available to the
radiation protection group. These actions were completed by February 6,
1997. (Note: Due to access control concerns, the licensee has not changed
out the reactor water decant tank room (elevation 116' radwaste) lock or the
Unit 3 subpile room door. These locks are to be changed when the areas are
available for entry. Access to those individual doors are controlled by other
,
high radiation area keys that are controlled.)
I
The licensee placed the new keys (master keys and lock set changing keys)
e
under the administrative control of the radiation protection manager.
As discussed above, Technical Specification 5.7.2 requires that all door and gate
keys to a high radiation area with dose rates greater than 1.0 rem /hr at
30 centimeters from the radiation source be maintained under the administrative
control of radiation protection personnel. The inspector noted that unauthorized
keys to locked high radiation areas, not under the administrative control of radiation
protection personnel, were available to personnel (fire protection personnel) since
about 1993 through about November 1996. This is a violation of Technica!
Specification 5.7.2. (VIO 50 277,278/97-04-03)
c.
Conclusions
PECO implemented a generally effective external exposure control program. A
violation was identified associated with failure to administratively control keys for
R1.5 Control of Radioactive Materials and Contamination
a.
Scoce (83750)
The inspector selectively reviewed radioactive material and contamination control
practices including calibration and performance checks of survey and monitoring
instruments and the use of personal contamination monitors and friskers. The
inspector also evaluated personnel skin contaminations and skin dose assessment
methodology.
b.
Observations and Findinos
PECO implemented generally effective contamination control work techniques and
prompt correction and cleanup of contamination. At the time of this inspection, the
station exhibited approximately 2% of accessible floor areas as contamination areas
(excluding the drywell). Contaminated areas exhibited generally low levels of
contamination. Calibrated and response checked survey instrumentation was
available throughout the station. PECO tracked and trended personnel
contaminations for programmatic corrective action purposes. No personnel
contaminations resulted in any significant dose assessments.
~
.
29
i
The licensee continued to implement room-specific control of those areas of the
station which exhibited electron capture decay nuclides (e.g., Zinc-65) to provide
enhanced monitoring of material removed from these rooms and work performed
therein.
A number of individuals were observed at the RCA control point with low level
external contamination (principally clothing contamination) attributable to short-lived
particulate daughter products of fission gasses. The short-lived daughter products
(i.e., Cs-138 and Rb-88) adhered to personnel clothing, were readily detectable at
the licensee radiological controlled egress points by whole body friskers, and did not
represent an internal exposure control concern.
.
The contamination was attributable to low activity noble gas from the ventilation
.
system located on the 195-foot elevation of the Unit 3 turbine building. The gas
was believed by the licensee to be generated as a result of fission of residual tramp
uranium remaining on incore surfaces. The ventilation system draws air from the
"F" moisture separato: area of the turbine building. A large steam leak at a flange
in the area is reported to be the source of the gas. The licensce attempted to repair
i
the leak, but has not been successful and has decided to wait until the upcoming
October 1997 Unit 3 refueling outage to repair the leak.
The licensee plugged minor holes in the duct and indicated that ALARA cost benefit
analyser did not indicate repair activities (e.g., during reactor downpowering) were
cost beneficial.
The inspector noted the contaminations did not result in any significant personnel
exposures. The inspector expressed concern that the frequency of such
contaminations and the need for personnel to remain inside the RCA could
-
desensitize personnel to the need to continue to remain vigilant regarding personnel
contamination monitoring. The licensee's radiation protection manager indicated
articles have been published in employee newspapers to alert personnel to the
matter and the locations of the low activity noble gas daughter products. Potential
submersion doses, as indicated by the licensee's radiation protection manager, were
measurable by the licensee's TLDs.
The licensee experienced an average of about 200 personnel contaminations for
1995 and 1996. The licensee had, as of the date of this inspection, sustained
approximately 15 personnel contaminations for 1997, which, according to the
licensee, was 50% less than the previous year. Contamination of personnel by low-
level noble gas daughter products was not included in this data and was not tracked
and trended.
The inspector evaluated skin dose assessments previously performed by the
licensee for two individuals selected by the inspector who sustained skin
contamination. The following was noted:
The licensee used an industry code (VARSKIN MOD 1) to perform the skin
dose calculations. The inspector evaluated the dose using a revised code
(VARSKIN MOD 2) which included gamma dose contributions. The
inspector's independent dose calculations indicated approximately a 30%
.
