ML16343A312
| ML16343A312 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 06/20/1995 |
| From: | Pellet J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML16342C960 | List: |
| References | |
| 50-275-95-04, 50-275-95-4, 50-323-95-04, 50-323-95-4, NUDOCS 9506280649 | |
| Download: ML16343A312 (20) | |
See also: IR 05000275/1995004
Text
ENCLOSURE 2
U.S.
NUCLEAR REGULATORY COMMISSION
REGION IV
Inspection
Report:
50-275/95-04
50-323/95-04
Licenses:
DPR-82
Licensee:
Pacific
Gas
and Electric Company
77 Beale Street,
Room
1451
P.O.
Box 770000
San Francisco,
Facility Name:
Diablo Canyon Nuclear
Power Plant, Units
1 and
2
Inspection At:
Diablo Canyon Nuclear
Power Plant,
Units
1 and
2
Inspection
Conducted:
May 22-25,
1995
Inspectors:
Joseph I. Tapia,
Examiner/Inspector,
Operations
Branch,
Division of Reactor Safety
Stephen
L. McCrory, Examiner/Inspector,
Operations
Branch,
Division of Reactor Safety
Approved:
o
.
e
et,
ie
,
perations
rane
Division of Reactor Safety
ate
Ins ection
Summar
Areas
Ins ected
Units
1
and>> 2
Routine,
announced
inspection of the
qualifications of applicants for operating licenses,
the licensed operator
requalification program,
and ability to detect
and respond to steam generator
tube leaks
and tube ruptures.
Results
Units
1 and
2
~0eratioas
Both applicants for initial licenses
passed
the examinations
(Section 1).
The reference
material
provided
by the training department for
examination
development
was
good
and adequately
supported
examination
development
(Section
1. 1).
9506280649
95062i
ADQCK 05000275
8
The examiners
observed generally
good communication practices
by
applicants
during the conduct of the examination
(Section 1.2).
Operator
performance
during the requalification examination
was
satisfactory with some weaknesses
in communications
and
command
and
control.
Examples of nondirected
communications,
informal wording,
multiple conversations
and failures to repeat
back were noted
(Section 2.2).
~
The licensee's
handling of generic
communications
related to steam
generator
tube integrity was considered
appropriate,
however,
one
example of untimely implementation of generic
communications
was noted
(Section 3.1).
~
The licensee
exhibited
good capability and used diverse
methods for
detection of primary-to-secondary
leakage.
Appropriate alarm setpoints
were utilized to provide rapid notification of increasing
leakage
(Section 3.2
and 3.3).
.
The procedures
and supporting training that were
used
by operators
in
response
to primary-to-secondary
leakage
were good (Section 3.4).
Summar
of Ins ection Findin s:
~
The practice of allowing licensed reactor operators
to fill the position
of senior control operator
was not consistent with the commitment
made
by the licensee
in the
FSAR and resulted
in
a Deviation (Section 2.2).
~
One deviation
was identified during this inspection
(275;323/9504-01).
Attachments:
~
Attachment
1 - Persons
Contacted
and Exit Heeting
~
Attachment
2
Simulation Facility Report
- ~
-3-
DETAILS
1
LICENSED OPERATOR APPLICANT INITIALQUALIFICATION EVALUATION (NUREG-1021)
During the inspection,
the examiners
evaluated
the qualifications of one
license applicant for senior reactor operator
and
one applicant for reactor
operator.
The inspection
assessed
the eligibility and administrative
and
technical
competency of the applicants to be issued
licenses
to operate
and
direct the operation of the reactivity controls of a commercial
nuclear
power
facility in accordance
with 10 CFR Part 55 and
"Operator License
Examiner Standards,"
Revision 7,
Supplement
1, Sections
200 (series),
300
(series),
and
400 (series).
Further,
the inspection
included evaluations of
facility materials,
procedures,
and simulation capability used to support
development
and administration of the examinations.
These
areas
were
evaluated
using the guidance
provided in the areas of NUREG-1021 cited above.
Both applicants
were being reexamined.
Accordingly, the written examination
was waived based
on prior successful
completion in accordance
with 10 CFR 55.35
and
Similarly, the administrative topics
and control
room
systems/facility walk-through part of the examination
was waived for both
applicants.
