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                                                                                            I
U. S. NUCLEAR REGUL.ATORY COMMISSION
                          U. S. NUCLEAR REGUL.ATORY COMMISSION                             i
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                                                                                            J
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                                              REGION I
REGION I
    Docket / Report No. 50-277/97-04                                 License Nos. DPR-44
Docket / Report No.
                          50-278/97-04                                             DPR-56
50-277/97-04
    Licensee:           PECO Energy Company
License Nos. DPR-44
                        P. O. Box 195
50-278/97-04
                        Wayne, PA 19087-0195
DPR-56
                                                                                            l
Licensee:
    Facility Name:       Peach Bottom Atomic Power Statinn Units 2 and 3                   I
PECO Energy Company
    Dates:               May 4 - June 7,1997
P. O. Box 195
    Inspectors:         W. L. Schrnidt, Senior Resident inspector
Wayne, PA 19087-0195
                          M. J. Buckley, Resident inspector
l
                          B. D. Welling, Resident inspector
Facility Name:
                        J. W. Shea, NRR Project Manager
Peach Bottom Atomic Power Statinn Units 2 and 3
                          R. L. Fuhrmeister, Sr. Reactor Engineer
Dates:
                          R. L. Nimitz, Sr. Radiation Specialist
May 4 - June 7,1997
    Approved By:         P. D. Swetland, Acting Chief                                     l
Inspectors:
                          Reactor Projects Branch 4                                         i
W. L. Schrnidt, Senior Resident inspector
                          Division of Reactor Projects
M. J. Buckley, Resident inspector
                                  ,
B. D. Welling, Resident inspector
  9707310261 970724
J. W. Shea, NRR Project Manager
  PDR   ADOCK 05000277
R. L. Fuhrmeister, Sr. Reactor Engineer
  G                   PDR
R. L. Nimitz, Sr. Radiation Specialist
Approved By:
P. D. Swetland, Acting Chief
Reactor Projects Branch 4
i
Division of Reactor Projects
9707310261 970724
,
PDR
ADOCK 05000277
PDR
G


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                                            EXECUTIVE SUMMARY
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                                      Peach Bottom Atomic Power Station                                         i
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                            NRC Inspection Report 50-277/97-04, 50 278/97-04                                     i
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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             i
. - -
        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                     !
l
        areas including operations, maintenance, security, and emergency planning. The written                   i
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        record of the surveillance showed proper scope and good documentation of conclusions.
a
                                                                                                                  1
EXECUTIVE SUMMARY
        Plant Operations:                                                                                         1
Peach Bottom Atomic Power Station
                                                                                                                  1
i
                                                                                                                  l
NRC Inspection Report 50-277/97-04, 50 278/97-04
        Operators performed routine activities wellincluding controls over the plant during on-line
i
        control rod hydraulic control unit (HCU) maintenance and removal from service of the Unit
This integrated inspection report includes aspects of resident and region based inspection
        3 fifth stage feedwater heaters during Unit 3 coastdown operation. Operators also
of routine and reactive activities in: operations; surveillance and maintenance; engineering
        responded well when they identified, during clearance application, that a control rod that
i
        was not scheduled to be worked had been inserted, based on reactor engineering direction,
and technical support; and plant support areas.
        for HCU work. They also performed well when several abnormal conditions developed
Overall Assurance of Quality:
        during the HCU work, including: an individual Unit 2 control rod scrammed due to a
PECO Energy (PECO) operated both units safely over the period.
        clearance application error by nuclear material department personnel, several scram air
The PECO quality assurance department (QA) conducted surveillance in a broad range of
        header leaks and low pressure alarms, and a control rod that drifted from fully withdrawn
areas including operations, maintenance, security, and emergency planning. The written
        to fully inserted, due to scram inlet isolation valve damage and leakage.
record of the surveillance showed proper scope and good documentation of conclusions.
        Control room operators responded well to a faulty automatic high temperature isolation
1
        switch for the high pressure coolant injection (HPCI) system at Unit 2. However, during
Plant Operations:
        jumpering of the faulty instrument, an operator lifted the wrong lead on the terminal strip,
1
        causing a partial loss of logic power and inability of HPCI to automatically start. Operators
l
        responded to the loss of logic power and restored the system to operable status within
Operators performed routine activities wellincluding controls over the plant during on-line
        several minutes. While the operator made a mistake, the procedural guidance on the
control rod hydraulic control unit (HCU) maintenance and removal from service of the Unit
        installation of jumpers did not provide specific information on where to install jumpere snd
3 fifth stage feedwater heaters during Unit 3 coastdown operation. Operators also
        possible problems that could result if not installed properly.
responded well when they identified, during clearance application, that a control rod that
        Plant housekeeping was generally excellent, however PECO needed to take actions to
was not scheduled to be worked had been inserted, based on reactor engineering direction,
        address continued leaking of emergency diesel generator (EDG) lubricating and fuel oil, to
for HCU work. They also performed well when several abnormal conditions developed
        preclude an additional fire hazard.
during the HCU work, including: an individual Unit 2 control rod scrammed due to a
        The inspectors reviewed and closed three licensee event reports (LERs), finding that two
clearance application error by nuclear material department personnel, several scram air
        represented technical violations of the operating license core thermal power limit, which
header leaks and low pressure alarms, and a control rod that drifted from fully withdrawn
        the licensee identified, properly reported and corrected. These failures constituted
to fully inserted, due to scram inlet isolation valve damage and leakage.
        licensee-identified and corrected violations are being treated as Non-Cited Violations
Control room operators responded well to a faulty automatic high temperature isolation
        consistent with Section Vll.B.1 of the NRC Enforcement Policy.
switch for the high pressure coolant injection (HPCI) system at Unit 2.
                                                        ii
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.
ii


.
.
.
.
  Maintenance:
Maintenance:
  The inspectors found that PECO personnel conducted maintenance and surveillance
The inspectors found that PECO personnel conducted maintenance and surveillance
  activities acceptably during the period.                                                     !
activities acceptably during the period.
                                                                                              l
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  PECO planned and coordinated the HCU on-line maintenance generally well. Control room
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       '
staff exhibited very good control of post-maintenance testing. However, a maintenance
  technician caused the inadvertent scram of a single Unit 2 control rod due to an error while 1
'
  pulling fuses to establish electricalisolation. The inspectors noted that this was the first '
technician caused the inadvertent scram of a single Unit 2 control rod due to an error while
  time that PECO allowed maintenance technicians to establish their own isolation for work.
1
  Further, this event revealed weaknesses in double verification techniques and supervisory
pulling fuses to establish electricalisolation. The inspectors noted that this was the first
  oversight.
time that PECO allowed maintenance technicians to establish their own isolation for work.
  The inspectors concluded that, overall, the maintenance outage on the E-3 and E-1
Further, this event revealed weaknesses in double verification techniques and supervisory
  emergency diesel generators (EDGs) were effectively planned and implemented. PECO
oversight.
  corrected an out-of-specification crankshaft strain measurement on both machines,
The inspectors concluded that, overall, the maintenance outage on the E-3 and E-1
  implementing enhanced vendor information. Several minor problems, some involving
emergency diesel generators (EDGs) were effectively planned and implemented. PECO
  maintenance rework issues, caused delays in the post-maintenance testing and restoration
corrected an out-of-specification crankshaft strain measurement on both machines,
  to an operable status. PECO contingency planning allowed the problems to be resolved in
implementing enhanced vendor information. Several minor problems, some involving
  a timely manner.
maintenance rework issues, caused delays in the post-maintenance testing and restoration
  Operators adequately performed quarterly standby liquid control (SLC) system surveillance
to an operable status. PECO contingency planning allowed the problems to be resolved in
  testing in accordance with procedures. Operators identified some minor weaknesses and
a timely manner.
  opportunities for improvements in the recantly revised surveillance test procedure and in
Operators adequately performed quarterly standby liquid control (SLC) system surveillance
  the use of test equipment.
testing in accordance with procedures. Operators identified some minor weaknesses and
  The inspectors reviewed an issue that occurred during the Unit 21996 refueling outage
opportunities for improvements in the recantly revised surveillance test procedure and in
  where a maintenance work group pulled the wrong fuse during preparations for
the use of test equipment.
  containment isolation valve local leak rate testing (LLRT) work. This personnel error
The inspectors reviewed an issue that occurred during the Unit 21996 refueling outage
  resulted in'a loss of electrical power to the mechanical vacuum isolation valves, but no
where a maintenance work group pulled the wrong fuse during preparations for
  actual valve motion resulted since the valves were closed. Although this activity resulted
containment isolation valve local leak rate testing (LLRT) work. This personnel error
  in insignificant safety impact, it represents poor maintenance activity performance.
resulted in'a loss of electrical power to the mechanical vacuum isolation valves, but no
  Unfamiliarity with fuse removal and self-checking methodologies by the personnel involved,
actual valve motion resulted since the valves were closed. Although this activity resulted
  and an inadequate pre-job briefing contributed to the event.
in insignificant safety impact, it represents poor maintenance activity performance.
  In review of PECO's implementation of the improved technical specification (ITS), the
Unfamiliarity with fuse removal and self-checking methodologies by the personnel involved,
  independent safety engineering group (ISEG) and the NRC noticed difference between the
and an inadequate pre-job briefing contributed to the event.
  ITS and the old custom specification, in the definition of an instrument channel functional
In review of PECO's implementation of the improved technical specification (ITS), the
  test (CFT). This difference caused ISEG to question whether PECO was conducting
independent safety engineering group (ISEG) and the NRC noticed difference between the
  adequate testing. The NRC Nuclear Reactor Regulation (NRR) staff to reviewed the issue
ITS and the old custom specification, in the definition of an instrument channel functional
  and concluded that PECO satisfied the objectives of the ITS CFT requirements, by verifying
test (CFT). This difference caused ISEG to question whether PECO was conducting
  the function of one contact in all relays supplying a signal to engineered safety function
adequate testing. The NRC Nuclear Reactor Regulation (NRR) staff to reviewed the issue
  system logic. The inspectors also concluded that the ISEG review had led to an improved
and concluded that PECO satisfied the objectives of the ITS CFT requirements, by verifying
  understanding of the regulatory requirements in this area.
the function of one contact in all relays supplying a signal to engineered safety function
                                                iii
system logic. The inspectors also concluded that the ISEG review had led to an improved
understanding of the regulatory requirements in this area.
iii


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-                                                                                                      .
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                                                                                                        i
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  Enoineerino:
Enoineerino:
  Engineering department management and system mangers provided very good support to
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 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.                               1
the crankshaft strain issues discussed in sections M1.2 and M1.3 above.
  The reactor engineers performed well during on-line HCU maintenance, properly reviewing
1
  the needed initial conditions and the thermal limit effects prior to inserting control rods to
The reactor engineers performed well during on-line HCU maintenance, properly reviewing
  be worked, and conducting the post-work scram testing. In one instance, reactor
the needed initial conditions and the thermal limit effects prior to inserting control rods to
  engineering directed the operators to insert the wrong control rod. There was no adverse             l
be worked, and conducting the post-work scram testing. In one instance, reactor
  effect on thermal limits, and operators identified and corrected the mistake before work
engineering directed the operators to insert the wrong control rod. There was no adverse
  began.
effect on thermal limits, and operators identified and corrected the mistake before work
  Plant Sucoort:
began.
  PECO implemented an effective fire protection program, maintaining fire fighting equipment
Plant Sucoort:
  accessible and in good condition. One discrepancy was noted in the incorporation of a PEP             ,
PECO implemented an effective fire protection program, maintaining fire fighting equipment
  issue lesson learned into a lesson plan for fire watch training. Housekeeping in the plants           I
accessible and in good condition. One discrepancy was noted in the incorporation of a PEP
  was noted to be excellent. Evaluation of, and corrective actions for discrepancies
,
  identified by audits and self-assessments, were comprehensive and well focused.
issue lesson learned into a lesson plan for fire watch training. Housekeeping in the plants
  Overall effective radiological controls were implemented including planning and preparation
was noted to be excellent. Evaluation of, and corrective actions for discrepancies
  for the Unit 3 outage. The ALARA program was effectively implemented. The external
identified by audits and self-assessments, were comprehensive and well focused.
  and internal exposure controls program were effective. Weaknesses were identified in the
Overall effective radiological controls were implemented including planning and preparation
  evaluation and control of non-routine effluent / material release paths indicating a need for
for the Unit 3 outage. The ALARA program was effectively implemented. The external
  enhanced attention to detailin this area. A violation for lack of proper controls over high
and internal exposure controls program were effective. Weaknesses were identified in the
  radiation area keys was identified by the licensee and corrected. Although it does not
evaluation and control of non-routine effluent / material release paths indicating a need for
  appear that the keys were misused, this violation was cited because several keys were
enhanced attention to detailin this area. A violation for lack of proper controls over high
  uncontrolled for a number of years.
radiation area keys was identified by the licensee and corrected. Although it does not
                                                                                                        I
appear that the keys were misused, this violation was cited because several keys were
                                                                                                        1
uncontrolled for a number of years.
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iv
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j
                                                iv                                                    j
                                                                                                        i


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n.
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                                                                                                                                        \
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:                                             TABLE OF CONTENTS
:
        EX EC U TIV E S U M M ARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ii
TABLE OF CONTENTS
        SUMM ARY OF PLANT ACTIVITIES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
EX EC U TIV E S U M M ARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ii
        1.   Operations ................................................. 1
SUMM ARY OF PLANT ACTIVITIES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
            01     Conduct of Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1.
            04     Operator Knowledge and Performance . . . . . . . . . . . . . . . . . . . . . . . . . 1
Operations
                    04.1 High Pressure Coolant injection inoperable Due to Lifted Lead -
................................................. 1
                            Unit 2........................................... 1
01
            07     Quality Assurance in Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
Conduct of Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
            08     Miscellaneous Operations issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
04
                    08.1 (Closed) Licensee Event Reports 2-97-001, 2-97-002, and 3-
Operator Knowledge and Performance . . . . . . . . . . . . . . . . . . . . . . . . . 1
                            97-002.......................................... 3                                                         l
04.1 High Pressure Coolant injection inoperable Due to Lifted Lead -
                                                                                                                                        !
Unit 2...........................................
        II. Maintenance................................................                                                           4
1
            M1'     Conduct of Maintenance and Surveillance . . . . . . . . . . . . . . . . . . . . . . 4
07
                    M 1.1 Hydraulic Control Unit On-line Maintenance . . . . . . . . . . . . . . . . 4
Quality Assurance in Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
                    M1.2 E-3 Emergency Diesel Generator Maintenance                                   ..............             6  ;
08
                    M1.3 E-1 Emergency Diesel Generator Maintenance                                   ..............             8  I
Miscellaneous Operations issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3
                    M1.4 Surveillance Activities ...............................                                                   8
08.1 (Closed) Licensee Event Reports 2-97-001, 2-97-002, and 3-
            M4     Maintenance Staff Knowledge and Performance . . . . . . . . . . . . . . . . . . 9                                 j
97-002..........................................
                    M4.1 incorrect Fuse Removal During Local Leak Rate Testing ......                                             10
3
                                                                                                                                        '
II.
            M8      Miscellaneous Maintenance issues (92902) . . . . . . . . . . . . . . . . . . . . 10
Maintenance................................................
                    M8.1 Review of Instrument Channel Functional Test Practices . . . . . . 11
4
        Ill. E n g i n e e ri n g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .12
M1'
            E1     General Engineering Comments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12
Conduct of Maintenance and Surveillance . . . . . . . . . . . . . . . . . . . . . . 4
        IV. Plant Support   ..............................................                                                       13
M 1.1 Hydraulic Control Unit On-line Maintenance . . . . . . . . . . . . . . . . 4
            F2     Status of Fire Protection Facilities and Equipment . . . . . . . . . . . . . . . . 13
M1.2 E-3 Emergency Diesel Generator Maintenance
                    F2.1 Material Condition Inspection and Equipment Inventories .....                                           13
6
            F3     Fire Protection Procedures and Documentation . . . . . . . . . . . . . . . . . . 15
..............
                    F3.1   Fire Protection Procedure Reviews                     .....................                         15
M1.3 E-1 Emergency Diesel Generator Maintenance
            F5     Fire Protection Staff Training and Qualification . . . . . . . . . . . . . . . . . . 17
8
                    F5.1   Fire Brigade Training           ...............................                                     17
..............
                    F.5.2 (OpenIFl 50-277/278/97-04-01) Fire Watch Training . . . . . . . . 18
M1.4 Surveillance Activities
            F7     Quality Assurance in Fire Protection                 .........................                               19
8
                    F7.1   Fire Protection Program Audits . . . . . . . . . . . . . . . . . . . . . . . . 19
...............................
                    F7.2 Review of August 10,1994 .......................... 20
M4
            F8     Miscellaneous Fire Protection issues . . . . . . . . . . . . . . . . . . . . . . . . . 21
Maintenance Staff Knowledge and Performance . . . . . . . . . . . . . . . . . . 9
                    F8.1   Conformance to Updated Final Safety Analysis Report
j
l                           D e s cription . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21
M4.1 incorrect Fuse Removal During Local Leak Rate Testing
                    Radiological Protection and Chemistry (RP&C) Controls . . . . . . . . . . . . 22
10
      *
'
i-          R1
......
l                    R 1.1 Radiological Controls (Program Changes) . . . . . . . . . . . . . . . . . 22
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
13
..............................................
F2
Status of Fire Protection Facilities and Equipment . . . . . . . . . . . . . . . . 13
F2.1
Material Condition Inspection and Equipment Inventories
13
.....
F3
Fire Protection Procedures and Documentation . . . . . . . . . . . . . . . . . . 15
F3.1
Fire Protection Procedure Reviews
15
.....................
F5
Fire Protection Staff Training and Qualification . . . . . . . . . . . . . . . . . . 17
F5.1
Fire Brigade Training
17
...............................
F.5.2 (OpenIFl 50-277/278/97-04-01) Fire Watch Training . . . . . . . . 18
F7
Quality Assurance in Fire Protection
19
.........................
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
l
D e s cription . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21
i-
R1
Radiological Protection and Chemistry (RP&C) Controls . . . . . . . . . . . . 22
*
l
R 1.1 Radiological Controls (Program Changes) . . . . . . . . . . . . . . . . . 22
l
l
i
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            -- _ _ -                . .  .      ._ - - -            -- - -                              -          -        .-
,
  , - -.
                                                                                                                                      l
..
          Table of Contents
                                                                                                                                      !
                            R1.2 ALARA Program and Unit 3 Refueling Outage                                                          l
                                    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 ...                                  25
                            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) .............................................                                                32
                            R7.1 Radiological Event Reports . . . . . . . . . . . . . . . . . . . . . . . . . . . 32
                      R8    Miscellaneous RP&C Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33              i
                            R8.1  (Closed) Unresolved item 96-06-04: Review of Radioactive
i                                  Material Storage Locations Versus Updated Final Safety
                                    Analysis (UFSAR) Descriptions . . . . . . . . . . . . . . . . . . . . . . . . 33
                            R8.2 Housekeeping ....................................                                                35
;                          R8.3 Verification of Updated Final Safety Analysis Commitments ...                                    35
          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
-.
          INSPECTION PROCEDURES USED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 36
-- _ _ -
          ITEMS OPENED, CLOSED, AND DISCUSSED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 36
. .
          LIST O F ACRO N YM S U SED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 7
.
                                                              vi
._ - - -
-- - -
-
-
.-
..
Table of Contents
R1.2 ALARA Program and Unit 3 Refueling Outage
l
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
25
...
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)
32
.............................................
R7.1 Radiological Event Reports . . . . . . . . . . . . . . . . . . . . . . . . . . . 32
R8
Miscellaneous RP&C Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33
R8.1
(Closed) Unresolved item 96-06-04: Review of Radioactive
i
Material Storage Locations Versus Updated Final Safety
Analysis (UFSAR) Descriptions . . . . . . . . . . . . . . . . . . . . . . . . 33
R8.2 Housekeeping
35
....................................
R8.3 Verification of Updated Final Safety Analysis Commitments
35
;
...
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
-
INSPECTION PROCEDURES USED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 36
ITEMS OPENED, CLOSED, AND DISCUSSED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 36
LIST O F ACRO N YM S U SED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 7
vi