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,
y
,
m-.
.,.
-
y
y
-__,
,
,
., - - - -
_
i
e
30
higher dose to the skin, assuming the contamination was a point source, as
'
also assumed by the licensee. The inspector indicated that although the skin
contamination resulted in generally low skin dose (well within applicable NRC
'
limits) the licensee should evaluate its skin dose methodology particularly as
it relates to determination of gamma dose and use of latest computer codes.
The licensee subsequently obtained VARSKIN MOD 2, performed similar
calculations and obtained essentially the same results. The licensee
subsequently developed a health physics job standard (HPJS - 9.8) to
provide guidance for use of the updated code and indicated that previous
skin dose assessments (for 1996 and 1997) would be reevaluated. The
licensee indicated the workers' doses (discussed above) would be updated.
On November 22,1996, the licensee became aware that a welder unit, released
from the radiological controlled area at the Peach Bottom station to an offsite
vendor facility, was found to contain contaminated tools. The welder unit had been
released from Peach Bottom Station on September 27,1996. PECO sent an HP
supervisor from Limerick Station to the vendor facility and found one contaminated
gasket, a pair of contaminated snips, and a contaminated screwdriver. The
maximum removable contamination found was 14,000 dpm (beta / gamma
contamination). The HP supervisor surveyed personnel and did not identify any
personnel contamination. Also, the HP supervisor surveyed floors, tables, storage
areas, tool boxes, waste cans and did not identify any contamination. The
supervisor collected the bag of material, and transported it back to Peach Bottom.
Technical Specification 5.4.1 requires that written procedures be established,
implemented and maintained covering the applicable procedures recommended in
Regulatory Guide 1.33, Appendix A, November 1972. The referenced appendix
recommends in Section G that procedures for control of radioactivity and for limiting
materials released to the environment be established. Licensee radiation protection
procedure HP-C-810, Revision 1, Radioactive Material (RAM) Control, specifies in
Section 7.5.1, that material to be released meet conditions specified therein (i.e.,
less than 1000 disintegrations per minute per 100 square centimeters
(dpm/100cm2) removable beta gamma contamination and less than 5,000
dpm/100cm2 total fixed and removable contamination.
The licensee determined that the welder unit had not been used since it was boxed
up and shipped from Peach Bottom on September 27,1996.
As interim corrective actions, the license initiated an internal review, suspended all
release of large machinery (e.g., welder units) from the RCA except with
supervisory approval, surveyed other welder equipment on site, and instructed
station HP personnel regarding the event.
The licensee took the following additional actions:
e
The licensee initiated an event report (PEP 1006341) for the matter and
initiated an investigation.
I
n
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. - - . - - -
. - . - . - - . - - - _ . _
- -
- - - .
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31
e
The licensee evaluated the reportability of the event and determined it was
not reportable.
e
The licensee established a staging area (early December 1996) on the turbine
building 116-foot elevation for survey of material. Material to be surveyed
was to be placed in the staging area for survey and immediately released
upon survey.
The licensee reviewed the adequacy of the general employee training
e
program relative to the event for potential enhancements and did not identify
any weaknesses.
The licensee took action to revise procedures for release of large objects as
e
expected during outages. The licensee was expected to complete the
'
revision by June 30,1997.
The licensee concluded that the non-surveyed welder unit, with the slightly
!
contaminated tools, was inadvertently released from the reactor building Unit 2
l
railroad door when it was placed next to a similar welder unit that had been
'
surveyed at the same location. The door was used following the outage to release
i
outage-related equipment and has since been closed.
The inspector noted that failure to implement procedures recommended in
Appendix A of Regulatory Guide 1.33,1972 is an apparent violation. The inspector
reviewed this violation with respect to the criteria for exercise of discretion outlined
l
in Section Vll.B.1 of the " General Staternent of Policy and Procedure for NRC
Enforcement Actions," (60 FR 34381; June 30,1995). The inspector noted that
even though the above issue was identified by the licensee, it did not appear to be
an issue that could have been prevented by a previous violation, it did not appear to
be willful, and corrective actions were taken as discussed above. The inspector
concluded the above matters constituted a licensee-identified and corrected
violation, which is considered non-cited, consistent with Section Vll.B.1 of the NRC
c.
Conclusions
PECO implemented a generally effective contamination control program. One non-
cited violation was identified regarding the release of a contaminated welding
machine.