After completion of the evaluations,
the examiners
recommended
that both
applicants satisfied the requirements
and both have
been
issued
the appropriate
licenses.
1. 1
Facilit
Haterials
Submitted for Examination
Develo ment
Host of the materials
had
been
submitted for development of the initial
reactor operator
examinations
administered
by the
NRC in October
1994.
In
addition, the licensee
updated
the materials prior to the re-examination.
The
materials
were adequate
in scope,
depth,
and variety for examination
development.
1.2
0 eratin
Examinations
The examiners
developed
comprehensive
operating tests
in accordance
with the
guidelines of NUREG-1021,
Revision 7, Supplement
1, Section
301.
The
operating examinations
consisted of an evaluation of integrated plant
operations.
The examiners
previewed
and validated the operating
examination
on Hay 22,
1995, with the assistance
of facility training personnel
under
security agreement.
The examination
was administered
on Hay 22,
1995.
The examiners
evaluated
the two applicants
on two scenarios
using the Diablo
Canyon plant-specific simulation facility.
The licensee
provided
a training
instructor to act
as
a surrogate for staffing of a three
person operating
crew
for purposes
of the examination.
The examiners
compared applicants'ctual
performance
during the scenarios
with expected
performance
in accordance
with
the requirements
of NUREG-1021,
Revision 7,
Supplement
1, Section
303, to
evaluate applicants'ompetency
on the operating
examination.
0
The applicants
demonstrated
good
command
and communication discipline during
the simulator scenarios.
The applicants
exhibited familiarity with facility
procedures
and were quick to evaluate
which procedure
was required
and then
locate
and reference that procedure.
Crew feedback
was solicited
and
appropriately
incorporated
into responses
to events.
The prioritization of
event responses
was appropriate.
1.3
Simulator Fidelit
During the preparation
and conduct of the operating
examination,
the examiners
observed
one minor discrepancy
in simulator fidelity, as described
in
Attachment 2.
The discrepancy
did not affect the validity of the operating
examination.
2
LICENSED OPERATOR REgUALIFICATION PROGRAM EVALUATION (IP 71001)
During the inspection,
ongoing annual requalification examinations
were
observed
and programmatic
requirements
were reviewed to assess
whether the
licensee's
requalification program
was evaluating operators'astery
of
training objectives
in accordance
with 10 CFR Part 55.
This included review
of the examination material
and
an assessment
of the examination
evaluators'ffectiveness
in conducting examinations.
2. 1
0 eratin
Examination
The inspectors
evaluated
the dynamic scenarios
being used in the operating
examination with respect
to the guidelines of NUREG-1021, section
604.
The
licensee
examined operating'rews utilizing two different scenarios
for each
crew during the week of the inspection.
The format and content of the
scenarios
were consistent
with the guidelines of NUREG-1021
and adequate
to
discriminate safe operator performance.
The initial conditions of the
scenarios
were realistic
and the scenarios
consisted of related events.
The
scenarios
had
been previously validated
by the training staff and allowed the
evaluators
to measure
the examinees'ompetencies
commensurate
with the
scenario objectives.
2.2
0 erator
Performance
An inspector
observed
one shift crew and
one crew composed of on-shift and
staff personnel
on two scenarios
each in the dynamic simulator examinations.
The crew composition included
a shift supervisor,
a shift foreman,
a senior
control operator,
two control operators,
and
a shift technical
advisor.
For
the shift crew, the inspector
noted satisfactory
performance of the crew and
individual operators,
with adequate
communications,
command
and control.
Some
examples
of nondirected
communications,
informal wording, multiple
conversations,
and failures to repeat
back were noted.
The mixed crew also
exhibited satisfactory
performance,
however,
the inspector
observed
conditions
which challenged
the
command,
control
and communication function during
response
to an emergency condition.
0
Diablo Canyon Administrative Procedure
OPI.DC11,
Rev.
1,
"Conduct of Control
Operations-Abnormal
Plant Conditions," establishes
the structure of the
control
room staff during abnormal
plant operations.
In accordance
with this
procedure,
the shift foreman provides overall direction
and supervision of the
control
room activities while the senior control operator
assumes
the duty of
procedure
reader.