                  __.   _       .       _ ._           _   . _ -     . _ .   __         _   _         _     __
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  ,
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_ ._
_
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. _ .
__
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_
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1*
1*
    SUMMARY OF PLANT ACTIVITIES
SUMMARY OF PLANT ACTIVITIES
    PECO operated both units safely over the period.
PECO operated both units safely over the period.
    Unit 2 remained at essentially 100% power, until May 18, when operators reduced power
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
to approximately 70% to allow control rod (CR) hydraulic control unit (HCU) on-line
l   maintenance. The nuclear maintenance division (NMD) completed the maintenance and
l
l  operators began restoring the unit to 100% power on May 23.
maintenance. The nuclear maintenance division (NMD) completed the maintenance and
l
l
l   Unit 3 entered the period at 100 % power, operators reduced power to allow on-line HCU
operators began restoring the unit to 100% power on May 23.
l
l
Unit 3 entered the period at 100 % power, operators reduced power to allow on-line HCU
'
'
    maintenance on May 4 and began returning the unit to 100% power on May 11, following
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
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
at approximately 98 % power, af ter removal of the fifth stage feedwater heaters on June
    1.
1.
                                                1. Operations
1. Operations
                                                                                                                    l
01
    01      Conduct of Operations'
Conduct of Operations'
            Operators performed routine activities wellincluding control of the plant during on-
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
line HCU maintenance and removal of the Unit 3 fifth stage feedwater heaters from
            service. Operators also responded well when they identified, during clearance
service. Operators also responded well when they identified, during clearance
            application, that a control rod that was not scheduled to be worked had been
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                               l
inserted, based on reactor engineering direction, for HCU work. They also
            performed well when several abnormal conditions developed during the HCU work,
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
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
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
control rod that drifted from fully withdrawn to fully inserted, due to scram insert
            valve damage and leakage.
valve damage and leakage.
            Plant housekeeping was generally excellent, however PECO needed to take actions
Plant housekeeping was generally excellent, however PECO needed to take actions
            to preclude an additional fire hazard from continued leaking of emergency diesel
to preclude an additional fire hazard from continued leaking of emergency diesel
            generator (EDG) lubricating and fuel oil.
generator (EDG) lubricating and fuel oil.
    04     Operator Knowledge and Performance
04
    04.1 Hiah Pressure Coolant Iniection Inocerable Due to Lifted Lead - Unit 2
Operator Knowledge and Performance
      a.   Scope
04.1 Hiah Pressure Coolant Iniection Inocerable Due to Lifted Lead - Unit 2
            The inspectors reviewed the circumstances leading to a short period of inoperability
a.
            of the Unit 2 high pressure coolant injection (HPCI) system on June 1, while
Scope
            operators tried to install a jumper to place a failed high area temperature isolation
The inspectors reviewed the circumstances leading to a short period of inoperability
            relay in the tripped condition.
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
    InaiJll.'t d[o*rf.'".". "of".".Sa"ta .llan''In"o'd
relay in the tripped condition.
                                              u      i".'?"IU"" """ '"* * "'"'*"'"' "'**** *""** * * ""'' "'""'
InaiJll.'t d[o*rf.'".". "of".".Sa"ta .llan''In"o'd i".'?"IU"" """ '"* * "'"'*"'"' "'**** *""** * * ""'' "'""'
"
'*
u
.


    - _ .       .           -                       .   -. .-         - -     . .-       -   .   _.
,
  ,
- _ .
                                                                                                        .,
.
  .
-
                                                        2
.
          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       )
. .-
l             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
2
              there was a test jack installed on the terminal strip external connection, with no test
b.
              connections on the internal side of the strip. The operator looked at the other
Observations and Findinos
              connection and decided to install the jumper on the internal side of the terminal         !
During operator rounds it was noted that the instrument (TE 4944D) did not pass
              strip. After he lifted the lead on the internal side of the strip, he noted that there     I
the required daily channel check. As such, the operators properly decided to insert
              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
l
              continuity through the terminal, and resulted in a loss of power to this channel of
the trip from this instrument in accordance with technical specifications (TSs) and
              HPCIlogic, making the system inoperable for automatic actuation. The operator
general procedure (GP) 25.
                                                                                                        "
GP 25 directed that the operator install a jumper between two terminals'that would
              immediately identified the mistake and re-landed the lifted lead within two minutes
bypass the temperature switch contact and cause the high temperature isolation
              of lifting it.
relay to energize, placing the channel in the tripped condition per TSs.
              In review of GP 25 and the operations manual (OM) Section 7.7 that covers the
When the operator went to install the jumper he saw that on one of the terminals
              installation of jumpers by operators, the inspector found that:
there was a test jack installed on the terminal strip external connection, with no test
              *       GP 25 did not reference OM 7.7
connections on the internal side of the strip. The operator looked at the other
              *       Neither procedure was specific as to which side of a terminal strip was the       I
connection and decided to install the jumper on the internal side of the terminal
                      external or the internal wiring point.
strip. After he lifted the lead on the internal side of the strip, he noted that there
              *       The procedures did not address the possibility of installation of jumpers
was a flat metal "C" jumper installed between the terminal and the next terminal
                      causing a loss of power to other components in the same circuit, since the
below. Lifting the lead by removing the screw had caused a loss of power
                      terminal screws need to be removed to install a round lugged jumper.
continuity through the terminal, and resulted in a loss of power to this channel of
              *       There was no discussion of possible flat "C" jumpers that may be hard to
HPCIlogic, making the system inoperable for automatic actuation. The operator
                      see.
"
          C. Conclusions
immediately identified the mistake and re-landed the lifted lead within two minutes
                                                                                                        i
of lifting it.
              The operators responded well to the event and limited the duration of HPCI
In review of GP 25 and the operations manual (OM) Section 7.7 that covers the
              inoperability. The operator made a mistake by lifting a lead on the internal
installation of jumpers by operators, the inspector found that:
              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       ,
GP 25 did not reference OM 7.7
              inspector will review the PECO corrective actions upon receipt of the licensee event       I
*
              report.
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
i
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
,
inspector will review the PECO corrective actions upon receipt of the licensee event
report.
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Line 356: Line 497:
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                                                  3
3
    07-   Quality Assurance in Operations
07-
    a.     Scoce
Quality Assurance in Operations
          The inspectors reviewed the quality assurance department's surveillance completed
a.
          over the last several months.
Scoce
      b.   Conclusions
The inspectors reviewed the quality assurance department's surveillance completed
          PECO QA conducted surveillance in a broad range of areas including operations,
over the last several months.
          maintenance, security, and emergency planning. The written record of the
b.
          surveillance showed proper scope and good documentation of conclusions.
Conclusions
                                                                                                l
PECO QA conducted surveillance in a broad range of areas including operations,
    08    M!scellaneous Operations issues:
maintenance, security, and emergency planning. The written record of the
    08.1 (Closed) Licensee Event Reports 2-97-001. 2-97-002. and 3-97-002 ~
surveillance showed proper scope and good documentation of conclusions.
                                                                                                l
08
      a.   Scooe
M!scellaneous Operations issues:
          The inspector reviewed the issues documented in several licensee event reports
08.1 (Closed) Licensee Event Reports 2-97-001. 2-97-002. and 3-97-002 ~
          (LERs).
l
      b.   Observations and Findinas
a.
          LER 2-97-001: Non-Conservative single loop average power range monitor (APRM)
Scooe
          flow biased scram setpoints.
The inspector reviewed the issues documented in several licensee event reports
          PECO reactor engineering identified this issue as they reviewed previous core power
(LERs).
  '
b.
          and flow data from single loop operation in the 1992 time period. The error
Observations and Findinas
          involved was small(less than 1%) and was due to the method used to establish the
LER 2-97-001: Non-Conservative single loop average power range monitor (APRM)
          flow biased APRM setpoints required by TS during single loop operations. PECO
flow biased scram setpoints.
          took appropriate actions to correct and report this issue.
PECO reactor engineering identified this issue as they reviewed previous core power
          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.
and flow data from single loop operation in the 1992 time period. The error
          PECO determined that due to inaccuracies in the instrumentation used to measure
involved was small(less than 1%) and was due to the method used to establish the
          the recirculation pump electrical energy core thermal power may have exceeded, by
flow biased APRM setpoints required by TS during single loop operations. PECO
          1.5 megawatts, the Units 1 and 2 license limit of 3458 megawatts. PECO found
took appropriate actions to correct and report this issue.
          that the watt transducers for the recirculation pumps did not account for the power
LER 2-97-002: Reactor power slightly greater than licensed thermal power - due to
          factor of the electrical supply. Accounting for the power factor would lessen the
inaccurate accounting of recirculation pump power in the thermal heat balance.
          amount of energy actually being used to drive the pump and thus lessen the amount
PECO determined that due to inaccuracies in the instrumentation used to measure
          of actual energy being added to the reactor coolant by the recirculation pump.
the recirculation pump electrical energy core thermal power may have exceeded, by
          Since the calculation subtracts the recirculation pump energy from the core thermal
1.5 megawatts, the Units 1 and 2 license limit of 3458 megawatts. PECO found
          power (i.e., allowing reactor power to be increased by the amount being added by
that the watt transducers for the recirculation pumps did not account for the power
          the recirculation pumps), not accounting for the power factor allowed actual reactor
factor of the electrical supply. Accounting for the power factor would lessen the
          power to be above the license limit.
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.


                      ,                         .       .=.         _ - -       .           .
,
    ,
,
.
.=.
_ - -
.
.
.. .
.. .
                                                    4
4
            LER 3-97-002: Manual Reactor Scram due to Natural Circulation Operation
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
,
,
            On March 9,1997, while in single loop operation to investigate a low lube oil
condition indicated on the idle recirculation pump, Unit 3 experienced a loss cf the
            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
            operating recirculation pump due to a faulty interlock between the main generator
and the recirculation pump power supply. The resultant manual reactor scram and
            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
            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
            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
.
.
            for a turbine trip, an auxiliary breaker position contact failed to change position
indicating that the main generator was still supplying the recirculation pump
            indicating that the main generator was still supplying the recirculation pump
switchgear. When the turbine was tripped, the logic for 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
            thought that the recirculation pump was still powered from the main generator and
tripped the recirculation pump motor generator supply breaker open in anticipation
            tripped the recirculation pump motor generator supply breaker open in anticipation
of the main generator output breaker opening. With the resultant natural circulation
            of the main generator output breaker opening. With the resultant natural circulation
reactor condition, procedures called for a manual reactor scram. PECO repaired the
            reactor condition, procedures called for a manual reactor scram. PECO repaired the
breaker auxiliary contact and after other plant repairs restarted the unit. PECO
,
,
            breaker auxiliary contact and after other plant repairs restarted the unit. PECO
committed to review the other 13 KV breakers installed at the site to determine if
            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
            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
            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-
            KV breaker switches will be followed during a subsequent inspection. (IFl 50-
278/97-04-01)
            278/97-04-01)
c.
      c.   Conclusions
Conclusions
            PECO adequately documented the conditions described in the LERs reviewed above.
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
With respect to LERs 2-97-001 and 2-97-002, the inspectors found that these
.
.
2
events represented technical violations of the operating license that the licensee
            events represented technical violations of the operating license that the licensee
2
            identified, properly reported and corrected. These failures constituted licensee
identified, properly reported and corrected. These failures constituted licensee
            identified and corrected violations and are being treated as Non-Cited Violations
identified and corrected violations and are being treated as Non-Cited Violations
            consistent with Section Vll.B.1 of the NRC Enforcement Policy.
consistent with Section Vll.B.1 of the NRC Enforcement Policy.
                                    II. Maintenance and Surveillance
II. Maintenance and Surveillance
      M1     Conduct of Maintenance and Surveillance
M1
Conduct of Maintenance and Surveillance
i
i
            The inspectors found that PECO personnel conducted activities acceptably during
The inspectors found that PECO personnel conducted activities acceptably during
the period. There were several performance and equipment issues that developed
,
,
            the period. There were several performance and equipment issues that developed
;
;
            during HCU and EDG maintenance, as discussed below.
during HCU and EDG maintenance, as discussed below.
1
1
M1.1 Hydraulic Control Unit On-line Maintenance
'
'
      M1.1 Hydraulic Control Unit On-line Maintenance
a.
      a.    Scone (62707)
Scone (62707)
            During the period PECO conducted HCU on-line maintenance at both units. These
During the period PECO conducted HCU on-line maintenance at both units. These
            activities included scram pilot valve (SSPV) replacement and general HCU
activities included scram pilot valve (SSPV) replacement and general HCU
            preventive maintenance. The inspector reviewed
preventive maintenance. The inspector reviewed
  i
i
  ;         e       The Unit 3 activities conducted between May 4 and May 11,1997.
;
4
e
The Unit 3 activities conducted between May 4 and May 11,1997.
4


  .
.
  .
.
                                                5
5
      *        The Unit 2 activities performed during the week of May 19,1997.
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
b.
      involvement was generally apparent, and procedures and work controls were
Observations and Findinas
      adequately used. The inspector noted that the work was generally well-planned and     l
Overall, PECO NMD personnel conducted the HCU work well. Supervisor
      controlled. Control room staff displayed good command and control, and conducted
involvement was generally apparent, and procedures and work controls were
      effective briefings on several occasions to ensure that the numerous control rod
adequately used. The inspector noted that the work was generally well-planned and
      movements to support the maintenance and post-maintenance testing, were
controlled. Control room staff displayed good command and control, and conducted
      performed without error. Good coordination and communications were evident
effective briefings on several occasions to ensure that the numerous control rod
      between maintenance technicians, operators, and reactor engineers.
movements to support the maintenance and post-maintenance testing, were
      Maintenance technician performance was generally good. However, during the
performed without error. Good coordination and communications were evident
      activities several issued developed:
between maintenance technicians, operators, and reactor engineers.
      *       Inadvertent scram of one control rod - Unit 2
Maintenance technician performance was generally good. However, during the
      On May 22,1997, technicians pulled fuses for the wrong control rod, resulting in
activities several issued developed:
      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-
Inadvertent scram of one control rod - Unit 2
      normal procedure and determined that no thermallimits were exceeded as a result
On May 22,1997, technicians pulled fuses for the wrong control rod, resulting in
      of the event. The rod was returned to its original position.
an inadvertent single control rod scram. The technicians were directed by the shift
      PECO determined that the technicians had pulled fuses in the wrong fuse panel
supervisor to re-install the fuses. Operators correctly entered the applicable off-
      during the application of a clearance. Specifically, they were to pull fuses for
normal procedure and determined that no thermallimits were exceeded as a result
      control rod 26-11, which corresponded to the 14th fuse row down in panel
of the event. The rod was returned to its original position.
      2HC068. Instead, they pulled the fuses for control rod 18-59, which was the 14th
PECO determined that the technicians had pulled fuses in the wrong fuse panel
      row in panel 2CC068. Although the task involved double verification (one
during the application of a clearance. Specifically, they were to pull fuses for
      technician observing another), the second technician's verification focused on the
control rod 26-11, which corresponded to the 14th fuse row down in panel
      14th row and missed the check of the correct panel.
2HC068. Instead, they pulled the fuses for control rod 18-59, which was the 14th
      PECO found that this event revealed weaknesses in double verification techniques,
row in panel 2CC068. Although the task involved double verification (one
      attention to detail, and supervisory oversight. PECO's immediate corrective actions
technician observing another), the second technician's verification focused on the
      were to relieve maintenance personnel from performing HCU clearance work.
14th row and missed the check of the correct panel.
      Operators have applied all subsequent HCU clearances pending formal assessment
PECO found that this event revealed weaknesses in double verification techniques,
      under the performance enhancement process.
attention to detail, and supervisory oversight. PECO's immediate corrective actions
      The inspector's review of this issue determined that poor labeling inside of the fuse
were to relieve maintenance personnel from performing HCU clearance work.
      cabinets was also a contributor in this event. Inside the panel was a single label
Operators have applied all subsequent HCU clearances pending formal assessment
      that consisted of the control rod numbers cross-referenced to fuse row numbers
under the performance enhancement process.
      written with a grease pencil. Thus, the task of counting the row numbers added
The inspector's review of this issue determined that poor labeling inside of the fuse
      difficulty to the job, and apparently allowed the technicians to lose focus on other
cabinets was also a contributor in this event. Inside the panel was a single label
      verification steps. Furthermore, the inspector noted that the event occurred late in
that consisted of the control rod numbers cross-referenced to fuse row numbers
      the workers' shift, when errors can be more likely to occur.
written with a grease pencil. Thus, the task of counting the row numbers added
      *       Scram Air Header Leaks
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.
*
Scram Air Header Leaks
l
l
,
,


  ,- .   - - -           -.     -.     ..   -     - - . . ---.. ..--             .- - - - -         - - -
,-
  ..
.
                                                              6
- - -
i              There were several minor issues dealing with scram air header leaks during work on
-.
l               the SSPVs. Specifically, while PECO worked on these valves at Unit 3, the scram
-.
j               air header supply tubing upstream of the closed manual isolation valve was
..
l               unsupported. During the maintenance, some of the upstream compression fittings
-
(               came loose. This leak caused low scram air header pressure problems during the
- - . . ---.. ..--
i               work. In one case, the operators quickly placed the backup air supply into service
.- - - - -
I               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
..
i               stop. The deformed isolation valve seat allowed air leakage, which resulted in a
6
                fully withdrawn control rod drifting to the fully inserted position. PECO took
There were several minor issues dealing with scram air header leaks during work on
                adequate actions to preclude the air leaks from the scram air tubing by designing a
i
l               clamp that provided extra support to this section of tubing during the maintenance
l
                activity, at Unit 2.
the SSPVs. Specifically, while PECO worked on these valves at Unit 3, the scram
                e       insertion of a Wrong Control Rod                                                   -
j
              -The reactor engineers had generally good control over the HCU activities and
air header supply tubing upstream of the closed manual isolation valve was
                coordinated with the operators and NMD crew to properly position the control rods
l
                to be worked. The required post-maintenance testing, which included control rod
unsupported. During the maintenance, some of the upstream compression fittings
                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
came loose. This leak caused low scram air header pressure problems during the
                was in the groups of rods to be worked, was not in the sequence for work at that
i
                time. In this event, the reactor engineers had conducted the proper pre-insertion
work. In one case, the operators quickly placed the backup air supply into service
                core thermal limits analysis, which predicted no negative result from the insertion.
I
                After the rod was inserted, the operators began to apply the maintenance clearance
to prevent a low pressure condition. The pressure surge caused one on the control
                and noted that the wrong rod had been inserted. The operators and reactor
rod scram inlet isolation valves to deform because of an improperly set over travel
                engineer then completed a review of the issue and returned the control rod to the .
i
                fully withdrawn position,
stop. The deformed isolation valve seat allowed air leakage, which resulted in a
        c.     Conclusions
fully withdrawn control rod drifting to the fully inserted position. PECO took
                The inspector concluded that the planning and coordination associated with the Unit
adequate actions to preclude the air leaks from the scram air tubing by designing a
                2 HCU on-line maintenance was generally good. However, leakage from
l
                unsupported, disconnected air lines caused some operational problems. Also, a
clamp that provided extra support to this section of tubing during the maintenance
                work sequencing error by the reactor engineer resulted in inserting the wrong
activity, at Unit 2.
                control rod. Control room staff exhibited very good control of post-maintenance
e
                testing. Although maintenance technician performance was generally good, an error
insertion of a Wrong Control Rod
                resulted in pulling the fuses for an incorrect control rod. This event revealed
-
                weaknesses in double verification techniques and supervisory oversight.
-The reactor engineers had generally good control over the HCU activities and
      M1.2 E-3 Emeraency Diesel Generator Maintenance
coordinated with the operators and NMD crew to properly position the control rods
        a.     Scope (62707)
to be worked. The required post-maintenance testing, which included control rod
:              The inspectors reviewed the planned maintenance and post-maintenance testing
movement and scram testing was conducted very well. However, in one instance
i               performed on the E-3 emergency diesel generator (EDG) during the period May 12-
the reactor engineer directed that the operators insert a control rod that, while it
;               23,1997.
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
:
i
performed on the E-3 emergency diesel generator (EDG) during the period May 12-
;
23,1997.
i
i
f
f
I
I
i
i
                                                                                                        _
_
_