1
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o
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32
R5
Staff Training and Qualification in Radiation Protection and Chemistry
R5.1 Radiation Workers /Radioloaical Controls Personnel
a.
Scope (83750)
The inspector reviewed the training and qualification records of a worker who was
fit-tested for use of respiratory protection equipment, the qualifications of the
radiation protection technician who fit-tested the worker, and reviewed the training
documentation and completion of required surveys by selected advanced radiation
workers. The inspector also reviewed the training provided radiation protection
technicians. The inspector evaluated the training and qualification of these
individuals relative to applicable technical specification requirements, procedural
requirements, and 10 CFR 50.120. The inspector reviewed training records and
discussed qualification criteria with cognizant personnel.
- b.
Observations and Findinas
PECO provided training and qualification, as appropriate, for the individuals selected
by the inspector. The licensee established and implemented a health physics
technician continuing training course plan. The licensee provided 102 hours0.00118 days <br />0.0283 hours <br />1.686508e-4 weeks <br />3.8811e-5 months <br /> of
training per technician for 1995,92 hours0.00106 days <br />0.0256 hours <br />1.521164e-4 weeks <br />3.5006e-5 months <br /> per technician for 1996, and developed a
j
course plan for 1997. The 1997 plan includes industry events, design basis
analysis and impact of hydrogen water chemistry.
c.
Conclusions
PECO provided training of radiological controls personnel providing respirator fit ~
testing, workers identified to wear respirators, and training of advanced radiation
workers. PECO was also providing continuing training to radiation protection
technicians.
R7
Quality Assurance in Radiological Protection and Chemistry Activities (83750)
R7.1 Radioloaical Event Reoorts
a.
Scope (83750)
The inspector selectively reviewed oversight activities for radiological controls. In
particular, the inspector reviewed PECO's evaluations and actions associated with
self-identified issues and concerns documented in its self-identification programs
(e.g., personnel contamination reports, radiological occurrence reports, performance
enhancement issues, quality assurance surveillance items, and industry audits).
-
.
33
b.
Observations and Findinas
The inspector reviewed selected licensee self-identified issues covering calendar
year 1996 and 1997 up to the time of the inspection. The inspector's review
indicated that the licensee took effective and timely action on self-identified
concerns. The inspector noted generally good oversight of activities.
The inspector reviewed various audits and surveillances and noted the use of
applicable industry standards as audit criteria. The licensee developed a program
(procedure HP-C-109) to periodically review the radiation protection program
content and implementation as outlined in 10 CFR 20.1101. Technical experts
were used to support the reviews. The licensee evaluated industry experience for
potential programmatic enhancements and developed a quarterly trend report using
numerous inputs on radiation protection performance for use in assessing the
effectiveness of the radiation protection program.
c.
Conclusions
PECO implemented an effective program for self-identifying and correcting self-
identified issues and concerns. No violations or safety concerns were identified.
R8
Miscellaneous RP&C Activities
R8.1 (Closed) Unresolved item 96-06-04: Review of Radioactive Material Storaae
Locations Versus Uodated Final Safety Analysis (UFSAR) Descriptions
a.
Scoce (83750)
During NRC Combined Inspection No. 50-277;278/95-27, (conducted
November 26,1995, through January 13,1996) and 50-277;278/96-06,
(conducted July 7,1996, through September 7,1996) the inspector reviewed the
conformance of the licensee's radioactive waste storage and processing facilities
relative to descriptions within the UFSAR. Insufficient review of this issue resulted
in an unresolved item.
b.
Observations and Findinas
During the inspection, the inspector met with cognizant licensee personnel and
discussed the actions taken on the earlier identified discrepancies as described
below,
e
The inspector noted that, relative to the liquid radioactive waste system, the
licensee identified lack of neutralization of the chemical waste tank contents
prior to transfer to radwaste floor drain sumps as indicated in UFSAR
Section 9.2.4.2.3. The licensee performed a safety evaluation (June 12,
1996) for the current mode of operation, did not identify any safety concerns
and initiated an engineering change request (May 3,1996) to update the
m.___
___ __._ _
e
34
The UFSAR did not contain any apparent specific information relative to
outdoor storage of radioactive materials / radioactive waste. However, a
licensee 10 CFR 50.59 evaluation for outdoor storage, identified several
outdoor radioactive material storage / staging areas, some of which were not
used. The 10 CFR 50.59 evaluation, performed for outdoor _ storage of
radioactive material, did not address the storage of sea van trailers behind
the 135' elevation of the radioactive waste building. The inspector noted
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that the trailer storage appeared to be well within restrictions on radiation
- dose rates presented in the 10 CFR 50.59 evaluation for other storage
locations. The licensee updated the 10 CFR 50.59 (May 5,1997) to identify
specific storage locations, removed unnecessary sea vans, and initiated an
engineering change request to update the UFSAR to reflect outdoor storage.