The senior control operator is expected
to interact with
the shift foreman whenever there
are procedure transitions,
entry points,
or
questions
as to the effectiveness
of the procedure.
During observations
of
the mixed crew, the inspector
noted that the shift foreman spent
a
considerable
amount of time in activities that did not lend themselves
to
maintaining overall direction of control
room activities.
These
included
protracted
conversations
with both the shift supervisor
and the shift
technical
advisor,
and telephonic
communications
with plant personnel.
At the
same time, the senior control operator
adopted
a role beyond that of procedure
reader
and included directing
and providing responses
to questions
from
control board operators.
The consequence
of the assumption of those roles
by
the shift foreman
and the senior control operator
was that
command
and control
of control
room activities
and communications
among operators
suffered.
Overall coordination in response
to the scenario
events
became
focused
on the
senior control operator
and the involvement of the shift foreman
was
fragmented.
As
a result of the performance
exhibited by the mixed crew, the inspector
reviewed the licensee's
programmatic
requirements
which delineate
the general
authorities
and responsibilities of operating shift personnel.
During this
review, the inspector
noted that the position of senior control operator
was
allowed to be filled by either
a licensed reactor operator or licensed senior
reactor operator.
Further review disclosed that seven of the nineteen
designated
senior control operators
are licensed reactor operators
while the
remaining twelve have senior reactor operator licenses.
The senior control
operator in the mixed requalification examination
crew was
a licensed reactor
operator.
In addition, the inspector verified that,
on Hay 23 and 24,
1995,
the on-shift Unit 2 senior control operator
was
a licensed reactor operator
and not
a licensed
senior reactor operator.
The inspector
expressed
concern
that
a licensed reactor operator
was being placed in a position of directing
the licensed activities of other licensed reactor operators.
10 CFR Part 50.54(l) specifically authorizes
only those licensed
as senior reactor
operators
to direct the licensed activities of licensed operators.
The
licensee
agreed that the observations
made
by the inspector of the constituted
crew were of concern,
but did not agree that the senior control operator
was
directing the activities of other licensed reactor operators for actual
operating
crews.
The licensee's
Updated Final Safety Analysis Report Section
13. 1.2.3, "Shift
Crew Composition," provides
a description of the minimum shift organization
that the licensee
provided to satisfy
The Final
Safety Analysis Report
was originally submitted
as part of the application for
the operating license
and
has
been periodically updated
in accordance
with 10 CFR Part 50.71.
The most recent
update,
Revision
10,
was issued
on April 27,
1995.
Figure
13. 1-3 is referenced
by Section
13. 1.2.3
and describes
the
senior control operator
as requiring
an
NRC senior reactor operator license.
The licensee's
current practice of allowing licensed reactor operators
to fill
the position is not consistent with the Updated Final Safety Analysis Report
and represents
a deviation from a commitment
(275;323/9504-01).
2.3
Evaluations
The inspector
observed
the post scenario
evaluations.
The observations
and
analyses
made
by the training department
evaluators
were detailed
and well
focused.
The facility evaluators
identified crew strengths
and weaknesses
as
well as individual strengths
and weaknesses.
The evaluators
rated the
examinees'ompetencies
by comparing actual
performance
during the scenarios
against
expected
performance
in accordance
with the guidance of NUREG-1021.
The evaluators
did not record that the senior. control operator,
a licensed
reactor operator,
was directing the activities of other licensed reactor
operators.
As
a result,
no performance
feedback
was provided to the operator.
3
PRIHARY-TO-SECONDARY LEAKAGE NONITORING AND RESPONSE
An inspection
was performed to determine
the effectiveness
of the licensee
programs
and actions
concerned
with monitoring of and response
to steam
generator
primary-to-secondary
leakage.
The areas
reviewed included handling
of generic communications
related to steam generator
tube integrity, the
adequacy of procedures
and equipment to provide real time information on leak
rate
and rate of change of leak rate,
the adequacy of alarm set points
on
radiation monitors
used for detection of leakage
and for alerting operators
to
any increasing
primary-to-secondary
leak rate,
and operator training.
3. 1
Licensee
Res
onse to Generic
Communications
The inspectors
reviewed the licensee's
handling of specific generic
communications
related to steam generator
tube integrity.