                                                    .. -             - .                    .
,
,                                                                            _-.         .
.. -
  .
- .
                                              7
_-.
    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
7
      approved procedures and technicians were knowledgeable of their work
b.
      assignments.
Observations and Findinas
                                                                                              1
The inspectors observed several planned preventive and corrective maintenance
      One of the significant activities involved the discovery of an out-of-specification
actions completed on the E-3 EDG. Overall, the work was effectively planned and
      crankshaft strain measurement. During this outage, PECO began using a new
implemented. The inspectors found that maintenance was performed according to
      evaluation criteria for crankshaft strain measurements. The out-of-specification
approved procedures and technicians were knowledgeable of their work
      measurement indicated that the generator could be set too high for alignment with     j
assignments.
      the diesel engine. Measurements on the diesel engine #14 bearing provided
1
      additional data that indicated a small alignment problem. After evaluation and         ;
One of the significant activities involved the discovery of an out-of-specification
      consultation with the vendor, PECO determined that the generator was set too high     l
crankshaft strain measurement. During this outage, PECO began using a new
      for the engine. PECO lowered the generator by removing shims on the bedplate.         I
evaluation criteria for crankshaft strain measurements. The out-of-specification
      Technicians also replaced the #14 diesel engine bearing because of out-of-
measurement indicated that the generator could be set too high for alignment with
      specification measurements, although no sign of bearing damage was evident.
j
      These actions returned the crankshaft strain measurement to within specifications.
the diesel engine. Measurements on the diesel engine #14 bearing provided
      PECO confirmed that the small misalignment had not affected EDG operability.
additional data that indicated a small alignment problem. After evaluation and
      Because of the additional maintenance associated with the bearing replacement, an
consultation with the vendor, PECO determined that the generator was set too high
      extended post-maintenance test of the EDG was required.
l
      A number of minor problems caused delays in the post-maintenance testing. First,
for the engine. PECO lowered the generator by removing shims on the bedplate.
      on May 20, operators observed a high crankcase vacuum reading and shut down
I
      the EDG. Technicians found that the exhaust eductor restricting orifice area had
Technicians also replaced the #14 diesel engine bearing because of out-of-
      opened up. Several Icaking fuel injectors were also identified, including some that
specification measurements, although no sign of bearing damage was evident.
      had been replaced during the maintenance period. PECO noted that there was some
These actions returned the crankshaft strain measurement to within specifications.
      confusion over the leakage acceptance critorion, and was pursuing the cause of the
PECO confirmed that the small misalignment had not affected EDG operability.
      leaking injectors with the vendor. Also, an intermittent ground caused some delays
Because of the additional maintenance associated with the bearing replacement, an
      and was traced to a fuel rack reset mechanism limit switch. Finally, the EDG lube
extended post-maintenance test of the EDG was required.
      oil standby circulating pump, which helps keep the lube oil warm, tripped repeatedly
A number of minor problems caused delays in the post-maintenance testing. First,
      between EDG test runs. PECO considered this to be a repeat maintenance / rework
on May 20, operators observed a high crankcase vacuum reading and shut down
      problem because the pump was worked and reworked during the maintenance
the EDG. Technicians found that the exhaust eductor restricting orifice area had
      period. PECO was evaluating the cause of the repeat maintenance problem at the
opened up. Several Icaking fuel injectors were also identified, including some that
      end of this inspection period. Despite the problems during the post-maintenance
had been replaced during the maintenance period. PECO noted that there was some
      testing, PECO contingency planning was adequate to allow the problems to be
confusion over the leakage acceptance critorion, and was pursuing the cause of the
      resolved in a generally timely manner. The problems were entered into the
leaking injectors with the vendor. Also, an intermittent ground caused some delays
      corrective action process for tracking and causal analysis,
and was traced to a fuel rack reset mechanism limit switch. Finally, the EDG lube
    c. Conclusions
oil standby circulating pump, which helps keep the lube oil warm, tripped repeatedly
      The inspectors concluded that, overall, the maintenance outage on the E-3 EDG was
between EDG test runs. PECO considered this to be a repeat maintenance / rework
      effectively planned and implemented. PECO corrected an out-of-specification
problem because the pump was worked and reworked during the maintenance
      crankshaft strain measurement by lowering the generator. Several minor problems,
period. PECO was evaluating the cause of the repeat maintenance problem at the
      some of which were maintenance rework issues, caused delays in the post-
end of this inspection period. Despite the problems during the post-maintenance
      maintenance testing and restoration to an operable status. PECO contingency
testing, PECO contingency planning was adequate to allow the problems to be
      planning allowed the problems to be resolved in a timely manner.
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.


!*
!*
  .
.
l
l
                                                    8
8
    M1.3 E-1 Emeraency Diesel Generator Maintenance
M1.3 E-1 Emeraency Diesel Generator Maintenance
    a.   Scooe (62707)
a.
          A review by the inspectors, of the planned on-line maintenance and testing, for the
Scooe (62707)
          E-1 emergency diesel generator (EDG) overhaul started on June 1 and continued
A review by the inspectors, of the planned on-line maintenance and testing, for the
                                                                                                i
E-1 emergency diesel generator (EDG) overhaul started on June 1 and continued
          through the end of the inspection period to verify that maintenance activities       l
i
          performed provided for assurance of reliable and safe operation.
through the end of the inspection period to verify that maintenance activities
    b.   Observations and Findinas
performed provided for assurance of reliable and safe operation.
                                                                                                :
b.
          The inspectors observed several effectively planned and executed E-1 EDG             1
Observations and Findinas
          preventive and corrective maintenance actions. Maintenance personnel performed
The inspectors observed several effectively planned and executed E-1 EDG
          work according to approved procedures, and technicians were aware and
1
          knowledgeable of their work assignments. Maintenance supervision coordinated
preventive and corrective maintenance actions. Maintenance personnel performed
          well with system managers.
work according to approved procedures, and technicians were aware and
          The new crankshaft strain measurement criterion resulted in significant work
knowledgeable of their work assignments. Maintenance supervision coordinated
          activity for PECO during the overhaul. Another generator move would be required
well with system managers.
          during the overhaul. After verification of the out-of-specification crankshaft stain
The new crankshaft strain measurement criterion resulted in significant work
          measurement and supporting measurement from the #14 bearing, maintenance
activity for PECO during the overhaul. Another generator move would be required
          moved the generator in the downward direction by removing shims. PECO lowered
during the overhaul. After verification of the out-of-specification crankshaft stain
          the generator in the same method used during the successful overhaul of the E-3
measurement and supporting measurement from the #14 bearing, maintenance
          EDG. Also a lateral generator move to bring the crankshaft stain and bearing
moved the generator in the downward direction by removing shims. PECO lowered
          measurements within specification was required after consultation with the vendor.
the generator in the same method used during the successful overhaul of the E-3
          The licensee chose to calculate the required move, but to move the generator
EDG. Also a lateral generator move to bring the crankshaft stain and bearing
          incrementally, thereby minimizing the flexing on the generator and diesel shafts.
measurements within specification was required after consultation with the vendor.
          During post-maintenance testing to return the E1 EDG to service, PECO identified a
The licensee chose to calculate the required move, but to move the generator
          field flash circuit relay (K1) reset problem that needed to be addressed before
incrementally, thereby minimizing the flexing on the generator and diesel shafts.
          operability of the El could be assured. Specifically, the K1 relay wouldn't reset
During post-maintenance testing to return the E1 EDG to service, PECO identified a
          following an EDG run. K1 must operate to reset the field flash after an engine
field flash circuit relay (K1) reset problem that needed to be addressed before
          shutdown to ensure proper operation of the emergency generator on a subsequent
operability of the El could be assured. Specifically, the K1 relay wouldn't reset
          automatic start. PECO appropriately concluded that this could affect EDG
following an EDG run. K1 must operate to reset the field flash after an engine
          operability and proceeded with a relay replacement. A satisfactory reset verification
shutdown to ensure proper operation of the emergency generator on a subsequent
          of the K1 relay on the other EDGs, performed by PECO, addressed the possibility of
automatic start. PECO appropriately concluded that this could affect EDG
          a generic failure. This unexpected relay replacement caused a delay of the return of
operability and proceeded with a relay replacement. A satisfactory reset verification
          the E1 EDG beyond the initial schedule.
of the K1 relay on the other EDGs, performed by PECO, addressed the possibility of
    c.   Conclusions
a generic failure. This unexpected relay replacement caused a delay of the return of
          The inspectors concluded that PECO had effectively planned and implemented the
the E1 EDG beyond the initial schedule.
          maintenance outage activities on the E-1 EDG. PECO competently corrected the
c.
          crankshaft strain measurement and field flash relay problems.
Conclusions
!   M 1.4 Surveillance Activities
The inspectors concluded that PECO had effectively planned and implemented the
          e       Standby Liquid Control System Surveillance Test - Unit 2
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
l
l


  , -.       .     .       .       .     .-       . - _ -       -     -           . - -   - . . .-
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                                                    9
9
        a. Scoce (61726)
a.
Scoce (61726)
4
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.                   )
The inspector observed the performance of surveillance test ST-O-011301-2,
        b. Observations and Findinos
?
" Standby Liquid Control Pump Functional Test for IST," on May 27,1997.
)
b.
Observations and Findinos
4
4
'
'
The inspector found that the Unit 2 quarterly standby liquid control (SBLC) system
'
'
          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
          surveillance, which verifies the operability and performance of the SBLC pumps and
check valves, was conducted adequately according to the surveillance test
.          check valves, was conducted adequately according to the surveillance test
.
,          procedure. The inspector observed operators performing preparations, valve
procedure. The inspector observed operators performing preparations, valve
,
i
i
          positioning, and other actions specified by the test procedure. Operators
positioning, and other actions specified by the test procedure. Operators
          appropriately held a pre-job briefing to ensure that test participants were aware of
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
the actions necessary to restore the system should it be needed to perform its
'
'
i         safety function.                                                                           !
i
safety function.
!
"
"
          The operators identified a number of minor weaknesses in the recently revised (April
The operators identified a number of minor weaknesses in the recently revised (April
            1997) surveillance test procedure. For example, some revised steps requiring
1997) surveillance test procedure. For example, some revised steps requiring
          operators to take pump oil samples caused considerable delays in the conduct of
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
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
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
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
resulted in considerable difficulty in drawing the oil sample, in addition, operators
          identified confusing directions as to whether to mark as "not applicable" procedural
identified confusing directions as to whether to mark as "not applicable" procedural
          sub steps or complete the steps of the procedure.
sub steps or complete the steps of the procedure.
          The operators also recognized some opportunities for improvement in their
The operators also recognized some opportunities for improvement in their
          familiarity with the test procedures and equipment. The inspector observed that
familiarity with the test procedures and equipment. The inspector observed that
          when problem areas arose, the operators appropriately consulted with shift
when problem areas arose, the operators appropriately consulted with shift
          supervision for clarification and guidance.
supervision for clarification and guidance.
          The inspector noted that a review of the SBLC pump standby oillevel band or
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.
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
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.
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.
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.
This discrepancy was brought to the attention of the operators.
        c. Conclusions
c.
          The inspector concluded that the quarterly SBLC system surveillance test was
Conclusions
          performed adequately in accordance with test procedures. Operators identified
The inspector concluded that the quarterly SBLC system surveillance test was
          some minor weaknesses in the recently revised surveillance test procedure and
performed adequately in accordance with test procedures. Operators identified
          identified some opportunities for improvement in their femiliarity with the test
some minor weaknesses in the recently revised surveillance test procedure and
          procedure and equipment. The inspector observed a weakness associated with the
identified some opportunities for improvement in their femiliarity with the test
          methodology or procedural guidance for taking pump oil samples.
procedure and equipment. The inspector observed a weakness associated with the
      M4 Maintenance Staff Knowledge and Performance
methodology or procedural guidance for taking pump oil samples.
M4
Maintenance Staff Knowledge and Performance


.
.
.
.
                                              10
10
  M4.1 incorrect Fuse Removal Durina Local Leak Rate Testina
M4.1 incorrect Fuse Removal Durina Local Leak Rate Testina
  a.   Scoce (62707)
a.
        The inspector reviewed an issue that occurred during thr. Unit 2 refueling outage in
Scoce (62707)
        September 1996 as it related to human performance dur;ng localleak rate testing
The inspector reviewed an issue that occurred during thr. Unit 2 refueling outage in
        (LLRT).
September 1996 as it related to human performance dur;ng localleak rate testing
  b.   Observations and Findinas
(LLRT).
        On September 20,1996, with Unit 3 shutdown, the maintenance personnel
b.
        mistakenly pulled the primary containment isolation system (PCIS) inboard and
Observations and Findinas
        outboard mechanical vacuum pump trip logic fuses (F6 fuses in panels 20C041 and
On September 20,1996, with Unit 3 shutdown, the maintenance personnel
        20C042) while working on a localleak rate test activity. This would have caused a
mistakenly pulled the primary containment isolation system (PCIS) inboard and
        these valves to close if they had not been closed at the time. These valves were
outboard mechanical vacuum pump trip logic fuses (F6 fuses in panels 20C041 and
        not primary containment isolation valves, but do receive a PCIS signal to close in
20C042) while working on a localleak rate test activity. This would have caused a
        response to a main steam line isolation, to limit release of radioactive contamination
these valves to close if they had not been closed at the time. These valves were
        from the condenser.
not primary containment isolation valves, but do receive a PCIS signal to close in
        The inspector discussed this activity with a maintenance supervisor with respect to
response to a main steam line isolation, to limit release of radioactive contamination
        the expectations related to fuse manipulations by maintenance personnel. Also, the
from the condenser.
        inspector verified the persons involved had received some training in LLRT and the
The inspector discussed this activity with a maintenance supervisor with respect to
        associated fuse removal. During LLRTs and post-maintenance testing on the main
the expectations related to fuse manipulations by maintenance personnel. Also, the
        steam line sample valves, fuses were to be removed and reinstalled several times.
inspector verified the persons involved had received some training in LLRT and the
        The lead maintenance technician doing the tast had installed the fuses, but he had
associated fuse removal. During LLRTs and post-maintenance testing on the main
        other maintenance personnel remove the fuseo. During this process they removed
steam line sample valves, fuses were to be removed and reinstalled several times.
        the wrong fuses.
The lead maintenance technician doing the tast had installed the fuses, but he had
        PECO initiated a PEP (10006104) to investigate the, cause of the wrong fuse
other maintenance personnel remove the fuseo. During this process they removed
        manipulation. During this evaluation, PECO identified several causes including less
the wrong fuses.
        than adequate use of self-check methods, the workers selected for the ta.sk may not
PECO initiated a PEP (10006104) to investigate the, cause of the wrong fuse
        have been familiar with fuse labeling and wire labelir:g conventions used at PBAPS,
manipulation. During this evaluation, PECO identified several causes including less
        and an inadequate pre-job briefing,
than adequate use of self-check methods, the workers selected for the ta.sk may not
  c.   Conclusions
have been familiar with fuse labeling and wire labelir:g conventions used at PBAPS,
        The maintenance work group personnel error of pulling the wrong fuses during
and an inadequate pre-job briefing,
                                                                  .
c.
        preparations for containment isolation valve LLRT work resulted in a loss of
Conclusions
        electrical feed to the mechanical vacuum isolation valve's, but no actual valve
The maintenance work group personnel error of pulling the wrong fuses during
        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
preparations for containment isolation valve LLRT work resulted in a loss of
        consequences. However, it represents pocr maintenance activity performance.
electrical feed to the mechanical vacuum isolation valve's, but no actual valve
        Unfamiliarity with fuse removal and self-checking methodologies by the personnel
motion resulted. These valves are not primary containment isolation valves and the
        involved, and an inadequate pre-job briefing contributed to the event.
plant was shut down at the time. Therefore, this problem had no safety
  M8     Miscellaneous Maintenance issues (92902)
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)


.
.
.
.
                                                                                                l
11
                                                11
M8.1 Review of Instrument Channel Functional Test Practices
  M8.1 Review of Instrument Channel Functional Test Practices
a. Scope
  a. Scope
During the period, the inspectors completed a review of how PECO implemented the
        During the period, the inspectors completed a review of how PECO implemented the
improved standard technical specification (ITS) definition of an instrument channel
        improved standard technical specification (ITS) definition of an instrument channel
functional test (CFT). The NRC staff became aware of this issue when inspectors
        functional test (CFT). The NRC staff became aware of this issue when inspectors
and PECO ISEG identified some possible interpretation problems with the ITS
        and PECO ISEG identified some possible interpretation problems with the ITS
definition. ISEG Report 96-29, " Review of Channel Functional Tests for Steam Leak
        definition. ISEG Report 96-29, " Review of Channel Functional Tests for Steam Leak
Detection Instrumentation at Peach Bottom Units 2 and 3," questioned whether
        Detection Instrumentation at Peach Bottom Units 2 and 3," questioned whether
channel functional test practices for the HPCI and RCIC system complied with
        channel functional test practices for the HPCI and RCIC system complied with
Peach Bottom ITS requirements. The inspectors reviewed the issue with assistance
        Peach Bottom ITS requirements. The inspectors reviewed the issue with assistance
from the Office of Nuclear Reactor Regulation (NRR).
        from the Office of Nuclear Reactor Regulation (NRR).                                     l
b. Obsgrvations and Findinas
  b. Obsgrvations and Findinas
ISEG and the inspectors found that Peach Bottom did not test all channel relay
        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
        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
        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
        be within the scope of the channel, PECO was complying with the ITS requirements
for channel system functional tests.
        for channel system functional tests.
The NRC found that the issue revolved around the differences in the wording of the
        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.
        ITS and the old custom TS definitions of a CFT.
Old Custom TS CFT definition:
        Old Custom TS CFT definition:
Prior to implementing the ITS, Peach Bottom operated with custom TS. The
        Prior to implementing the ITS, Peach Bottom operated with custom TS. The
previous TS contained a definition for " Instrument or Channel Functional Test,"
        previous TS contained a definition for " Instrument or Channel Functional Test,"
which stated:
        which stated:
"An instrument or channel functional test means the injection of a simulated signal
        "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
        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."
        the proper instrument channel response, alarm and/or initiating action."
NRC review of this definition found that the intent was to demonstrate channel
        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
        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
        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
        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
        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
        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).
        contacts would be included in the ITS required logic system functional test (LSFT).