The inspector's review did not identify any apparent significant safety concerns
_
,
associated with the findings. The licensee took actions (e.g., Action Requests) to
, l
review the findings and update the UFSAR and applicable drawings, as appropriate.
The licensee indicated a UFSAR update would be submitted on or about
June 30,1997, to reflect the changes. The licensee took generic actions to train
appropriate personnel that station changes were to be processed through the
10 CFR 50.59 process.
j
The inspector noted that 10 CFR 50.59(a) states that the licensee may make
changes to the facility as described in the safety evaluation report without prior
commission approval provided that the change does not involve a change to the
technical specifications or an unreviewed safety question.10 CFR 50.9(b) requires
j
that records of the changes include a safety evaluation, which provides the basis
that the change did not involve an unreviewed safety question. Further,10 CFR 50.71(e)(4) requires that the UFSAR be updated to reflect the changes. The
inspector's review indicated 10 CFR 50.59 evaluations were not made for the
changes discussed above. The licensee subsequently took corrective actions as
described above. The inspector's review of the changes did not identify any
significant safety concerns,
i
Based on the above, the inspector noted that the failure to perform a 10 CFR 50.59
i
evaluations constitute violations of no safety consequence, and are being treated as
non-cited violations, consistent with Sections IV and Vll.B.1 of the NRC
Enforcement Pohcy.
c.
Conclusions
j
The licensee initiated action to update the UFSAR to reflect current practices for
storage and staging of low level radioactive / contaminated materialin the yard areas
of the station and operation of radioactive waste systems. Unresolved item
(URI 50-277; 50-278/96-06-04) associated with UFSAR discrepancies is closed. A
non-cited violation of 10 CFR 50.59 was identified.
,
- _ _ _ .
.
.
-
. _ -
..
.
. _ - -
__
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e
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R8.2 Housekeepina
The inspector toured the facility and noted overall very good plant conditions
including areas outside the station. Areas were generally neat and no leaking
equipment was noted. The licensee took action to clean and paint, as appropriate,
areas previously noted by the inspector to exhibit poor conditions (e.g., waste
collector / floor drain collector tank room, condensate backwash receiving tank
rooms).
R8.3 Verification of Updated Final Safety Analysis Commitmen_ts
a.
Scope (83750)
A recent discovery of a licensee operating their facility in a manner contrary to the
UFSAR description highlighted the need for a special, focused review that compares
plant practices, procedures and/or parameters to the UFSAR description. While
performing the inspections discussed in this report, the inspectors reviewed the
applicable portions of the UFSAR that related to the areas inspected.
b.
Observations and Findinas
The inspector reviewed turbine building ventilation systems associated with
personnel contamination by noble gas daughter products. The inspector discussed
the ventilation system design bases and any changes made to the ventilation
system relative to UFSAR descriptions.
c.
Conclusions
No inconsistencies were identified.
V. Manaaement Meetinas
X1
Exit Meeting Summary
An exit meeting was conducted on April 25,1997, at which the results of the inspection
were presented. PECO representatives acknowledged, and did not contest, the findings at
that time.
X2
Review of UFSAR Commitments
A recent discovery of a licensee operating their facility in a manner contrary to the Updated
Final Safety Analysis Report (UFSAR) description highlighted the need for a special focused
review that compares plant practices, procedures and/or parameters to the UFSAR
description. While performing the inspections discussed in this report, the inspector
reviewed the application portions of the UFSAR that related to the areas inspected. The
inspector verified that the UFSAR wording was consistent with the observed plant
practices, procedure and/or parameters.