The inspectors
reviewed the licensee
evaluations of Information Notices 88-99,
"Detection
and Honitoring of Sudden
and/or Rapidly Increasing
Primary-to-
Secondary
Leakage";
91-43,
"Recent Incidents Involving Rapid Increases
in
Primary-to-Secondary
Leak Rate";
and 93-56,
"Weakness
in Emergency Operating
Procedures
Found
as
a Result of Steam Generator
Tube Rupture."
The licensee's
actions
in response
to these
generic
communications
were found to be
appropriate
and included implementation of applicable mitigation strategies.
The inspector
reviewed the licensee's
actions
in response
to the industry's
Significant Operating
Experience
Report
(SOER) 93-1,
"Diagnosis
and Hitigation
Leakage
Including Steam Generator
Tube Ruptures."
This document
was evaluated
by the licensee's
Independent
Safety Engineering
Group
(ISEG)
and resulted
in eight recommendations,
dated
June 7,
1994.
The
Plant Staff Review Committee
(PSRC)
concurred with six of the eight
recommendations
in June,
1994.
The necessary
"Action Requests"
to implement
the resulting
recommendations
were not generated
by the
ISEG until Harch,
1995.
This untimely action resulted
because
of informal processes
which
allowed the generation of the Action Requests
to be overlooked.
The delay in
implementing the recommendations,
which resulted
from the
ISEG evaluation of
an industry report,
indicates
a potential vulnerabili'ty which, in this case,
did not significantly impact the safety of the plant.
3.2
Procedures
and
E ui ment Ade uac
for Leak Rate Information
Chemical Analysis Procedures
AP-I, Revision 2,
"Prompt Steam Generator
Leak
Identification Procedure,"
and D-15, Revision 6,
Leak Rate
Determination,"
were reviewed
by the inspector.
These
procedures
were found
to provide good bases for promptly identifying which steam generator
is
leaking or ruptured
and for quantifying the primary-to-secondary
leak rate.
The referenced
radiation monitoring equipment
was capable of measuring
the
concentrations
of contaminants
in the secondary
system
and comparing those to
the concentrations
in the primary system,
identifying and quantifying the leak
rate
and thereby minimizing the potential for large-scale
secondary
contamination
and effluent releases
by isolating the affected
3.3
Alarm Set pints
on Radiation Monitors
The inspector
reviewed the alarm set points
on radiation monitors
used in each
unit for detection of leakage
and alerting operators
to an increasing
leak
rate.
The inspector
determined that the radiation monitors in both units,
that- were used for detection of leakage,
were set suitably low to ensure rapid
notification of any increasing
primary-to-secondary
leak rate.
Confirmation
of any primary-to-secondary
leak rate indications were to be
made through
independent
chemistry measurements.
The first indication of primary-to-secondary
system leakage
was provided
by
redundant
condenser air ejector monitors which were Victoreen beta
scintillators that monitor noble gases.
These monitors provided alerts at
settings equivalent to 20 gallons per day and high alarms at
50 gallons per
day.
The chemistry department
provided operators daily values for equating
gallons per day to counts per minute.
This allowed operators
to linearly
scale
the air ejector readings
and monitor changing
leakage rates.
Any
increase of 50 gallons per day in four hours
was then considered
a good
indicator for a potential
tube rupture
and appropriate
action
was to be taken.
Main steam lines were monitored with upgraded
gamma detectors
which alarm
an
alert at the equivalent of six gallons per minute
and high alarm at 46 gallons
per minute.
These detectors
provided indications of significant leakage.
More accurate
results
were provided
by counting
samples
taken from each
steam
generator within 30 to 45 minutes.
A common
blow-down monitor
served
as
a backup to the other two methods
and alarmed at
an equivalent
250
gallons per day.
Sample lines from each
also
passed
over
radiation detectors,
serving
as
an additional
backup.
0
3.4
Ade uac
of Emer enc
0 eratin
Procedures
and
0 erator Trainin
I
The inspectors
reviewed the adequacy of the licensee's
emergency
operating
procedure
(EOP)
E-3, Revision l2,
Tube Rupture,"
and
abnormal
operating
procedures
(AOPs) AP-3A, Revision 6,
Tube Leak,"
and AP-3B, Revision 8,
Tube Failure."