,
,
  .
.
                                                  12
12
          ITS CFT definition:
ITS CFT definition:
          A CFT is defined in the ITS as follows:
A CFT is defined in the ITS as follows:
          "A CHANNEL FUNCTIONAL TEST shall be the injection of a simulated or actual
"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,
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 ,
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
trips. The CHANNEL FUNCTIONAL TEST may be performed by means of any series
          tested."
of sequential, overlapping, or total channel steps, so that the entire channel is
          In this ITS definition, the "and/or" language from the previous TS was removed and
tested."
          the term " required" was added to characterize the intent of "and/or." However,
In this ITS definition, the "and/or" language from the previous TS was removed and
          that language change may cause the interpretation that all relay contacts associated
the term " required" was added to characterize the intent of "and/or." However,
          with a channel are required to be verified operable during a CFT.
that language change may cause the interpretation that all relay contacts associated
          The purpose of CFT was not intended to test the change of state of all channel
with a channel are required to be verified operable during a CFT.
          relay contacts, in the ITS conversion, the licensee considered the definition change
The purpose of CFT was not intended to test the change of state of all channel
          as administrative and the staff accepted the new definition because this change in
relay contacts, in the ITS conversion, the licensee considered the definition change
          definition was not intended to increase the scope of CFT. Testing of all contacts-
as administrative and the staff accepted the new definition because this change in
          during CFT is considered unnecessary since change of state of one contact could
definition was not intended to increase the scope of CFT. Testing of all contacts-
          infer change of state of all other contacts associated with a relay. The operability
during CFT is considered unnecessary since change of state of one contact could
          of all contacts is assured during LSFT. The probability of failure of individual-
infer change of state of all other contacts associated with a relay. The operability
          contacts, based on past history, is low.
of all contacts is assured during LSFT. The probability of failure of individual-
                                                                                                I
contacts, based on past history, is low.
    c. Conclusion
c. Conclusion
          The staff has not required licensees to monitor all relay contacts during the more
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     ,
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     l
,
          review had led to an improved understanding of regulatory requirements in this area.
the Peach Bottom ITS CFT requirements. The inspectors also concluded that ISEG's
          In order to avoid further misinterpretation of the TS requirements for CFT, the NRC
review had led to an improved understanding of regulatory requirements in this area.
          staff has undertaken an initiative, with industry representatives, to develop a more l
In order to avoid further misinterpretation of the TS requirements for CFT, the NRC
          clearly worded CFT definition in the ITS.                                             ;
staff has undertaken an initiative, with industry representatives, to develop a more
                                                                                                1
clearly worded CFT definition in the ITS.
                                          111. Enaineerina                                     l
111. Enaineerina
    E1   General Engineering Comments
l
    a.   Scone                                                                                 i
E1
          The inspectors reviewed the general support provided by engineering to day-to-day
General Engineering Comments
          operations of both units.
a.
Scone
i
The inspectors reviewed the general support provided by engineering to day-to-day
operations of both units.


                                                          . .                 -               _-
,
,                                                    .
.
  *
. .
                                                                                                  l
-
                                                                                                  l
_-
                                                  13
*
    b.   Conclusions
l
          Engineering department management and system mangers provided very good
13
          support to the E1 and E3 maintenance outage, particularly in the review and
b.
          dispositioning of the crankshaft strain issues discussed in Sections M1.2 and M1.3       l
Conclusions
          above.                                                                                   j
Engineering department management and system mangers provided very good
                                            IV. Plant Support
support to the E1 and E3 maintenance outage, particularly in the review and
    F2   Status of Fire Protection Facilities and Equipment
dispositioning of the crankshaft strain issues discussed in Sections M1.2 and M1.3
    F2.1 Material Condition Insoection and Eauioment Inventories
above.
    a.   Scone
j
          The inspector conducted a walkthrough of the facility with PECO's fire protection       ,
IV. Plant Support
          staff for PBAPS. During the walkthrough, the inspector noted the material condition     l
F2
          of the fire protection equipment, the material condition and housekeeping of the
Status of Fire Protection Facilities and Equipment
          facility, the readiness of the fire protection equipment for use, and discussed the
F2.1
          upgrades of the equipment, which had been completed. The inspector conducted
Material Condition Insoection and Eauioment Inventories
          an inventory check of the fire brigade locker and the hose cart house at the south-
a.
          east corner of the facility. In addition, the inspector reviewed the unified log to     ,
Scone
          evaluate the out-of-service times for the Cardox systems during the current year.
The inspector conducted a walkthrough of the facility with PECO's fire protection
          The inspector also reviewed the records of the following fire protection equipment
,
          functional tests:
staff for PBAPS. During the walkthrough, the inspector noted the material condition
        -e       ST-O-37C-360-2, Rev. 4, " Motor Driven Fire Pump Operability Test,"
of the fire protection equipment, the material condition and housekeeping of the
                  conducted April 2,1997.
facility, the readiness of the fire protection equipment for use, and discussed the
          *       ST-O-37D-370-2, Rev. 6, " Diesel Driven Fire Pump Operability Test,"
upgrades of the equipment, which had been completed. The inspector conducted
                  conducted March 27,1997.
an inventory check of the fire brigade locker and the hose cart house at the south-
          *       RT-F-37G-392-2, Rev. 2, "E-2 Diesel Generator Cardox System Simulated
east corner of the facility. In addition, the inspector reviewed the unified log to
                  Actuation and Air Flow Test," conducted June 5,1996.
,
          *       RT-0-100-990-2, Rev.1, " Participants Record Docuruentatbn for STs, RTs,
evaluate the out-of-service times for the Cardox systems during the current year.
                  Clearances, and Check Off Lists (COLs)."
The inspector also reviewed the records of the following fire protection equipment
    b.   Observations and Findinas
functional tests:
          The material condition and housekeeping in the facility were excellent. No build up
-e
          of trash or combustible material was noted in the plant during the tours. The fire
ST-O-37C-360-2, Rev. 4, " Motor Driven Fire Pump Operability Test,"
          fighting equipment in the plant, both manual and automatic, was observed to be in
conducted April 2,1997.
          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.
ST-O-37D-370-2, Rev. 6, " Diesel Driven Fire Pump Operability Test,"
          The fire protection alarm panels in the control room have been replaced with newer
conducted March 27,1997.
          models. The newer models allow for acknowledging and resetting alarms
*
          individually. This improvement will allow control room operators to acknowledge an
RT-F-37G-392-2, Rev. 2, "E-2 Diesel Generator Cardox System Simulated
          alarm without masking any additional alarms which might occur.
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.


  , _ .   -     - .... .   . -       . . - - . . -         . -     . - - . -       . - . . -   .-.
,
                                                                                                              !
_ .
-
- .... .
. -
. . - - . . -
. -
. - - . -
. - . . -
.-.
'
'
    e
e
4
4
                                                        14
14
!
!
              Some oil and carbon blowout was noted in'the vicinity of the exhaust system and
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
underneath the turbochargers of the emergency diesel generators (EDGs). The
inspector identified this condition.to PECO personnel, who initiated actions to
d
d
              inspector identified this condition.to PECO personnel, who initiated actions to
*
*
              commence cleanup. Due to the constricted access to the area below the
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             {
turbocharger end of the engine, the CO, flooding system for the room is required to
j             be disabled while work is in progress to limit the safety hazard to personnel. The             j
{
j
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.
$
$
              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
;
;
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
^
^
              being tested approximately every two weeks. During the review of the completed                  I
functional tests on the fire pumps, the inspector noted that Section 10.0,
              functional tests on the fire pumps, the inspector noted that Section 10.0,
;
;             " Participants Record," had not been completed. This section provides the initials
" Participants Record," had not been completed. This section provides the initials
              and printed names of the individuals who participated in the test. When the                     ;
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             '
;
j.
inspector questioned this, PECO personnel provided a copy of " Participants Record
'
'
              Documentation for STs, RTs, Clearances and Check Off Lists," RT-0-100-990 2,
'
I             Rev.1. This procedure is performed annually, remaining open for the entire year, to
Documentation for STs, RTs, Clearances and Check Off Lists," RT-0-100-990 2,
j-             permit operations personnel to omit their names and initials in STs, RTs, tagging
I
;              orders and check lists, which they conduct on a daily basis.
Rev.1. This procedure is performed annually, remaining open for the entire year, to
j-
permit operations personnel to omit their names and initials in STs, RTs, tagging
orders and check lists, which they conduct on a daily basis.
;
.
.
l             The testino of the diesel generator CO, flooding sysmn was conducted once every
l
              eighteer ionths. The test verifies the system activation time and discharge
The testino of the diesel generator CO, flooding sysmn was conducted once every
                            s
eighteer
                                                                                                              '
ionths. The test verifies the system activation time and discharge
              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
s
              conduct of the test, the appropriate impairments and fire watches are implemented
duration. F.w test also verifies that the flow path is unobstructed by performing an
              by the procedure.
air blow from the CO, discharge header to the discharge nozzles. During the
              The review of the out-of service times of the Cardox systems, based upon the out-
conduct of the test, the appropriate impairments and fire watches are implemented
              of-service and return-to service times recorded in the unified log, indicated that of
by the procedure.
              the approximately 2,760 hours in the year-to-date (at the time of the inspection),
The review of the out-of service times of the Cardox systems, based upon the out-
              Cardox systems had been out-of-service for approximately 11 hours due to system
of-service and return-to service times recorded in the unified log, indicated that of
              problems. :n contrast, Cardox systems had been removed from service for
the approximately 2,760 hours in the year-to-date (at the time of the inspection),
              personnel p.otection approximately 490 hours. The inspector considered this to be
Cardox systems had been out-of-service for approximately 11 hours due to system
              indicative of highly reliable fire suppression systems.
problems. :n contrast, Cardox systems had been removed from service for
          c.   Conclusions
personnel p.otection approximately 490 hours. The inspector considered this to be
              The inspector concluded, based upon the observed conditions, the results of the
indicative of highly reliable fire suppression systems.
              reviews of the operability tests, and evaluation of Cardox system outage times that
c.
              the fire protection equipment at the PBAPS is in good repair and is ready to perform
Conclusions
              its intended function.
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.
-
_
_


  , . - - - . - .         - . -       . . . . - -.         , . . - . - . .   . . . .-. - . . - . - . - _ - . - - -
, . - - - . - .
- . -
. . . . - -.
, . . - . - . .
. . .
.-. - . . - . - . - _ - . - - -
!
!
                                                                                                                    '
'
;i
;i
S
S
4
4
i                                                                     15
i
1                 F3   Fire Protection Procedures and Documentation
15
                                                                                                                    l
1
                  F3,1 Fire Protection Procedure Reviews
F3
Fire Protection Procedures and Documentation
F3,1
Fire Protection Procedure Reviews
)
,
,
                                                                                                                    )
.
.                      The inspector reviewed the procedures controlling fire protection activities at the
The inspector reviewed the procedures controlling fire protection activities at the
l                     facility to determine what management controls had been developed to prevent fires
l
;                     and rapidly suppress any fire which might occur. The following specific procedures
facility to determine what management controls had been developed to prevent fires
i                     were reviewed:
;
and rapidly suppress any fire which might occur. The following specific procedures
i
were reviewed:
I
I
                      e       Nuclear Generation Group Policy No. NP-FP-1, Rev. O, Fire Protection
e
Nuclear Generation Group Policy No. NP-FP-1, Rev. O, Fire Protection
e
FF-01, Rev. 3, Fire Brigade
4
4
                      e        FF-01, Rev. 3, Fire Brigade
2
2
                      e       A-C-920, Rev. O, Nuclear Generation Group Fire Protection Program
e
;                     e       AG-CG-012.02, Rev.1, Control of Combustible and Flammable Materials
A-C-920, Rev. O, Nuclear Generation Group Fire Protection Program
!                     e       AG-CG-012.01, Rev. 2, Actions for Fire Protection impairments
;
e
AG-CG-012.02, Rev.1, Control of Combustible and Flammable Materials
!
e
AG-CG-012.01, Rev. 2, Actions for Fire Protection impairments
*
*
                      e       AG-CG-012, Rev. O, Hot Work Guideline                                               ,
e
                      o       NE-C-250, Rev.1, Fire Protection Review
AG-CG-012, Rev. O, Hot Work Guideline
                      e       NE-C-250-2, Rev. O, Fire Protection Review Checklist
,
                      e       HZ C-5-4, Rev.1, Safety Storage Equipment
o
                      e       HZ-C-5-1, Rev.-1, Request for Permanent Chemical Storage
NE-C-250, Rev.1, Fire Protection Review
                  b. Observations and Findinas
e
                      The fire protection policy statement and administrative procedures have been
NE-C-250-2, Rev. O, Fire Protection Review Checklist
                      revised and reissued, with new nurnbers, since the last fire protection inspection.
e
                      This effort was conducted, in part, to provide for standardization between PECO's
HZ C-5-4, Rev.1, Safety Storage Equipment
                      nuciear stai!ons at Limerick and Peach Bottom. This will aid in ensuring compliance
e
                      with requirements as personnel move between the sites in response to outage and
HZ-C-5-1, Rev.-1, Request for Permanent Chemical Storage
                      specialty maintenance requirements. The exception to this is FF-01, " Fire Brigade,"
b.
                      which remains site specific for Peach Bottom.
Observations and Findinas
                      A C-920, Nuclear Generation Group Fire Protection Program
The fire protection policy statement and administrative procedures have been
                      This procedure is a common procedure between the stations and provides the broad
revised and reissued, with new nurnbers, since the last fire protection inspection.
                      administrative guidelines for fire protection and extends the concept of defense-in-
This effort was conducted, in part, to provide for standardization between PECO's
                      depth to fire protection activities. This procedure provides the overall goals to be
nuciear stai!ons at Limerick and Peach Bottom. This will aid in ensuring compliance
                      achieved by the fire protection program, and specifies lower tier procedures for
with requirements as personnel move between the sites in response to outage and
                      carrying out specific functions within the program. This procedure also provides
specialty maintenance requirements. The exception to this is FF-01, " Fire Brigade,"
                      general guidance for responding to fires, control of hot work, and review of all work
which remains site specific for Peach Bottom.
                      activities to ensure that protection against fires has been included in the work
A C-920, Nuclear Generation Group Fire Protection Program
                      planning.
This procedure is a common procedure between the stations and provides the broad
                      AG-CG-012, Hot Work Guideline
administrative guidelines for fire protection and extends the concept of defense-in-
                      This procedure minimizes potential fire hazards by specifying a hot work permit
depth to fire protection activities. This procedure provides the overall goals to be
                      system to control work involving ignition hazards. The procedure specifies those
achieved by the fire protection program, and specifies lower tier procedures for
                      areas wherein a hot work permit is required, lists examples of typical ignition
carrying out specific functions within the program. This procedure also provides
                      sources, and provides instructions for generating hot work permits. This procedure
general guidance for responding to fires, control of hot work, and review of all work
                      also assigns responsibilities to the personnel involved, such as the hot 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


                                  , . - . - . - .   .       ._ - ..     - - - - - . - _         -
, . - . - . - .
                                                                                                      . . - . . - . -
.
  .
._ - ..
                                                            16
- - - - - . - _
                  opf rator, fire watch, work supervisor, and industrial risk management (IRM)
-
                  penonnel. The procedure also requires that communications equipment, such as                         j
. . - . . - . -
                  r    7t page, plant phone, or portable radio, is available in the immediate vicinity of             l
.
                  the hot work area. This last requirement was added as part of the corrective                         l
16
                  actions for the fire, which occurred at the site on August 10,1994 (see Section F                     !
opf rator, fire watch, work supervisor, and industrial risk management (IRM)
                  7.2).                                                                                                 '
penonnel. The procedure also requires that communications equipment, such as
                  AG-CG-012.01, Actions for Fire Protection impairments                                               l
j
                  This procedure providos guidance for reporting, tracking, and ensuring restoration of
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).
'
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
i
,
,
                  fire protection systems and features required by the NRC or American Nuclear
-
                                                                      -
insurers (ANI). The requirements of this procedure include designation and
                                                                                                                      i
i
                  insurers (ANI). The requirements of this procedure include designation and                           i
;
;                 scheduling of fire watch personnel by the work supervisor who is responsible for                     i
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
j
,
,
                  the work which will cause the impairment (security force provides fire watches for                  j
-
-
                  impairments caused by broken equipment). Shift management is responsible to                         J
impairments caused by broken equipment). Shift management is responsible to
                  perform a review of compensatory actions prior to issuing the impairment, and the
J
perform a review of compensatory actions prior to issuing the impairment, and the
work group is responsible for implementing the compensatory actions before taking
-
-
                  work group is responsible for implementing the compensatory actions before taking
the equipment out of service. For those situations where there are questions
                  the equipment out of service. For those situations where there are questions
i
i                 regarding the effect on fire protection systems or features, IRM provides resolution.
regarding the effect on fire protection systems or features, IRM provides resolution.
l
l
AG-CG-012.02, Control of Combustible and Flammable Materials
1
,
,
                  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
                  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
                  the use of flammable or combustible materials in the plants. The procedure
provides for the reporting of fire hazards, control and storage of combustible
                  provides for the reporting of fire hazards, control and storage of combustible
materials, designation of combustible free zones to prevent the spread of fire,
                  materials, designation of combustible free zones to prevent the spread of fire,
l
l                 controls on smoking in the faci!ity, and disposal guidelines for combustible
controls on smoking in the faci!ity, and disposal guidelines for combustible
                  materials.
l
l
materials.
<
<
                  NE-C 250, Fire Protection Review
NE-C 250, Fire Protection Review
                  This procedure provides guidelines for conducting fire protection reviews of
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,
modification packages, and for completing the fire protection checklist. The
i-               which require prior review by the fire protection / safe shutdown (FP/SS) branch prior
procedure also establishes threshold levels of changes in combustible loadings,
l                 to implementation of the modification. Changes less than these threshold values
i-
                  are reported to the FP/SS branch for incorporation into controlled documents.
which require prior review by the fire protection / safe shutdown (FP/SS) branch prior
l
to implementation of the modification. Changes less than these threshold values
are reported to the FP/SS branch for incorporation into controlled documents.
.
.
4                FF-01, Fire Brigade
FF-01, Fire Brigade
                                                                                                                      )
4
i                 This procedure provides guidelines for responding to a fire at the station, both
)
i
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
i
considered to be general guidance, and the brigade leader is permitted to adjust the
j
actions of the brigade to suit the conditions encountered during each individual
response.
<
:
:
.                within and without the protected area. Responsibilities for specific personnel are                  ;
                  given, and requirements for brigade response are listed. The procedure is                            i
                  considered to be general guidance, and the brigade leader is permitted to adjust the
j                actions of the brigade to suit the conditions encountered during each individual
<
                  response.
:
;
;
r
r
                                                                                                                      .
.
  ., - , , . , -                                       ..,-                                                     .-
.,
-
,
, . , -
..,-
.-


  e-
e-
. .
. .
                                                  17
17
      c. Conclusions
c.
          Based upon the results of the procedure reviews, the inspector concluded that the
Conclusions
          fire protection procedures provide good guidance and controls to prevent fires,
Based upon the results of the procedure reviews, the inspector concluded that the
          maintain the fire fighting capabilities of the organization, and respond to any fires
fire protection procedures provide good guidance and controls to prevent fires,
          which might occur.
maintain the fire fighting capabilities of the organization, and respond to any fires
      F5   Fire Protection Staff Training and Qualification
which might occur.
      F5.1 Fire Briaade Trainina
F5
      a. Scope
Fire Protection Staff Training and Qualification
          The inspector observed a fire brigade training session for health physics personnel;
F5.1
          reviewed lesson plan PHPCT-94-02C, Rev. No. 001, "HP Introduction to Fire
Fire Briaade Trainina
          Brigade Response," dated April 19,1995; reviewed the documentation of the first
a.
          quarter 1997 fire brigade meeting; computerized records of fire brigade training for
Scope
          the past two years; and Issue Evaluation Report 10002682, " Unit 2 Vent Stack Fire
The inspector observed a fire brigade training session for health physics personnel;
          and Unusual Event Declaration," dated December 19,1994.
reviewed lesson plan PHPCT-94-02C, Rev. No. 001, "HP Introduction to Fire
      b. Observations and Findinas
Brigade Response," dated April 19,1995; reviewed the documentation of the first
          As discussed in Section F7.2, during the vent stack fire of August 1994, some
quarter 1997 fire brigade meeting; computerized records of fire brigade training for
          confusion existed regarding the chain of command due to the presence of senior
the past two years; and Issue Evaluation Report 10002682, " Unit 2 Vent Stack Fire
          operations management personnel on the refueling floor. To alleviate the problem,
and Unusual Event Declaration," dated December 19,1994.
          PECO training for fire brigade responders emphasizes that the fire brigade leader is
b.
          the on-scene commander at the fire. All other personnel at the scene are under the
Observations and Findinas
          direction of the brigade leader.
As discussed in Section F7.2, during the vent stack fire of August 1994, some
          The training focuses on fire field activities and fire response command structure.
confusion existed regarding the chain of command due to the presence of senior
          The training indicates that the duties of the health physics (HP) technician
operations management personnel on the refueling floor. To alleviate the problem,
          responding to the scene of the fire are the same as those normally performed,
PECO training for fire brigade responders emphasizes that the fire brigade leader is
          namely evaluating the radiological hazards and proposing protective measures for
the on-scene commander at the fire. All other personnel at the scene are under the
          the personnel involved.
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
i
          The training records for the fire brigade members and fire brigade leaders are
The training records for the fire brigade members and fire brigade leaders are
          maintained in the plant information management system (PIMS). The training status
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
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
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
for the shift personnel to determine training and qualification status when making
          fire brigade assignments.
fire brigade assignments.
      c. Conclusions
c.
          Based upon the review of lesso7 plans, observation of training in progress, and
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
4
          review of the computer printout of training and qualification records, the inspector
concluded that the training is sufficient to ensure that an adequate pool of trained
          concluded that the training is sufficient to ensure that an adequate pool of trained