._
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INSPECTION PROCEDURES USED
IP 37551: Onsite Engineering Observations
IP 40500: Effectiveness of Licensee Controls in identifying, Resolving,and Preventing
Problems
IP 61726: Surveillance Observations
IP 62707: Maintenance Observation
IP 64704: Fire Protection Program
IP 71707: Plant Operations
IP 71750: Plant Support Observations
IP 83750: Occupational Exposure
IP 92700: Onsite Follow of Written Reports of Nonroutine Events at Power Reactor
Facilities
IP 92901: Operations Followup
IP 92902: Followup - Engineer
IP 92903: Followup - Maintenance
IP 92904: Plant Support Followup
iP 93702: Prompt Onsite Response to Events at Operating Power Reactors
ITEMS OPENED, CLOSED, AND DISCUSSED
i
Ooened
IFl 50-278/97-04-01
Review Maintenance Rule Program Application to 13 kV
Breaker Switches
IFl 50-277; 50/278/97-04-02
Review Revisions to Fire Protection Training Lesson
Plans
Closed
VIO 50-277; 50-278/97-04-03
Violation of locked high radiation area key control
URI 50-277; 50-278/96-06-04
Updating of the UFSAR in accordance with 10 CFR 50.71 (e).
'
LER 2-97-001
Flow Biased Scram Setpoints
LER 2-97-002
Recirc Pump Motor PF issues
LER 3-97-002
Reactor Scram due to Natural Circulation
^
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LIST OF ACRONYMS USED
action request (AR)
action statement (AS)
administrative guideline (AG)
APRM gain adjust factor (AGAF)
as-low-as-reasonably-achievable (ALARA)
average power range monitors - neutron (APRMs)
control rod drives (CRDs)
control room deficiency list NRDL)
'
control room emergency ventilation (CREV)
core power and flow log (CPFL)
core spray (CS)
core thermal power (CTP)
design input document (DID)
diaphragm alternative response test (DART)
disintegrations per minute (DPM)
electro-hydraulic control (EHC)
i
eleventh refueling outage (2R11)
emergency core cooling system (ECCS)
emergency diesel generators (EDG)
'
emergency service water (ESW)
end-of-cycle (EOC)
engineering change request (ECR)
engineered safety feature (ESF)
equipment study list (ESL)
functional testing (FT)
general procedure (GP)
Generic Letter (GL)
health physics (HP)
high pressure coolant injection (HPCI)
high pressure service water (HPSW)
hydraulic control unit (HCU)
improved TS (ITS)
independent safety engineering group (ISEG)
inservice inspection (ISI)
inspector followup items (IFis)
instrument and control (l&C)
intermediate range monitor - neutron (IRM)
licensee event report (LER)
limited senior reactor operators (LSROs)
limiting conditions for operation (LCO)
load tap changer (LTC)
local leak rate test (LLRT)
loss of coolant accident (LOCA)
loss of off-site power (LOOP)
low pressure coolant injection (LPCI)
lubricating oil (LO)
modification (MOD)
motor generator (MG)
~
4
i
2
nuclear maintenance division (NMD)
nuclear review board (NRB)
offsite dose calculation manual (ODCM)
offsite power start-up source #2 (2SU)
offsite power start-up source #3 (3SU)
!
Peco Energy (PECO)
performance enhancement program (PEP)
plant equipment operator (PEO)
plant operations review committee (PORC)
post-maintenance testing (PMT)
i
primary containment (PC)
primary containment isolation system (PCIS)
primary containment isolation valve (PCIV)
protected area (PA)
4
quality assurance (QA)
radiation protection manager (RPM)
radiologically controlled area (RCA)
'
rated thermal power (RTP)
.
reactor core isolation cooling (RCIC)
!
reactor engineer (RE)
'
reactor feed pump (RFP)
reactor operator (RO)
.,
reactor protection system (RPS)
I
reliability centered maintenance (ROM)
,
'
safety evaluation report (SER)
safety related structures, system and components (SSC)
2
scram solenoid pilot valve (SSPV)
-
senior reactor operator (SRO)
shift update notice (SUN)
source range monitor (SRM)
i
spent fuel pool (SFP)
standby gas treatment (SGTS)
station blackout (SBO)
structure, system and component (SSC)
surveillance requirement (SR)
surveillance test (ST)
systerns approach to training (SAT)
technical requirements manual (TRM)
technical specification (TS)
- temporary plant alteration (TPA)
7
2
V-
3-
turbine control valve (TCV)
' turbine stop valve (TSV)
e
unresolved item (URI)
updated final safety analysis report (UFSAR)
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)
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