The inspectors
determined that the
and
EOP required the operators
to
evaluate
the magnitude
and trend for various parameters
such
as pressurizer
level,
RCS pressure,
activity levels,
and primary-to-secondary
leak rate.
A
minor inconsistency
was noted
between
procedures
in the level of detail given
for securing the turbine-driven auxiliary feedwater
pump
and isolating it from
the affected
The licensee
issued
an On-The-Spot
Change which
added valve numbers for performing this activity to procedure
AP-3B and
resolved the inconsistency.
The inspectors
evaluated
the capability of an operating
crew to identify and
trend primary-to-secondary
leakage
by conducting
a dynamic scenario
in the
facility simulator.
The scenario
also served to verify the adequacy of
detection
and alarm indications in real time.
The scenario started with an
imposed
leakage of 0.01 gallons'er
minute which ramped to 0.03 gallons per
minute in 10 minutes.
A 0.03 gallon per minute leak equates
to about
50
gallons per day.
The operating
crew successfully
detected
the leakage within
the first minute.
A subsequent
step
change to 0. 1 gallon per minute, or 150
gallons per day,
was also detected.
Subsequent
step
changes
to
1 gallon per
minute
and
75 gallons per minute were also detected.
Crew actions in response
to the changing
leakage
were appropriate.
The reviewed procedures
adequately
provided control
room operators
the
guidance
necessary
for continued monitoring,
assessment,
and response
to
identified primary-to-secondary
leakage.
The symptoms
and entry conditions
were sufficiently diverse,
such that the operators
could correctly diagnose
a
tube rupture
and enter the steam generator
tube rupture
procedure.
The performance of an operating
crew in response
to simulated
leakage conditions
was good.
ATTACHHENT 1
PERSONS
CONTACT AND EXIT HEETING
1
PERSONS
CONTACTED
1. 1
Licensee
Personnel
D. Bahner,
Operations
Engineer
- J. Becker,
Operations
Director
- K. Bych,
ISEG Supervisor
- G. Deardorff, Senior Control Operator
- B. Exner,
TS Supervisor
- S. Fridley, Outage Services Director
W. Fujimoto, Vice-President
- J. Fuhriman,
Engineer
J.
Gardner,
Senior Engineer
- B. Glynn III, gA
- T. Grebel,
Regulatory Services Director
- J. Griffin, Learning Services Director
- C. Harbor,
NRC Interface
C. Hartz,
gA Engineer
- R. Jett,
Learning Services
Supervisor
- D. Hiklush, Operations
Hanager
D. Oatley, Acting Haintenance
Hanager
K. O'eil,
ILC Engineer
- D. Taggart,
Nuclear Safety Engineering Director
- D. Vosburg,
Engineering Services Director
- J. Welsch,
Learning Services
Supervisor
J.
Young,
NgS Acting Hanager
1.2
NRC Personnel
- H. Tschiltz, Senior Resident
Inspector
In addition to the personnel
listed above,
the inspectors
contacted
other
personnel
during this inspection period.
- Denotes personnel
that attended
the exit meeting.
2
EXIT HEETING
An exit meeting
was conducted
on Hay 25,
1995.
During this meeting,
the
inspector
reviewed the scope
and generic findings of the report.
The licensee
did not express
a position
on the inspection findings documented
in this
reports
The licensee
did not identify as proprietary
any information provided
to, or reviewed
by, the examiners.
0
-10-
ATTACHMENT 2
SINULATION FACILITY REPORT
'acility
Licensee:
Facility Docket Nos:
Pacific
Gas
& Electric
50-275/323
Operating Test Administered on:
Hay 23,
1995
This observation
does
not constitute
an audit or inspection finding and is
not, without further verification and review, indicative of noncompliance with
This observation
does not affect
NRC certification or
approval of the simulation facility other than to provide information which
may be used in future evaluations.
No licensee
action is required in response
to this observation.
While conducting the validation of scenarios
for the operating examination,
the seismic
ground acceleration
level required to trip the reactor
was 0.4 g.
The in-plant seismic trip setpoint value is 0.3 g.