,-
,-
.
.
                                                  18
18
          and qualified personnel are available to staff the fire brigade. In addition, having the
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
training and qualification status available on PIMS aids in making assignments to fire
          brigade positions and determining the training needs.
brigade positions and determining the training needs.
  F.5.2 (OcenIFl 50-277/278/97-04-Oli Fire Watch Trainina Revision to Fire Protection
F.5.2 (OcenIFl 50-277/278/97-04-Oli Fire Watch Trainina Revision to Fire Protection
          Trainina Lesson Plans
Trainina Lesson Plans
    a.   Scope
a.
          The inspector reviewed computerized records of fire watch training, issue evaluation
Scope
          report 10002682, " Unit 2 Vent Stack Fire and Unusual Event Declaration"; lesson
The inspector reviewed computerized records of fire watch training, issue evaluation
          plan MCTR-1075, Rev. 2, " Hot Work (Ignition Source) Firewatch Training"; and
report 10002682, " Unit 2 Vent Stack Fire and Unusual Event Declaration"; lesson
          lesson plan MCTJ-0035R, " Portable Fire Extinguisher Usage Requal."
plan MCTR-1075, Rev. 2, " Hot Work (Ignition Source) Firewatch Training"; and
    b.   Observations and Findinas
lesson plan MCTJ-0035R, " Portable Fire Extinguisher Usage Requal."
          The inspector's review of the firewatch training lesson plan determined that the
b.
          training covers the responsibilities of the ignition source fire watch. The training
Observations and Findinas
          also stresses that the firewatch should be actively engaged in setting up the hot
The inspector's review of the firewatch training lesson plan determined that the
          work area to ensure that all sparks, slag, and molten metal are contained in the hot
training covers the responsibilities of the ignition source fire watch. The training
          work area.
also stresses that the firewatch should be actively engaged in setting up the hot
          Issue evaluation report 10002682 reviewed the fire which occurred in the Unit 2
work area to ensure that all sparks, slag, and molten metal are contained in the hot
          plant vent stack on August 10,1994, for lessons learned. One of the identified
work area.
          problems was the lack of communications equipment on the building roofs. The
Issue evaluation report 10002682 reviewed the fire which occurred in the Unit 2
          nearest plant communications equipment available to the firewatch at the scene of
plant vent stack on August 10,1994, for lessons learned. One of the identified
          the fire was approximately 75 feet down the scaffolding and across the turbine
problems was the lack of communications equipment on the building roofs. The
          floor. The future corrective actions for this event included revising the hot work
nearest plant communications equipment available to the firewatch at the scene of
          firewatch lesson plan to include lessons learned. The revision of the lesson plan
the fire was approximately 75 feet down the scaffolding and across the turbine
          should have included the need for communications equipment in the immediate
floor. The future corrective actions for this event included revising the hot work
          vicinity of the hot work area, including the use of a portable radio if plant page or
firewatch lesson plan to include lessons learned. The revision of the lesson plan
          phone are not available. Instead, the lesson plan still reads that the firewatch is
should have included the need for communications equipment in the immediate
          responsible to know "... the location of the nearest page or phone, and fire alarm
vicinity of the hot work area, including the use of a portable radio if plant page or
          pullstation." The issue evaluation report is also not listed in the references section
phone are not available. Instead, the lesson plan still reads that the firewatch is
          of the lesson plan. When these deficiencies were identified to the fire protection
responsible to know "... the location of the nearest page or phone, and fire alarm
          staff, actions were initiated to make appropriate changes to the lesson plan. This
pullstation." The issue evaluation report is also not listed in the references section
          will require action by the Limerick generating station fire protection staff, since they
of the lesson plan. When these deficiencies were identified to the fire protection
          are responsible for this lesson plan. This revision will be verified during the next fire
staff, actions were initiated to make appropriate changes to the lesson plan. This
          protection program review (IFl 50-277, 278/97-04-02).                                     '
will require action by the Limerick generating station fire protection staff, since they
    c.     Conclusions
are responsible for this lesson plan. This revision will be verified during the next fire
          Based on the review of the lesson plans and the review of the computer printout of
protection program review (IFl 50-277, 278/97-04-02).
          training records, the inspector concluded that the training is sufficient to ensure that
'
          qualified and knowledgeable personnel are available for performing hot work fire
c.
          watch duties. In addition, having the training records available on PIMS ensures
Conclusions
          that work supervisors can easily determine the qualification status and training
Based on the review of the lesson plans and the review of the computer printout of
          needs of personnel in their group.                                                       i
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.
i


  _ _ _ .   _.             -.   _ -_.._ _ _           _   _ _ . _ _     . . _ _ _ _ . _ . _ _ _ _ .                     _ _
_ _ _
    e
.
                                                              19
_.
          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
_ _
l                 actions. The review included the following audits:
e
!                 e              Assessment Report No. A0808620, " Assessment of Fire Protection Program,
19
j                                 MAP Area D02," dated December 15,1994.
F7
                  e             Assessment Report No. A0900835, "PBAPS Triennial Fire Protection
Quality Assurance in Fire Protection
!                                 Program, MAP Area D02, Rev. 8," dated September 25,1995.
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
l
actions. The review included the following audits:
!
Assessment Report No. A0808620, " Assessment of Fire Protection Program,
e
j
MAP Area D02," dated December 15,1994.
e
Assessment Report No. A0900835, "PBAPS Triennial Fire Protection
!
Program, MAP Area D02, Rev. 8," dated September 25,1995.
4
4
                  e              Assessment Report No. A0967111, "PBAPS Annual Fire Protection Program
Assessment Report No. A0967111, "PBAPS Annual Fire Protection Program
j                                 Assessment, MAP Area D02, Rev. 9," dated July 17,1996.
e
i                 e             .PEPlssue 10005802, " Deterioration of CO, Extinguisher Hoses," initiated
j
j'                               June 20,1996.
Assessment, MAP Area D02, Rev. 9," dated July 17,1996.
                                                                                                                                i
i
            b.   Observations and Findinas
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
l_
i                 administrative controls, the status and condition of fire protection equipment, fire
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
l
brigade training, corrective actions and self-assessment activities. Outside expertise
'
'
                  brigade training, corrective actions and self-assessment activities. Outside expertise
was used in the performance of the triennial program assessment. The assessment
                  was used in the performance of the triennial program assessment. The assessment
j
j                 plans were reviewed and concurred in by the Nuclear Review Board (NRB). In
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
3
l                 specific issues, such as ALARA considerations in the placement of continuous fire
several cases, members of the NRB requested the assessment team to look into
j                 watch stations and priority for clearing fire system impairments.
l
l                 The audits generally concluded that the program was being conducted in a safe and
specific issues, such as ALARA considerations in the placement of continuous fire
j                 effective manner. The audits determined that the number of outstanding                                         i
j
j                 maintenance items against fire protection equipment had decreased over the three
watch stations and priority for clearing fire system impairments.
L                 year period.
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
$
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,
'
'
                  team's facility walkdowns, appropriate corrective actions were initiated. One issue
conducted under PEP issue 10005802, initiated June 20,1996, determined that the
;                was related to six CO, fire extinguishers being identified as having cracked hoses-
hoses were removed from the CO, extinguishers when the extinguishers were sent
                  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
3
                  out for hydrostatic testing. The hoses were placed onto extinguishers, which were
out for hydrostatic testing. The hoses were placed onto extinguishers, which were
J                 being returned from testing without the complete inspection required by the
J
{                 procedure. The use of procedures by the particular workgroup involved was
being returned from testing without the complete inspection required by the
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
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
1
i
i
.
.
.
.
I
I
    -           _ - - , - - . _               -. _                     .. -                         - -._ _ - - - _ _ _ .
-
_ - - , - - . _
-.
_
.. -
-
-.
_
- - - _ _ _ .


v
v
  e
e
                                                                                                  l
20
                                                  20
turnout gear missing. An action request was issued to replace the missing
          turnout gear missing. An action request was issued to replace the missing
equipment. In addition, PECO established a tracking system to determine if an
          equipment. In addition, PECO established a tracking system to determine if an
adverse trend developed regarding missing fire protection equipment.
          adverse trend developed regarding missing fire protection equipment.
c.
    c.   Conclusions
Conclusions
          Based upon the audits finding only minor deficiencies, and the plant staff taking
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
corrective actions to address the findings, the inspector concluded that the audits
          were effective in identifying problems and causing corrective action to be taken.
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
F7.2 Review of Auaust 10.1994 Vent Stack Fire Corrective Actions - Unit 2
    a.   Scone
a.
          The inspector reviewed the circumstances surrounding the August 10,1994, fire
Scone
          on-site to determine what conclusions PECO had drawn with regard to the cause of
The inspector reviewed the circumstances surrounding the August 10,1994, fire
          the fire, and what actions PECO initiated to improve the performance of the fire
on-site to determine what conclusions PECO had drawn with regard to the cause of
          fighting organization as a result. This was identified as the only fire in the facility
the fire, and what actions PECO initiated to improve the performance of the fire
          since the last inspection.
fighting organization as a result. This was identified as the only fire in the facility
          The inspector reviewed the following documents during the course of the review:
since the last inspection.
          e       NRC Combined Inspection Report 50-277/94-13 and 50-278/94-13, dated
The inspector reviewed the following documents during the course of the review:
                  September 19,1994.
e
          e       Licensee Event Report (LER) 50-277/94-07," Secondary Containment
NRC Combined Inspection Report 50-277/94-13 and 50-278/94-13, dated
                  Breached to Fight Fire," dated September 8,1994.
September 19,1994.
          e       Performance Enhancement Program (PEP) Issue 10002682, " Fire in U/2 Rx
e
                  Building Roof Vent Stack Caused by Weld Spark," initiated August 10,1994.
Licensee Event Report (LER) 50-277/94-07," Secondary Containment
          e       issue Evaluation Report 10002682, " Unit 2 Vent Stack Fire and Unusual
Breached to Fight Fire," dated September 8,1994.
                  Event Declaration," dated December 19,1994.
e
    b.   Observations and Findinas
Performance Enhancement Program (PEP) Issue 10002682, " Fire in U/2 Rx
          The circumstances which led to the fire are briefly described in NRC Inspection
Building Roof Vent Stack Caused by Weld Spark," initiated August 10,1994.
          Report 50-277&278/94-13. During modification work to upgrade the Unit 2
e
          ventilation stack radiation monitor, new pipe hangars were required to be attached
issue Evaluation Report 10002682, " Unit 2 Vent Stack Fire and Unusual
          to the structural steel for the vent stack. During welding, sparks escaped the
Event Declaration," dated December 19,1994.
          materialintended to contain them and ignited insulation and/or bird nests within the
b.
          stack's double wall. Attempts by the fire watch to extinguish the fire with a dry
Observations and Findinas
          chemical extinguisher were not successful, and the fire brigade was activated. The
The circumstances which led to the fire are briefly described in NRC Inspection
          fire brigade opened a hatch in the roof of the reactor building to bring a hose to the
Report 50-277&278/94-13. During modification work to upgrade the Unit 2
          scene. The fire was rapidly extinguirhed using a hose stream.
ventilation stack radiation monitor, new pipe hangars were required to be attached
          During their review of the event, PECO identified two contributing causes for the
to the structural steel for the vent stack. During welding, sparks escaped the
          fire and twelve extraneous conditions adverse to quality (ECAQ). The contributing
materialintended to contain them and ignited insulation and/or bird nests within the
          causes were determined to be lack of understanding of the requirements for staging
stack's double wall. Attempts by the fire watch to extinguish the fire with a dry
          the area on the part of the work supervisor, and the fire watch being provided with
chemical extinguisher were not successful, and the fire brigade was activated. The
          an incorrect type of extinguisher. The ECAQs included confusion regarding who
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
21
        was in control at the scene of the fire due to the presence of senior operations
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
management personnel, lack of plant communications equipment at the hotwork job
        site, misunderstanding on the part of the work supervisor regarding what
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
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
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
identified problems. These actions are discussed further under Section FS, " Fire
        Protection Staff Training and Qualification."
Protection Staff Training and Qualification."
    c. Conclusions
c.
        Based on the corrective actions carried out for the observed deficiencies during the
Conclusions
        program audits and fire fighting activities on August 10,1994, the inspector
Based on the corrective actions carried out for the observed deficiencies during the
        concluded that PECO is effectively identifying and correcting problems with fire
program audits and fire fighting activities on August 10,1994, the inspector
        protection activities.
concluded that PECO is effectively identifying and correcting problems with fire
    F8   Miscellaneous Fire Protection issues
protection activities.
    F8.1 Conformance to Uodated Final Safety Analysis Report Description
F8
    a. Scoce
Miscellaneous Fire Protection issues
        The inspector reviewed the Facility Operating Licenses DPR-44 and DPR-56; the
F8.1
        Peach Bottom Fire Protection Plan; the Peach Bottom Updated Final Safety Analysis
Conformance to Uodated Final Safety Analysis Report Description
        Report; and NRC Safety Evaluation Reports dated August 24,1994, September 16,
a.
        1993, November 24,1980, October 10,1980, September 15,1980, August 14,
Scoce
        1980, and May 23,1979.
The inspector reviewed the Facility Operating Licenses DPR-44 and DPR-56; the
    b. Observations and Findinas
Peach Bottom Fire Protection Plan; the Peach Bottom Updated Final Safety Analysis
        The fire protection requirements were transferred from the PBAPS technicai
Report; and NRC Safety Evaluation Reports dated August 24,1994, September 16,
        specifications to the UFSAR by Amendment No. 210 to License DPR-44, and
1993, November 24,1980, October 10,1980, September 15,1980, August 14,
        Amendment 214 to License DPR-56. The description is contained in the fire
1980, and May 23,1979.
        protection plan, which is incorporated into Section 10.12 of the UFSAR by
b.
        reference. The inspector determined that the fire protection program conforms to
Observations and Findinas
        the description in the UFSAR.
The fire protection requirements were transferred from the PBAPS technicai
        During the inspector's walkthroughs of the facility, particular attention was paid to
specifications to the UFSAR by Amendment No. 210 to License DPR-44, and
        fixed suppression systems and fire protection features described in the fire
Amendment 214 to License DPR-56. The description is contained in the fire
        protection plan and NRC safety evaluation reports. The fixed suppression systems
protection plan, which is incorporated into Section 10.12 of the UFSAR by
        and other features described in the fire protection plan and safety evaluation reports
reference. The inspector determined that the fire protection program conforms to
        have been maintained in effect.
the description in the UFSAR.
    c. Conclusions
During the inspector's walkthroughs of the facility, particular attention was paid to
        Based on the inspector's observation of the fixed suppression systems and other
fixed suppression systems and fire protection features described in the fire
        features described in the fire protection plan and the safety evaluation reports, the
protection plan and NRC safety evaluation reports. The fixed suppression systems
        inspector determiner.1 that the fire protection systems conform to the descriptien in
and other features described in the fire protection plan and safety evaluation reports
        the UFSAR.
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.


, _.   ___ _ __               _ _ _ _       _..   _   ._       _ _ _       _ _ _ _ _        _ ._ ____ _ _
, _.
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
0
                PECO implemented a radiological controls organization change in late 1996
22
                involving temporary assignment of the radiological engineering manager to the                 ;
R1
                radiation protection manager position. The acting radiation protection manager                 !
Radiological Protection and Chemistry (RP&C) Controls
                (RPM) met applicable qualification guidance of Regulatory Guide 1.8. and was later             ;
R1.1 Radioloaical Controls (Proaram Chances)
                selected as RPM.                                                                               I
a.
                A new radiological engineering manager was selected. The individual met
Scope (80750)
                applicable experience requirements, At the time of the inspection, the individual
The inspector reviewed selected radiological controls program changes. Areas
                was not familiar with applicable program procedures and current industry guidance
reviewed included organization and staffing, facilities and equipment, and procedure
                in the areas to be managed. Consequently, an experienced individual within the
changes.
                radiological engineering group was temporarily promoted to acting manager pending
b.
                the completion, by the newly selected manager, of a familiarization program for
Observations and Findinas
                department procedures and industry standards.
e
                An experienced individual was acting in the capacity as manager, technical support
Organization, Staffing, Training and Qualification
                in the absence of the incumbent due to the incumbent's temporary assignment at
PECO implemented a radiological controls organization change in late 1996
                another licensee's facility.
involving temporary assignment of the radiological engineering manager to the
                During a previous inspection, a new individual was selected to provide training of
radiation protection manager position. The acting radiation protection manager
                station personnel in the area of radioactive material shipping. This individual
(RPM) met applicable qualification guidance of Regulatory Guide 1.8. and was later
                appeared to have limited experience and training in the area. The licensee provided
selected as RPM.
                training to compensate for the individual's limited experience and knowledge in this
A new radiological engineering manager was selected. The individual met
                area. An experienced individual was used to provide training in the interim.
applicable experience requirements, At the time of the inspection, the individual
                e       Review of Dosimetry Equipment
was not familiar with applicable program procedures and current industry guidance
                During a previous inspection, the inspector noted that the licensee was
in the areas to be managed. Consequently, an experienced individual within the
                encountering difficulty with the low energy beta response of its vendor supplied
radiological engineering group was temporarily promoted to acting manager pending
                dosimetry. Although the test results of the dosimetry met applicable national
the completion, by the newly selected manager, of a familiarization program for
                testing criteria, the error associated with the test results was higher than the
department procedures and industry standards.
                licensee wished to accept. Consequently, the licensee recently changed its
An experienced individual was acting in the capacity as manager, technical support
                dosimetry vendor. The inspector reviewed applicable information and determined
in the absence of the incumbent due to the incumbent's temporary assignment at
                that the new dosimetry system met the requirements of 10 CFR 20.1501 relative to
another licensee's facility.
                accreditation by national testing standards. The new dosimetry was noted to
During a previous inspection, a new individual was selected to provide training of
                                                                        w
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


,
,
                                                                                                  1
O
  O
23
                                                                                                  I
. exhibit improved dosimetry performance (i.e., lower error) and was accredited in all
                                                  23
radiation test categories of the national standard.
        . exhibit improved dosimetry performance (i.e., lower error) and was accredited in all
o
          radiation test categories of the national standard.
Programs and Procedures
          o       Programs and Procedures
The licensee implemented hydrogen water chemistry and injection of depleted zinc
          The licensee implemented hydrogen water chemistry and injection of depleted zinc
into the reactor coolant system. (See Section R1.2.b of this report.)
          into the reactor coolant system. (See Section R1.2.b of this report.)
c.
    c.   Conclusions
Conclusions
                                                                                                    1
No program changes were identified that reduced the effectiveness of the
          No program changes were identified that reduced the effectiveness of the               '
'
          radiological controls program.
radiological controls program.
                                                                                                  1
1
                                                                                                  !
!
          No safety concerns or violations were identified.                                       ,
No safety concerns or violations were identified.
                                                                                                  I
,
    R1.2 ALARA Proaram and Unit 3 Refuelina Outaae Plannina, Preparation. Emeraent Work
I
                                                                                                  "
R1.2 ALARA Proaram and Unit 3 Refuelina Outaae Plannina, Preparation. Emeraent Work
          Control
"
    a.   Scope (83750)
Control
          The inspector selectively reviewed various ALARA program elements and reviewed           I
a.
          the planning and preparation for the Unit 3 refueling outage, including control and
Scope (83750)
          review of emergent work. The inspector reviewed records, discussed outage
The inspector selectively reviewed various ALARA program elements and reviewed
          planning, and observed activities to verify necessary planning, preparations, and
the planning and preparation for the Unit 3 refueling outage, including control and
          management support for the implementation of radiological controls. The inspector
review of emergent work. The inspector reviewed records, discussed outage
          reviewed lessons learned from previous outages to determine if they were
planning, and observed activities to verify necessary planning, preparations, and
          incorporated into planning and preparations for future outages.
management support for the implementation of radiological controls. The inspector
    b.   Observations and Findinas
reviewed lessons learned from previous outages to determine if they were
          The licensee continued to implement initiatives to reduce overall occupational
incorporated into planning and preparations for future outages.
          radiation exposure. The licensee recently implemented injection of " depleted zinc"
b.
          in order to reduce the dose rates attributable to production of Zinc-65. The licensee
Observations and Findinas
          had started injection of natural zine into the coolant in 1991 (Unit 2) to reduce
The licensee continued to implement initiatives to reduce overall occupational
          piping degradation. However, use of natural zine results in production of Zinc-65
radiation exposure. The licensee recently implemented injection of " depleted zinc"
          which increases drywell radiation dose rates. The licensee implemented use of the
in order to reduce the dose rates attributable to production of Zinc-65. The licensee
          depleted zinc at both units on October 28,1996, and expects about a 25%
had started injection of natural zine into the coolant in 1991 (Unit 2) to reduce
          reduction in drywell radiation dose rates (recirculation piping) in 3 to 4 fuel cycles.
piping degradation. However, use of natural zine results in production of Zinc-65
          The licensee continued to implement other activities to reduce unnecessary
which increases drywell radiation dose rates. The licensee implemented use of the
          occupational exposure including use of robotics, hot spot reductions, permanent
depleted zinc at both units on October 28,1996, and expects about a 25%
          shielding (e.g., scram discharge headers), station modifications, enhanced use of
reduction in drywell radiation dose rates (recirculation piping) in 3 to 4 fuel cycles.
          video cameras, and improved timeliness of drywell shielding installation. The
The licensee continued to implement other activities to reduce unnecessary
          licensee also installed station and component pictures on the station's local area
occupational exposure including use of robotics, hot spot reductions, permanent
          network for viewing during work planning. The licensee also enhanced its
shielding (e.g., scram discharge headers), station modifications, enhanced use of
          benchmarking of individual job tasks and has obtained a " gamma camera" for use in
video cameras, and improved timeliness of drywell shielding installation. The
          identifying elevated dose rate areas. The licensee develops reasonable occupational
licensee also installed station and component pictures on the station's local area
          exposure goals and meets those goals.
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.


~
~
  O
O
                                                    24                                             I
24
                                                                                                  !
The licensee also implemented ute of hydrogen water chemistry to limit system
          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
          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
          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
          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
            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
          increases. 'The licensee developed applicable radiation dose rate limits at various
locations at the station (e.g., site boundary, controlled area, restricted area) to
          locations at the station (e.g., site boundary, controlled area, restricted area) to
ensure conformance with applicable regulatory requirements. The licensee plans to
          ensure conformance with applicable regulatory requirements. The licensee plans to       j
j
          perform a TLD study to evaluate dose rate increases.                                   ,
perform a TLD study to evaluate dose rate increases.
                                                                                                  I
,
    c.   Conclusions
c.
          The licensee implemented an overall effective ALARA program. The inspector noted
Conclusions
          very good ALARA plans for significant radiological work activities. PECO
The licensee implemented an overall effective ALARA program. The inspector noted
          implemented overall effective ALARA planning for the Unit 3 refueling outage         .
very good ALARA plans for significant radiological work activities. PECO
          including emergent work.                                                               !
implemented overall effective ALARA planning for the Unit 3 refueling outage
          No safety concerns or violations were noted.
.  
    R1.3 Internal Exoosure Controls
including emergent work.
    a.   Scooe (83750)
!
          The inspector selectively examined the internal exposure control program. The
No safety concerns or violations were noted.
          inspector reviewed records, discussed the program with cognizant personnel and
R1.3 Internal Exoosure Controls
          observed exposure control practices during tours of the RCA.
a.
    b.   Observations and Findinas
Scooe (83750)
          There were no recorded internal exposures during the past two years. The               !
The inspector selectively examined the internal exposure control program. The
          inspector noted that the licensee's effective control of airborne radioactivity has
inspector reviewed records, discussed the program with cognizant personnel and
          resulted in a substantial reduction in use of respiratory protection equipment.         l
observed exposure control practices during tours of the RCA.
          Licensee data indicated respirator usage has declined from about 17,000 respirators
b.
          worn in 1991 to approximately 54 respirators worn in 1996, which included a
Observations and Findinas
          refueling outage. The licensee performed appropriate internal dose calculations and
There were no recorded internal exposures during the past two years. The
          DAC-hours were calculated and tracked, as necessary.
inspector noted that the licensee's effective control of airborne radioactivity has
          The inspector observed an individual being fit-tested for a respirator and checked
resulted in a substantial reduction in use of respiratory protection equipment.
          applicable fit test machine calibrations. The inspector noted that the printout from
Licensee data indicated respirator usage has declined from about 17,000 respirators
          the fit test machine referenced the incorrect national standard for calculation of
worn in 1991 to approximately 54 respirators worn in 1996, which included a
          respirator fit factor. The licensee initiated a review of this matter.
refueling outage. The licensee performed appropriate internal dose calculations and
          The following area for improvement was identified:
DAC-hours were calculated and tracked, as necessary.
          The inspector selectively reviewed the program for estimating internal exposure to
The inspector observed an individual being fit-tested for a respirator and checked
          transuranics (e.g., alpha emitters) which are not readily detectable by invivo
applicable fit test machine calibrations. The inspector noted that the printout from
          bioast,ay (e.g., whole body counting). The inspector did not identify any procedural
the fit test machine referenced the incorrect national standard for calculation of
          guidance for personnel to use to readily ascertain that if an intake of transuranics
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
O
                                                  25
25
          may have occurred, properly estimate the intake by evaluation of breathing zone air
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
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.
invitro bioassays (e.g., fecal analysis) of samples collected in a timely manner.
    c.   Conclusions
c.
          PECO implemented an effective internal exposure control program. However,
Conclusions
          although the licensee did not have any current problem with transuranics, there was
PECO implemented an effective internal exposure control program. However,
          no clearly defined program to perform internal dose assessment for these
although the licensee did not have any current problem with transuranics, there was
          radionuclides in the event of their appearance (e.g., following fuel failures).
no clearly defined program to perform internal dose assessment for these
          No violations were noted.
radionuclides in the event of their appearance (e.g., following fuel failures).
    R1.4 External Exoosure Controls: (Ocen) Violation 97-04-01: Inadeauste Controls over
No violations were noted.
          Locked Hiah Radiation Door Kevs
R1.4 External Exoosure Controls: (Ocen) Violation 97-04-01: Inadeauste Controls over
    a.   Scope (83750)
Locked Hiah Radiation Door Kevs
          The inspector selectively examined the external exposure control program. The
a.
          inspector reviewed records, discussed the program with cognizant personnel and
Scope (83750)
          observed exposure control practices during tours of the RCA and observation of
The inspector selectively examined the external exposure control program. The
          work activities. The inspector reviewed high radiation area controls and general
inspector reviewed records, discussed the program with cognizant personnel and
          radiological posting, implementation of the radiation work permit program, and
observed exposure control practices during tours of the RCA and observation of
          implementation of the dosimetry program.
work activities. The inspector reviewed high radiation area controls and general
    b.   Observations and Findinas
radiological posting, implementation of the radiation work permit program, and
          PECO continued to implement and maintain effective real time personnel exposure
implementation of the dosimetry program.
          control by use of an electronic dosimetry (ELD)/ access control system. The system
b.
          was set-up to preclude unauthorized individuals from signing onto invalid radiation
Observations and Findinas
          work permits (e.g., new permits or revised permits). Workers were observed to be
PECO continued to implement and maintain effective real time personnel exposure
          appropriately wearing dosimetry on their heads, as directed by radiation protection
control by use of an electronic dosimetry (ELD)/ access control system. The system
          personnel, when working in radiation dose rate gradients emanating from overhead.
was set-up to preclude unauthorized individuals from signing onto invalid radiation
          The inspector noted that areas (e.g., high radiation areas, radiation areas) were
work permits (e.g., new permits or revised permits). Workers were observed to be
          properly posted and locked (as appropriate). The inspector inventoried high
appropriately wearing dosimetry on their heads, as directed by radiation protection
          radiation area keys and noted all to be present and properly signed out if applicable.
personnel, when working in radiation dose rate gradients emanating from overhead.
          The inspector verified workers performing work activities in high radiation areas
The inspector noted that areas (e.g., high radiation areas, radiation areas) were
          were properly signed-in on applicable radiation work permits. The inspector noted
properly posted and locked (as appropriate). The inspector inventoried high
          that the licensee was providing neutron monitoring, in accordance with guidance in
radiation area keys and noted all to be present and properly signed out if applicable.
          NRC Regulatory Guide 8.14, of personnel working in neutron areas in preparation       ,
The inspector verified workers performing work activities in high radiation areas
          for the Unit 3 outage. The licensee used a calibrated neutron survey meter, source     i
were properly signed-in on applicable radiation work permits. The inspector noted
          checked for the range of use,                                                         j
that the licensee was providing neutron monitoring, in accordance with guidance in
          During a previous inspection, the inspector noted that the licensee established
NRC Regulatory Guide 8.14, of personnel working in neutron areas in preparation
          " 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
for the Unit 3 outage. The licensee used a calibrated neutron survey meter, source
          a statement (as work description) that other " approved work" was authorized. The
i
                                                                                                  l
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
26
    inspector questioned licensee personnel, including radiation protection control point
inspector questioned licensee personnel, including radiation protection control point
    personnel, as to what constituted " approved work." The licensee subsequently
personnel, as to what constituted " approved work." The licensee subsequently
    revised procedure HP-C-310 to include a definition as to what constituted
revised procedure HP-C-310 to include a definition as to what constituted
    " approved work." However, the statement indicated the following caveat:
" approved work." However, the statement indicated the following caveat:
    "For work which is not controlled by one of the work control documents specified
"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."
above, the verbal approval of a qualified RP technician is required."
    The inspector noted that this provision did not discuss obtaining applicable
The inspector noted that this provision did not discuss obtaining applicable
    supervisory approval (e.g., work group supervisor, radiation protection supervisor,
supervisory approval (e.g., work group supervisor, radiation protection supervisor,
    or operations supervisor). The licensee initiated a review of this matter.
or operations supervisor). The licensee initiated a review of this matter.
    The following observations were made:
The following observations were made:
    e      The licensee performed periodic calibration of each electronic dosimeter, but
The licensee performed periodic calibration of each electronic dosimeter, but
            there was no verification that the dose rate alarm would alarm at the pre-set
e
            dose rate. The integrated dose alarm feature was, however, tested. The
there was no verification that the dose rate alarm would alarm at the pre-set
            licensee relied, in part, on the dose rate alarm feature to alert workers to
dose rate. The integrated dose alarm feature was, however, tested. The
            changing conditions, but relied on the integrating dose alarm feature to
licensee relied, in part, on the dose rate alarm feature to alert workers to
            conform with Technical Specification high radiation area monitoring. The
changing conditions, but relied on the integrating dose alarm feature to
            licensee initiated a review of this matter.
conform with Technical Specification high radiation area monitoring. The
    *      The inspector observed workers building scaffolding in the Unit 3
licensee initiated a review of this matter.
            moisture separator area on May 5,1997. Workers were continuously
The inspector observed workers building scaffolding in the Unit 3
            monitored by use of teledosimetry systems. The inspector noted,
*
            however, that one worker's teledosimeter lost contact with the
moisture separator area on May 5,1997. Workers were continuously
            remote monitoring station. However, no action was taken by the
monitored by use of teledosimetry systems. The inspector noted,
            radiation protection technician at the remote monitoring location to
however, that one worker's teledosimeter lost contact with the
            attempt to understand the basis for the loss of contact (e.g., the
remote monitoring station. However, no action was taken by the
            worker exited the approved work location). Further, the worker later
radiation protection technician at the remote monitoring location to
            A;ted tir : area, walked past the technician, and despite the
attempt to understand the basis for the loss of contact (e.g., the
            teledosimetry still failing to make contact with the remote monitor,
worker exited the approved work location). Further, the worker later
            the technician did not challenge the worker. The inspector considered
A;ted tir : area, walked past the technician, and despite the
            this a weakness in oversight of activities. The licensee initiated a
teledosimetry still failing to make contact with the remote monitor,
            review of this matter.
the technician did not challenge the worker. The inspector considered
    *       Radiological signs / maps posted at the main radiological controlled area
this a weakness in oversight of activities. The licensee initiated a
            access were difficult to read. The licensee initiated a review of this matter.
review of this matter.
    The inspector reviewed the control of high radiation area access door keys. The
*
    following observations were made:
Radiological signs / maps posted at the main radiological controlled area
    e       The keys for the cabinets where Level 1 and Level 2 high radiation area door
access were difficult to read. The licensee initiated a review of this matter.
            keys are kept were not on the key inventory list. The licensee added the
The inspector reviewed the control of high radiation area access door keys. The
            keys to the key listing.
following observations were made:
    *       There was no guidance regarding timeliness of updating of the high radiation
e
            area key inventory list when a change to the list occurs.
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.


  ~
~
    .
.
                                              27
27
            On January 30,1997, the RPM became aware that master keys that could
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
be used to open locked high radiation area doors at the Limerick and Peach
            Bottom stations, were improperly controlled and in the possession of
Bottom stations, were improperly controlled and in the possession of
            unauthorized personnel between mid-1993 and November 1996. The keys
unauthorized personnel between mid-1993 and November 1996. The keys
            had been improperly made and distributed to fire protection personnel by the
had been improperly made and distributed to fire protection personnel by the
            licensee's corporate locksmith. The locksmith did not know the keys opened
licensee's corporate locksmith. The locksmith did not know the keys opened
            high radiation area doors, in addition, the licensee's radiation protection
high radiation area doors, in addition, the licensee's radiation protection
            manager was unaware of the existence of the master keys maintained by the
manager was unaware of the existence of the master keys maintained by the
            corporate locksmith. Consequently, the inspector concluded that the
corporate locksmith. Consequently, the inspector concluded that the
            licensee's administrative key control program for locked high radiation areas
licensee's administrative key control program for locked high radiation areas
            was not effeotive.
was not effeotive.
      The inspector noted that the keys possessed by the fire protection personnel were
The inspector noted that the keys possessed by the fire protection personnel were
      unauthorized and were not under the administrative control of radiation protection
unauthorized and were not under the administrative control of radiation protection
      personnel. As a result of the identification of the unauthorized keys, the licensee
personnel. As a result of the identification of the unauthorized keys, the licensee
      took the following actions:
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
The security access authorization was removed for the individuals known to
            received the keys.
possess the keys and other individuals within the work group who may have
      *     The locksmith who provided the master keys was relieved of his duties.
received the keys.
      *     The licensee initiated tours every 2 hours of locked high radiation area doors
*
            to review for unauthorized entries.
The locksmith who provided the master keys was relieved of his duties.
      *     The licensee initiated reviews for unplanned / unexplained radiation exposures
*
            for the affected areas using a combination of key card data and knowledge
The licensee initiated tours every 2 hours of locked high radiation area doors
            of the work areas of individuals known to possess the unauthorized keys.
to review for unauthorized entries.
            No unplanned or unusual exposures were noted.
*
      *     The licensee made a 1-hour report to the NRC on this rnatter.
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
,
,
      *      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
            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
            station and did not knowingly make the keys for unauthorized access
purposes at the station.
            purposes at the station.
*
      *     The licensee initiated an event report.
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
The licensee evaluated the occupational radiation exposures of the
            that the individuals did i.ot have unescorted access privileges to the
individuals determined to have copies of the keys. The licensee determined
            radiological controlled area, no unexplained or unplanned occupational
that the individuals did i.ot have unescorted access privileges to the
            exposures occurred, and no uncontrolled high radiation area access by those
radiological controlled area, no unexplained or unplanned occupational
            individuals occurred.
exposures occurred, and no uncontrolled high radiation area access by those
individuals occurred.


~
~
  .
.
                                                28
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
The licensee changed-out the lock cores of the affected doors (Level 1
                  10,000 Mr/hr) with special lock cores whose keys were only available to the
doors-greater than 1,000 Mr/hr and Level 2 doors- greater than
                  radiation protection group. These actions were completed by February 6,
10,000 Mr/hr) with special lock cores whose keys were only available to the
                  1997. (Note: Due to access control concerns, the licensee has not changed
radiation protection group. These actions were completed by February 6,
                  out the reactor water decant tank room (elevation 116' radwaste) lock or the
1997. (Note: Due to access control concerns, the licensee has not changed
                  Unit 3 subpile room door. These locks are to be changed when the areas are     !
out the reactor water decant tank room (elevation 116' radwaste) lock or the
                  available for entry. Access to those individual doors are controlled by other ,
Unit 3 subpile room door. These locks are to be changed when the areas are
                  high radiation area keys that are controlled.)                                 I
available for entry. Access to those individual doors are controlled by other
          e      The licensee placed the new keys (master keys and lock set changing keys)
,
                  under the administrative control of the radiation protection manager.
high radiation area keys that are controlled.)
          As discussed above, Technical Specification 5.7.2 requires that all door and gate
I
          keys to a high radiation area with dose rates greater than 1.0 rem /hr at
The licensee placed the new keys (master keys and lock set changing keys)
          30 centimeters from the radiation source be maintained under the administrative         l
e
          control of radiation protection personnel. The inspector noted that unauthorized
under the administrative control of the radiation protection manager.
          keys to locked high radiation areas, not under the administrative control of radiation
As discussed above, Technical Specification 5.7.2 requires that all door and gate
          protection personnel, were available to personnel (fire protection personnel) since
keys to a high radiation area with dose rates greater than 1.0 rem /hr at
          about 1993 through about November 1996. This is a violation of Technica!
30 centimeters from the radiation source be maintained under the administrative
          Specification 5.7.2. (VIO 50 277,278/97-04-03)
control of radiation protection personnel. The inspector noted that unauthorized
    c.   Conclusions
keys to locked high radiation areas, not under the administrative control of radiation
          PECO implemented a generally effective external exposure control program. A
protection personnel, were available to personnel (fire protection personnel) since
          violation was identified associated with failure to administratively control keys for
about 1993 through about November 1996. This is a violation of Technica!
          locked high radiation areas.
Specification 5.7.2. (VIO 50 277,278/97-04-03)
    R1.5 Control of Radioactive Materials and Contamination
c.
    a.   Scoce (83750)
Conclusions
          The inspector selectively reviewed radioactive material and contamination control
PECO implemented a generally effective external exposure control program. A
          practices including calibration and performance checks of survey and monitoring
violation was identified associated with failure to administratively control keys for
          instruments and the use of personal contamination monitors and friskers. The
locked high radiation areas.
          inspector also evaluated personnel skin contaminations and skin dose assessment
R1.5 Control of Radioactive Materials and Contamination
          methodology.
a.
    b.   Observations and Findinos
Scoce (83750)
          PECO implemented generally effective contamination control work techniques and
The inspector selectively reviewed radioactive material and contamination control
          prompt correction and cleanup of contamination. At the time of this inspection, the
practices including calibration and performance checks of survey and monitoring
          station exhibited approximately 2% of accessible floor areas as contamination areas
instruments and the use of personal contamination monitors and friskers. The
          (excluding the drywell). Contaminated areas exhibited generally low levels of
inspector also evaluated personnel skin contaminations and skin dose assessment
          contamination. Calibrated and response checked survey instrumentation was
methodology.
          available throughout the station. PECO tracked and trended personnel
b.
          contaminations for programmatic corrective action purposes. No personnel
Observations and Findinos
          contaminations resulted in any significant dose assessments.
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
29
    The licensee continued to implement room-specific control of those areas of the
i
    station which exhibited electron capture decay nuclides (e.g., Zinc-65) to provide
The licensee continued to implement room-specific control of those areas of the
    enhanced monitoring of material removed from these rooms and work performed
station which exhibited electron capture decay nuclides (e.g., Zinc-65) to provide
    therein.                                                                                               I
enhanced monitoring of material removed from these rooms and work performed
    A number of individuals were observed at the RCA control point with low level
therein.
    external contamination (principally clothing contamination) attributable to short-lived
A number of individuals were observed at the RCA control point with low level
    particulate daughter products of fission gasses. The short-lived daughter products                     l
external contamination (principally clothing contamination) attributable to short-lived
    (i.e., Cs-138 and Rb-88) adhered to personnel clothing, were readily detectable at
particulate daughter products of fission gasses. The short-lived daughter products
    the licensee radiological controlled egress points by whole body friskers, and did not
(i.e., Cs-138 and Rb-88) adhered to personnel clothing, were readily detectable at
    represent an internal exposure control concern.                                                       !
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
The contamination was attributable to low activity noble gas from the ventilation
    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
system located on the 195-foot elevation of the Unit 3 turbine building. The gas
    "F" moisture separato: area of the turbine building. A large steam leak at a flange                   i
was believed by the licensee to be generated as a result of fission of residual tramp
    in the area is reported to be the source of the gas. The licensce attempted to repair                 i
uranium remaining on incore surfaces. The ventilation system draws air from the
    the leak, but has not been successful and has decided to wait until the upcoming
"F" moisture separato: area of the turbine building. A large steam leak at a flange
    October 1997 Unit 3 refueling outage to repair the leak.
in the area is reported to be the source of the gas. The licensce attempted to repair
    The licensee plugged minor holes in the duct and indicated that ALARA cost benefit
i
    analyser did not indicate repair activities (e.g., during reactor downpowering) were
the leak, but has not been successful and has decided to wait until the upcoming
    cost beneficial.
October 1997 Unit 3 refueling outage to repair the leak.
    The inspector noted the contaminations did not result in any significant personnel
The licensee plugged minor holes in the duct and indicated that ALARA cost benefit
    exposures. The inspector expressed concern that the frequency of such
analyser did not indicate repair activities (e.g., during reactor downpowering) were
    contaminations and the need for personnel to remain inside the RCA could               -
cost beneficial.
    desensitize personnel to the need to continue to remain vigilant regarding personnel
The inspector noted the contaminations did not result in any significant personnel
    contamination monitoring. The licensee's radiation protection manager indicated
exposures. The inspector expressed concern that the frequency of such
    articles have been published in employee newspapers to alert personnel to the
contaminations and the need for personnel to remain inside the RCA could
    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
desensitize personnel to the need to continue to remain vigilant regarding personnel
    measurable by the licensee's TLDs.
contamination monitoring. The licensee's radiation protection manager indicated
    The licensee experienced an average of about 200 personnel contaminations for
articles have been published in employee newspapers to alert personnel to the
    1995 and 1996. The licensee had, as of the date of this inspection, sustained
matter and the locations of the low activity noble gas daughter products. Potential
    approximately 15 personnel contaminations for 1997, which, according to the
submersion doses, as indicated by the licensee's radiation protection manager, were
    licensee, was 50% less than the previous year. Contamination of personnel by low-
measurable by the licensee's TLDs.
    level noble gas daughter products was not included in this data and was not tracked
The licensee experienced an average of about 200 personnel contaminations for
    and trended.
1995 and 1996. The licensee had, as of the date of this inspection, sustained
    The inspector evaluated skin dose assessments previously performed by the
approximately 15 personnel contaminations for 1997, which, according to the
    licensee for two individuals selected by the inspector who sustained skin
licensee, was 50% less than the previous year. Contamination of personnel by low-
    contamination. The following was noted:
level noble gas daughter products was not included in this data and was not tracked
    *             The licensee used an industry code (VARSKIN MOD 1) to perform the skin
and trended.
                    dose calculations. The inspector evaluated the dose using a revised code
The inspector evaluated skin dose assessments previously performed by the
                    (VARSKIN MOD 2) which included gamma dose contributions. The
licensee for two individuals selected by the inspector who sustained skin
                    inspector's independent dose calculations indicated approximately a 30%
contamination. The following was noted:
                                                                                      .
*
    . . - . . . ,           , y       ,   .,.   -
The licensee used an industry code (VARSKIN MOD 1) to perform the skin
                                                          y y -__, ,                       , ., - - - -
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%
.
. . - . . . ,
,
y
,
m-.
.,.
-
y
y
-__,
,
,
., - - - -


_
_
                                                                                          i
i
  e
e
                                                                                          l
30
                                                                                          1
higher dose to the skin, assuming the contamination was a point source, as
                                          30                                             l
'
                                                                                          l
also assumed by the licensee. The inspector indicated that although the skin
          higher dose to the skin, assuming the contamination was a point source, as     '
contamination resulted in generally low skin dose (well within applicable NRC
          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
          limits) the licensee should evaluate its skin dose methodology particularly as
it relates to determination of gamma dose and use of latest computer codes.
          it relates to determination of gamma dose and use of latest computer codes.
The licensee subsequently obtained VARSKIN MOD 2, performed similar
          The licensee subsequently obtained VARSKIN MOD 2, performed similar
calculations and obtained essentially the same results. The licensee
          calculations and obtained essentially the same results. The licensee
subsequently developed a health physics job standard (HPJS - 9.8) to
          subsequently developed a health physics job standard (HPJS - 9.8) to
provide guidance for use of the updated code and indicated that previous
          provide guidance for use of the updated code and indicated that previous
skin dose assessments (for 1996 and 1997) would be reevaluated. The
          skin dose assessments (for 1996 and 1997) would be reevaluated. The
licensee indicated the workers' doses (discussed above) would be updated.
          licensee indicated the workers' doses (discussed above) would be updated.
On November 22,1996, the licensee became aware that a welder unit, released
    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
    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
    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
    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
    supervisor from Limerick Station to the vendor facility and found one contaminated
gasket, a pair of contaminated snips, and a contaminated screwdriver. The
    gasket, a pair of contaminated snips, and a contaminated screwdriver. The
maximum removable contamination found was 14,000 dpm (beta / gamma
    maximum removable contamination found was 14,000 dpm (beta / gamma
contamination). The HP supervisor surveyed personnel and did not identify any
    contamination). The HP supervisor surveyed personnel and did not identify any
personnel contamination. Also, the HP supervisor surveyed floors, tables, storage
    personnel contamination. Also, the HP supervisor surveyed floors, tables, storage
areas, tool boxes, waste cans and did not identify any contamination. The
    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.
    supervisor collected the bag of material, and transported it back to Peach Bottom.
Technical Specification 5.4.1 requires that written procedures be established,
    Technical Specification 5.4.1 requires that written procedures be established,
implemented and maintained covering the applicable procedures recommended in
    implemented and maintained covering the applicable procedures recommended in
Regulatory Guide 1.33, Appendix A, November 1972. The referenced appendix
    Regulatory Guide 1.33, Appendix A, November 1972. The referenced appendix
recommends in Section G that procedures for control of radioactivity and for limiting
    recommends in Section G that procedures for control of radioactivity and for limiting
materials released to the environment be established. Licensee radiation protection
    materials released to the environment be established. Licensee radiation protection
procedure HP-C-810, Revision 1, Radioactive Material (RAM) Control, specifies in
    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.,
    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
    less than 1000 disintegrations per minute per 100 square centimeters
(dpm/100cm2) removable beta gamma contamination and less than 5,000
    (dpm/100cm2) removable beta gamma contamination and less than 5,000
dpm/100cm2 total fixed and removable contamination.
    dpm/100cm2 total fixed and removable contamination.
The licensee determined that the welder unit had not been used since it was boxed
    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.
    up and shipped from Peach Bottom on September 27,1996.
As interim corrective actions, the license initiated an internal review, suspended all
    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
    release of large machinery (e.g., welder units) from the RCA except with
supervisory approval, surveyed other welder equipment on site, and instructed
    supervisory approval, surveyed other welder equipment on site, and instructed
station HP personnel regarding the event.
    station HP personnel regarding the event.
The licensee took the following additional actions:
    The licensee took the following additional actions:
e
    e     The licensee initiated an event report (PEP 1006341) for the matter and
The licensee initiated an event report (PEP 1006341) for the matter and
            initiated an investigation.
initiated an investigation.
                                                                                          I
I


n     . - - . . -         -     . - - . - - -         . - . - . - - . - - - _ . _             - -   - - - .
n
  O
. - - . . -
                                                                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
O
                            upon survey.
31
                                                                                                                  !
e
                  e        The licensee reviewed the adequacy of the general employee training
The licensee evaluated the reportability of the event and determined it was
                            program relative to the event for potential enhancements and did not identify
not reportable.
                            any weaknesses.
e
                  e        The licensee took action to revise procedures for release of large objects as
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
'
'
                            expected during outages. The licensee was expected to complete the
revision by June 30,1997.
                            revision by June 30,1997.
The licensee concluded that the non-surveyed welder unit, with the slightly
                                                                                                                  1
!
                  The licensee concluded that the non-surveyed welder unit, with the slightly                   !
contaminated tools, was inadvertently released from the reactor building Unit 2
                  contaminated tools, was inadvertently released from the reactor building Unit 2
l
l                 railroad door when it was placed next to a similar welder unit that had been
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
surveyed at the same location. The door was used following the outage to release
                  outage-related equipment and has since been closed.                                           !
i
                  The inspector noted that failure to implement procedures recommended in
outage-related equipment and has since been closed.
                  Appendix A of Regulatory Guide 1.33,1972 is an apparent violation. The inspector
The inspector noted that failure to implement procedures recommended in
                  reviewed this violation with respect to the criteria for exercise of discretion outlined
Appendix A of Regulatory Guide 1.33,1972 is an apparent violation. The inspector
l                 in Section Vll.B.1 of the " General Staternent of Policy and Procedure for NRC
reviewed this violation with respect to the criteria for exercise of discretion outlined
                  Enforcement Actions," (60 FR 34381; June 30,1995). The inspector noted that
l
                  even though the above issue was identified by the licensee, it did not appear to be
in Section Vll.B.1 of the " General Staternent of Policy and Procedure for NRC
                  an issue that could have been prevented by a previous violation, it did not appear to
Enforcement Actions," (60 FR 34381; June 30,1995). The inspector noted that
                  be willful, and corrective actions were taken as discussed above. The inspector
even though the above issue was identified by the licensee, it did not appear to be
                  concluded the above matters constituted a licensee-identified and corrected
an issue that could have been prevented by a previous violation, it did not appear to
                  violation, which is considered non-cited, consistent with Section Vll.B.1 of the NRC
be willful, and corrective actions were taken as discussed above. The inspector
                  Enforcement Policy.
concluded the above matters constituted a licensee-identified and corrected
            c.   Conclusions
violation, which is considered non-cited, consistent with Section Vll.B.1 of the NRC
                  PECO implemented a generally effective contamination control program. One non-
Enforcement Policy.
                  cited violation was identified regarding the release of a contaminated welding
c.
                  machine.
Conclusions
PECO implemented a generally effective contamination control program. One non-
cited violation was identified regarding the release of a contaminated welding
machine.
1
1
I
I
o
o
  -- ,               ---c     ,                                 .                     -_-                   - -
--
,
---c
,
.
-_-
- -


~
~
  e
e
                                                  32
32
    R5   Staff Training and Qualification in Radiation Protection and Chemistry
R5
    R5.1 Radiation Workers /Radioloaical Controls Personnel
Staff Training and Qualification in Radiation Protection and Chemistry
      a. Scope (83750)
R5.1 Radiation Workers /Radioloaical Controls Personnel
          The inspector reviewed the training and qualification records of a worker who was     ;
a.
          fit-tested for use of respiratory protection equipment, the qualifications of the     !
Scope (83750)
          radiation protection technician who fit-tested the worker, and reviewed the training
The inspector reviewed the training and qualification records of a worker who was
          documentation and completion of required surveys by selected advanced radiation       l
fit-tested for use of respiratory protection equipment, the qualifications of the
          workers. The inspector also reviewed the training provided radiation protection
radiation protection technician who fit-tested the worker, and reviewed the training
          technicians. The inspector evaluated the training and qualification of these
documentation and completion of required surveys by selected advanced radiation
          individuals relative to applicable technical specification requirements, procedural
workers. The inspector also reviewed the training provided radiation protection
          requirements, and 10 CFR 50.120. The inspector reviewed training records and
technicians. The inspector evaluated the training and qualification of these
          discussed qualification criteria with cognizant personnel.
individuals relative to applicable technical specification requirements, procedural
    - b. Observations and Findinas
requirements, and 10 CFR 50.120. The inspector reviewed training records and
          PECO provided training and qualification, as appropriate, for the individuals selected
discussed qualification criteria with cognizant personnel.
          by the inspector. The licensee established and implemented a health physics
- b.
          technician continuing training course plan. The licensee provided 102 hours of
Observations and Findinas
          training per technician for 1995,92 hours per technician for 1996, and developed a     j
PECO provided training and qualification, as appropriate, for the individuals selected
          course plan for 1997. The 1997 plan includes industry events, design basis             I
by the inspector. The licensee established and implemented a health physics
          analysis and impact of hydrogen water chemistry.
technician continuing training course plan. The licensee provided 102 hours of
      c. Conclusions
training per technician for 1995,92 hours per technician for 1996, and developed a
          PECO provided training of radiological controls personnel providing respirator fit ~
j
          testing, workers identified to wear respirators, and training of advanced radiation
course plan for 1997. The 1997 plan includes industry events, design basis
          workers. PECO was also providing continuing training to radiation protection
analysis and impact of hydrogen water chemistry.
          technicians.
c.
    R7   Quality Assurance in Radiological Protection and Chemistry Activities (83750)
Conclusions
    R7.1 Radioloaical Event Reoorts
PECO provided training of radiological controls personnel providing respirator fit ~
      a. Scope (83750)
testing, workers identified to wear respirators, and training of advanced radiation
          The inspector selectively reviewed oversight activities for radiological controls. In
workers. PECO was also providing continuing training to radiation protection
          particular, the inspector reviewed PECO's evaluations and actions associated with
technicians.
          self-identified issues and concerns documented in its self-identification programs
R7
          (e.g., personnel contamination reports, radiological occurrence reports, performance
Quality Assurance in Radiological Protection and Chemistry Activities (83750)
          enhancement issues, quality assurance surveillance items, and industry audits).
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
33
    b.   Observations and Findinas
b.
          The inspector reviewed selected licensee self-identified issues covering calendar
Observations and Findinas
          year 1996 and 1997 up to the time of the inspection. The inspector's review           l
The inspector reviewed selected licensee self-identified issues covering calendar
          indicated that the licensee took effective and timely action on self-identified
year 1996 and 1997 up to the time of the inspection. The inspector's review
          concerns. The inspector noted generally good oversight of activities.
indicated that the licensee took effective and timely action on self-identified
          The inspector reviewed various audits and surveillances and noted the use of
concerns. The inspector noted generally good oversight of activities.
          applicable industry standards as audit criteria. The licensee developed a program
The inspector reviewed various audits and surveillances and noted the use of
          (procedure HP-C-109) to periodically review the radiation protection program
applicable industry standards as audit criteria. The licensee developed a program
          content and implementation as outlined in 10 CFR 20.1101. Technical experts
(procedure HP-C-109) to periodically review the radiation protection program
          were used to support the reviews. The licensee evaluated industry experience for
content and implementation as outlined in 10 CFR 20.1101. Technical experts
          potential programmatic enhancements and developed a quarterly trend report using
were used to support the reviews. The licensee evaluated industry experience for
          numerous inputs on radiation protection performance for use in assessing the
potential programmatic enhancements and developed a quarterly trend report using
          effectiveness of the radiation protection program.
numerous inputs on radiation protection performance for use in assessing the
    c.   Conclusions
effectiveness of the radiation protection program.
          PECO implemented an effective program for self-identifying and correcting self-
c.
          identified issues and concerns. No violations or safety concerns were identified.
Conclusions
    R8   Miscellaneous RP&C Activities
PECO implemented an effective program for self-identifying and correcting self-
    R8.1 (Closed) Unresolved item 96-06-04: Review of Radioactive Material Storaae
identified issues and concerns. No violations or safety concerns were identified.
          Locations Versus Uodated Final Safety Analysis (UFSAR) Descriptions
R8
    a.   Scoce (83750)
Miscellaneous RP&C Activities
          During NRC Combined Inspection No. 50-277;278/95-27, (conducted
R8.1 (Closed) Unresolved item 96-06-04: Review of Radioactive Material Storaae
          November 26,1995, through January 13,1996) and 50-277;278/96-06,
Locations Versus Uodated Final Safety Analysis (UFSAR) Descriptions
          (conducted July 7,1996, through September 7,1996) the inspector reviewed the
a.
          conformance of the licensee's radioactive waste storage and processing facilities
Scoce (83750)
          relative to descriptions within the UFSAR. Insufficient review of this issue resulted
During NRC Combined Inspection No. 50-277;278/95-27, (conducted
          in an unresolved item.
November 26,1995, through January 13,1996) and 50-277;278/96-06,
    b.   Observations and Findinas
(conducted July 7,1996, through September 7,1996) the inspector reviewed the
          During the inspection, the inspector met with cognizant licensee personnel and
conformance of the licensee's radioactive waste storage and processing facilities
          discussed the actions taken on the earlier identified discrepancies as described
relative to descriptions within the UFSAR. Insufficient review of this issue resulted
          below,
in an unresolved item.
          e     The inspector noted that, relative to the liquid radioactive waste system, the
b.
                  licensee identified lack of neutralization of the chemical waste tank contents
Observations and Findinas
                  prior to transfer to radwaste floor drain sumps as indicated in UFSAR
During the inspection, the inspector met with cognizant licensee personnel and
                  Section 9.2.4.2.3. The licensee performed a safety evaluation (June 12,
discussed the actions taken on the earlier identified discrepancies as described
                  1996) for the current mode of operation, did not identify any safety concerns
below,
                  and initiated an engineering change request (May 3,1996) to update the
e
                  UFSAR.
The inspector noted that, relative to the liquid radioactive waste system, the
                                                                                                l
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
UFSAR.


                                                        m.___             ___ __._ _
m.___
___ __._ _
e
e
                                            34
34
    *       The UFSAR did not contain any apparent specific information relative to
*
            outdoor storage of radioactive materials / radioactive waste. However, a
The UFSAR did not contain any apparent specific information relative to
            licensee 10 CFR 50.59 evaluation for outdoor storage, identified several
outdoor storage of radioactive materials / radioactive waste. However, a
            outdoor radioactive material storage / staging areas, some of which were not
licensee 10 CFR 50.59 evaluation for outdoor storage, identified several
            used. The 10 CFR 50.59 evaluation, performed for outdoor _ storage of           ;
outdoor radioactive material storage / staging areas, some of which were not
            radioactive material, did not address the storage of sea van trailers behind   I
used. The 10 CFR 50.59 evaluation, performed for outdoor _ storage of
            the 135' elevation of the radioactive waste building. The inspector noted     j
radioactive material, did not address the storage of sea van trailers behind
            that the trailer storage appeared to be well within restrictions on radiation
the 135' elevation of the radioactive waste building. The inspector noted
            - dose rates presented in the 10 CFR 50.59 evaluation for other storage
j
            locations. The licensee updated the 10 CFR 50.59 (May 5,1997) to identify
that the trailer storage appeared to be well within restrictions on radiation
            specific storage locations, removed unnecessary sea vans, and initiated an
- dose rates presented in the 10 CFR 50.59 evaluation for other storage
            engineering change request to update the UFSAR to reflect outdoor storage.
locations. The licensee updated the 10 CFR 50.59 (May 5,1997) to identify
    The inspector's review did not identify any apparent significant safety concerns
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.       l
,
    The licensee indicated a UFSAR update would be submitted on or about
associated with the findings. The licensee took actions (e.g., Action Requests) to
    June 30,1997, to reflect the changes. The licensee took generic actions to train
, l
    appropriate personnel that station changes were to be processed through the
review the findings and update the UFSAR and applicable drawings, as appropriate.
    10 CFR 50.59 process.                                                                   j
The licensee indicated a UFSAR update would be submitted on or about
    The inspector noted that 10 CFR 50.59(a) states that the licensee may make
June 30,1997, to reflect the changes. The licensee took generic actions to train
    changes to the facility as described in the safety evaluation report without prior
appropriate personnel that station changes were to be processed through the
    commission approval provided that the change does not involve a change to the
10 CFR 50.59 process.
    technical specifications or an unreviewed safety question.10 CFR 50.9(b) requires     j
j
    that records of the changes include a safety evaluation, which provides the basis
The inspector noted that 10 CFR 50.59(a) states that the licensee may make
    that the change did not involve an unreviewed safety question. Further,10 CFR
changes to the facility as described in the safety evaluation report without prior
    50.71(e)(4) requires that the UFSAR be updated to reflect the changes. The
commission approval provided that the change does not involve a change to the
    inspector's review indicated 10 CFR 50.59 evaluations were not made for the
technical specifications or an unreviewed safety question.10 CFR 50.9(b) requires
    changes discussed above. The licensee subsequently took corrective actions as
j
    described above. The inspector's review of the changes did not identify any
that records of the changes include a safety evaluation, which provides the basis
    significant safety concerns,                                                           i
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
    Based on the above, the inspector noted that the failure to perform a 10 CFR 50.59     i
inspector's review indicated 10 CFR 50.59 evaluations were not made for the
    evaluations constitute violations of no safety consequence, and are being treated as
changes discussed above. The licensee subsequently took corrective actions as
    non-cited violations, consistent with Sections IV and Vll.B.1 of the NRC               ;
described above. The inspector's review of the changes did not identify any
    Enforcement Pohcy.                                                                     !
significant safety concerns,
  c. Conclusions                                                                             j
i
    The licensee initiated action to update the UFSAR to reflect current practices for
Based on the above, the inspector noted that the failure to perform a 10 CFR 50.59
    storage and staging of low level radioactive / contaminated materialin the yard areas
i
    of the station and operation of radioactive waste systems. Unresolved item
evaluations constitute violations of no safety consequence, and are being treated as
    (URI 50-277; 50-278/96-06-04) associated with UFSAR discrepancies is closed. A
non-cited violations, consistent with Sections IV and Vll.B.1 of the NRC
    non-cited violation of 10 CFR 50.59 was identified.
Enforcement Pohcy.
                                                                                            i
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.


,           - _ _ _ .       .           .         -           . _ - __ ..                     .     . _ - -
,
  A
- _ _ _ .
  e
.
                                                      35
.
    R8.2 Housekeepina                                                                                       ;
-
                                                                                                            l
. _ -
              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
A
              rooms).
e
    R8.3 Verification of Updated Final Safety Analysis Commitmen_ts
35
      a.     Scope (83750)
R8.2 Housekeepina
              A recent discovery of a licensee operating their facility in a manner contrary to the
The inspector toured the facility and noted overall very good plant conditions
              UFSAR description highlighted the need for a special, focused review that compares
including areas outside the station. Areas were generally neat and no leaking
              plant practices, procedures and/or parameters to the UFSAR description. While
equipment was noted. The licensee took action to clean and paint, as appropriate,
              performing the inspections discussed in this report, the inspectors reviewed the
areas previously noted by the inspector to exhibit poor conditions (e.g., waste
              applicable portions of the UFSAR that related to the areas inspected.
collector / floor drain collector tank room, condensate backwash receiving tank
      b.     Observations and Findinas
rooms).
            The inspector reviewed turbine building ventilation systems associated with
R8.3 Verification of Updated Final Safety Analysis Commitmen_ts
              personnel contamination by noble gas daughter products. The inspector discussed
a.
              the ventilation system design bases and any changes made to the ventilation
Scope (83750)
              system relative to UFSAR descriptions.
A recent discovery of a licensee operating their facility in a manner contrary to the
      c.     Conclusions
UFSAR description highlighted the need for a special, focused review that compares
              No inconsistencies were identified.
plant practices, procedures and/or parameters to the UFSAR description. While
                                          V. Manaaement Meetinas
performing the inspections discussed in this report, the inspectors reviewed the
    X1       Exit Meeting Summary
applicable portions of the UFSAR that related to the areas inspected.
    An exit meeting was conducted on April 25,1997, at which the results of the inspection
b.
    were presented. PECO representatives acknowledged, and did not contest, the findings at
Observations and Findinas
    that time.
The inspector reviewed turbine building ventilation systems associated with
    X2       Review of UFSAR Commitments
personnel contamination by noble gas daughter products. The inspector discussed
    A recent discovery of a licensee operating their facility in a manner contrary to the Updated
the ventilation system design bases and any changes made to the ventilation
    Final Safety Analysis Report (UFSAR) description highlighted the need for a special focused
system relative to UFSAR descriptions.
    review that compares plant practices, procedures and/or parameters to the UFSAR
c.
    description. While performing the inspections discussed in this report, the inspector
Conclusions
    reviewed the application portions of the UFSAR that related to the areas inspected. The
No inconsistencies were identified.
    inspector verified that the UFSAR wording was consistent with the observed plant
V. Manaaement Meetinas
    practices, procedure and/or parameters.
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.


                                                                                          ._
._
                                                                                            l
                                                                                            1
'
'
                              INSPECTION PROCEDURES USED
INSPECTION PROCEDURES USED
  IP 37551: Onsite Engineering Observations
IP 37551: Onsite Engineering Observations
  IP 40500: Effectiveness of Licensee Controls in identifying, Resolving,and Preventing
IP 40500: Effectiveness of Licensee Controls in identifying, Resolving,and Preventing
                Problems
Problems
  IP 61726: Surveillance Observations
IP 61726: Surveillance Observations
  IP 62707: Maintenance Observation
IP 62707: Maintenance Observation
  IP 64704: Fire Protection Program
IP 64704: Fire Protection Program
  IP 71707: Plant Operations
IP 71707: Plant Operations
  IP 71750: Plant Support Observations
IP 71750: Plant Support Observations
  IP 83750: Occupational Exposure
IP 83750: Occupational Exposure
  IP 92700: Onsite Follow of Written Reports of Nonroutine Events at Power Reactor
IP 92700: Onsite Follow of Written Reports of Nonroutine Events at Power Reactor
                Facilities
Facilities
  IP 92901: Operations Followup
IP 92901: Operations Followup
  IP 92902: Followup - Engineer
IP 92902: Followup - Engineer
  IP 92903: Followup - Maintenance
IP 92903: Followup - Maintenance
  IP 92904: Plant Support Followup
IP 92904: Plant Support Followup
  iP 93702: Prompt Onsite Response to Events at Operating Power Reactors
iP 93702: Prompt Onsite Response to Events at Operating Power Reactors
                          ITEMS OPENED, CLOSED, AND DISCUSSED
ITEMS OPENED, CLOSED, AND DISCUSSED
                                                                                            i
i
  Ooened                                                                                     !
Ooened
                                                                                            !
IFl 50-278/97-04-01
  IFl 50-278/97-04-01               Review Maintenance Rule Program Application to 13 kV
Review Maintenance Rule Program Application to 13 kV
                                    Breaker Switches
Breaker Switches
  IFl 50-277; 50/278/97-04-02       Review Revisions to Fire Protection Training Lesson
IFl 50-277; 50/278/97-04-02
                                    Plans
Review Revisions to Fire Protection Training Lesson
  Closed
Plans
  VIO 50-277; 50-278/97-04-03       Violation of locked high radiation area key control
Closed
  URI 50-277; 50-278/96-06-04       Updating of the UFSAR in accordance with 10 CFR
VIO 50-277; 50-278/97-04-03
                                    50.71 (e).                                             '
Violation of locked high radiation area key control
  LER 2-97-001                       Flow Biased Scram Setpoints
URI 50-277; 50-278/96-06-04
  LER 2-97-002                       Recirc Pump Motor PF issues
Updating of the UFSAR in accordance with 10 CFR
  LER 3-97-002                       Reactor Scram due to Natural Circulation
50.71 (e).
                                                                                            l
'
                                                                                            l
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


^
^
                                                            l
l
  *
l
                                                            l
*
  e
e
                                      LIST OF ACRONYMS USED
LIST OF ACRONYMS USED
    action request (AR)
action request (AR)
    action statement (AS)
action statement (AS)
    administrative guideline (AG)                           l
administrative guideline (AG)
    APRM gain adjust factor (AGAF)
APRM gain adjust factor (AGAF)
    as-low-as-reasonably-achievable (ALARA)
as-low-as-reasonably-achievable (ALARA)
    average power range monitors - neutron (APRMs)
average power range monitors - neutron (APRMs)
    control rod drives (CRDs)
control rod drives (CRDs)
    control room deficiency list NRDL)                     l
control room deficiency list NRDL)
                                                            '
'
    control room emergency ventilation (CREV)
control room emergency ventilation (CREV)
    core power and flow log (CPFL)
core power and flow log (CPFL)
    core spray (CS)
core spray (CS)
    core thermal power (CTP)
core thermal power (CTP)
    design input document (DID)
design input document (DID)
    diaphragm alternative response test (DART)
diaphragm alternative response test (DART)
    disintegrations per minute (DPM)                       ;
disintegrations per minute (DPM)
    electro-hydraulic control (EHC)                         i
electro-hydraulic control (EHC)
    eleventh refueling outage (2R11)
i
    emergency core cooling system (ECCS)                   l
eleventh refueling outage (2R11)
    emergency diesel generators (EDG)
emergency core cooling system (ECCS)
                                                            '
emergency diesel generators (EDG)
    emergency preparedness (EP)
'
    emergency service water (ESW)
emergency preparedness (EP)
    end-of-cycle (EOC)
emergency service water (ESW)
    engineering change request (ECR)
end-of-cycle (EOC)
    engineered safety feature (ESF)
engineering change request (ECR)
    equipment study list (ESL)
engineered safety feature (ESF)
    functional testing (FT)
equipment study list (ESL)
    general procedure (GP)
functional testing (FT)
    Generic Letter (GL)
general procedure (GP)
    health physics (HP)
Generic Letter (GL)
    high pressure coolant injection (HPCI)
health physics (HP)
    high pressure service water (HPSW)
high pressure coolant injection (HPCI)
    hydraulic control unit (HCU)
high pressure service water (HPSW)
    improved TS (ITS)
hydraulic control unit (HCU)
    independent safety engineering group (ISEG)
improved TS (ITS)
    inservice inspection (ISI)
independent safety engineering group (ISEG)
    inspector followup items (IFis)
inservice inspection (ISI)
    instrument and control (l&C)
inspector followup items (IFis)
    intermediate range monitor - neutron (IRM)
instrument and control (l&C)
    licensee event report (LER)
intermediate range monitor - neutron (IRM)
    limited senior reactor operators (LSROs)
licensee event report (LER)
    limiting conditions for operation (LCO)
limited senior reactor operators (LSROs)
    load tap changer (LTC)
limiting conditions for operation (LCO)
    local leak rate test (LLRT)
load tap changer (LTC)
    loss of coolant accident (LOCA)
local leak rate test (LLRT)
    loss of off-site power (LOOP)
loss of coolant accident (LOCA)
    low pressure coolant injection (LPCI)
loss of off-site power (LOOP)
    lubricating oil (LO)
low pressure coolant injection (LPCI)
    modification (MOD)
lubricating oil (LO)
    motor generator (MG)
modification (MOD)
motor generator (MG)


  ~
~
    4
4
    i
i
                                                    2
2
        nuclear maintenance division (NMD)
nuclear maintenance division (NMD)
        nuclear review board (NRB)
nuclear review board (NRB)
        offsite dose calculation manual (ODCM)
offsite dose calculation manual (ODCM)
        offsite power start-up source #2 (2SU)
offsite power start-up source #2 (2SU)
;      offsite power start-up source #3 (3SU)
offsite power start-up source #3 (3SU)
!       Peco Energy (PECO)
;
        performance enhancement program (PEP)
!
        plant equipment operator (PEO)
Peco Energy (PECO)
        plant operations review committee (PORC)
performance enhancement program (PEP)
        post-maintenance testing (PMT)
plant equipment operator (PEO)
i       primary containment (PC)
plant operations review committee (PORC)
        primary containment isolation system (PCIS)
post-maintenance testing (PMT)
i
primary containment (PC)
primary containment isolation system (PCIS)
#
#
        primary containment isolation valve (PCIV)
primary containment isolation valve (PCIV)
protected area (PA)
4
4
        protected area (PA)
quality assurance (QA)
        quality assurance (QA)
radiation protection manager (RPM)
        radiation protection manager (RPM)
radiologically controlled area (RCA)
'
rated thermal power (RTP)
.
reactor core isolation cooling (RCIC)
!
reactor engineer (RE)
'
'
        radiologically controlled area (RCA)
reactor feed pump (RFP)
        rated thermal power (RTP)
reactor operator (RO)
.
        reactor core isolation cooling (RCIC)
!      reactor engineer (RE)
'
        reactor feed pump (RFP)
.,
.,
        reactor operator (RO)
reactor protection system (RPS)
        reactor protection system (RPS)
I
I       reliability centered maintenance (ROM)
reliability centered maintenance (ROM)
,
,
'
residual heat removal (RHR)
        residual heat removal (RHR)
'
        residual heat removal (RHR)
residual heat removal (RHR)
        safety evaluation report (SER)
safety evaluation report (SER)
;       safety related structures, system and components (SSC)
;
2
safety related structures, system and components (SSC)
        safety relief valve (SRV)
safety relief valve (SRV)
2
scram solenoid pilot valve (SSPV)
-
-
        scram solenoid pilot valve (SSPV)
secondary containment (SC)
        secondary containment (SC)
;
;       senior reactor operator (SRO)
senior reactor operator (SRO)
:       shift technical advisor (STA)
:
        shift update notice (SUN)
shift technical advisor (STA)
        source range monitor (SRM)
shift update notice (SUN)
        specific gravity (SG)
source range monitor (SRM)
        spent fuel pool (SFP)                                 i
specific gravity (SG)
        standby gas treatment (SGTS)
i
        standby liquid control (SLC)
spent fuel pool (SFP)
        station blackout (SBO)
standby gas treatment (SGTS)
        structure, system and component (SSC)
standby liquid control (SLC)
        surveillance requirement (SR)
station blackout (SBO)
        surveillance test (ST)
structure, system and component (SSC)
        systerns approach to training (SAT)
surveillance requirement (SR)
        technical requirements manual (TRM)
surveillance test (ST)
        technical specification (TS)
systerns approach to training (SAT)
      - temporary plant alteration (TPA)
technical requirements manual (TRM)
        turbine bypass valve (BPV)
technical specification (TS)
- temporary plant alteration (TPA)
turbine bypass valve (BPV)


  7
7
  2
2
    V-
V-
                                                        3-
3-
          turbine control valve (TCV)
turbine control valve (TCV)
        ' turbine stop valve (TSV)
' turbine stop valve (TSV)
      e  undervoltage (UV)
undervoltage (UV)
          unresolved item (URI)
e
          updated final safety analysis report (UFSAR)
unresolved item (URI)
                                                              1
updated final safety analysis report (UFSAR)
                                                          ~l
~l
                                                            l
)
                                                            )
1
                                                            !
I
                                                            1
i
                                                            I
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l
                                                          '1
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;
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                                                            ;
                                                            I
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.
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Latest revision as of 16:19, 24 May 2025

Insp Repts 50-277/97-04 & 50-278/97-04 on 970504-0607. Violations Noted.Major Areas Inspected:Operations, Surveillance & Maintenance,Engineering & Technical Support & Plant Support
ML20149L295
Person / Time
Site: Peach Bottom  Constellation icon.png
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)
v • d • e


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

DPR-56

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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PDR

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

................................................. 1

01

Conduct of Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

04

Operator Knowledge and Performance . . . . . . . . . . . . . . . . . . . . . . . . . 1

04.1 High Pressure Coolant injection inoperable Due to Lifted Lead -

Unit 2...........................................

1

07

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..........................................

3

II.

Maintenance................................................

4

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

6

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M1.3 E-1 Emergency Diesel Generator Maintenance

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M1.4 Surveillance Activities

8

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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

13

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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

25

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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

01

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.

04

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.

Scram Air Header Leaks

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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

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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

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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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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

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Standby Liquid Control System Surveillance Test - Unit 2

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a.

Scoce (61726)

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The inspector observed the performance of surveillance test ST-O-011301-2,

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" Standby Liquid Control Pump Functional Test for IST," on May 27,1997.

)

b.

Observations and Findinos

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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

,

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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

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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)

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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).

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12

ITS CFT definition:

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

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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

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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

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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

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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:

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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

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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

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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

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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

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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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F3

Fire Protection Procedures and Documentation

F3,1

Fire Protection Procedure Reviews

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The inspector reviewed the procedures controlling fire protection activities at the

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facility to determine what management controls had been developed to prevent fires

and rapidly suppress any fire which might occur. The following specific procedures

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were reviewed:

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Nuclear Generation Group Policy No. NP-FP-1, Rev. O, Fire Protection

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FF-01, Rev. 3, Fire Brigade

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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

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7t page, plant phone, or portable radio, is available in the immediate vicinity of

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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

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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

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regarding the effect on fire protection systems or features, IRM provides resolution.

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AG-CG-012.02, Control of Combustible and Flammable Materials

1

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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,

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controls on smoking in the faci!ity, and disposal guidelines for combustible

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materials.

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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.

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FF-01, Fire Brigade

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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

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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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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).

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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

l

actions. The review included the following audits:

!

Assessment Report No. A0808620, " Assessment of Fire Protection Program,

e

j

MAP Area D02," dated December 15,1994.

e

Assessment Report No. A0900835, "PBAPS Triennial Fire Protection

!

Program, MAP Area D02, Rev. 8," dated September 25,1995.

4

Assessment Report No. A0967111, "PBAPS Annual Fire Protection Program

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Assessment, MAP Area D02, Rev. 9," dated July 17,1996.

i

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.PEPlssue 10005802, " Deterioration of CO, Extinguisher Hoses," initiated

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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

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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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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

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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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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

,

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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

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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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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.

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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

locked high radiation areas.

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.

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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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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

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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

Enforcement Policy.

c.

Conclusions

PECO implemented a generally effective contamination control program. One non-

cited violation was identified regarding the release of a contaminated welding

machine.

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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

UFSAR.

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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

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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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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

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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).

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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)

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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)

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eleventh refueling outage (2R11)

emergency core cooling system (ECCS)

emergency diesel generators (EDG)

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emergency preparedness (EP)

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)

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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)

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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)

,

residual heat removal (RHR)

'

residual heat removal (RHR)

safety evaluation report (SER)

safety related structures, system and components (SSC)

safety relief valve (SRV)

2

scram solenoid pilot valve (SSPV)

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secondary containment (SC)

senior reactor operator (SRO)

shift technical advisor (STA)

shift update notice (SUN)

source range monitor (SRM)

specific gravity (SG)

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spent fuel pool (SFP)

standby gas treatment (SGTS)

standby liquid control (SLC)

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)

turbine bypass valve (BPV)

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2

V-

3-

turbine control valve (TCV)

' turbine stop valve (TSV)

undervoltage (UV)

e

unresolved item (URI)

updated final safety analysis report (UFSAR)

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