ML17292A458
| ML17292A458 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 09/12/1996 |
| From: | Johnston G, Mckernon T, Tracy G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML17292A454 | List: |
| References | |
| 50-397-96-16, NUDOCS 9609190289 | |
| Download: ML17292A458 (23) | |
See also: IR 05000397/1996016
Text
ENCLOSURE 2
U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Docket No.:
License No.:
Repoit No.:
Licensee:
Facility:
Location:
Dates:
Inspectors:
Approved By:
50-397
50-397/96-1 6
Washington Public Power Supply System
Washington Nuclear Project-2
3000 George Washington Way
Richland, Washington
July 31 through August 9, 1996
Thomas O. McKernon, Operator Licensing Examiner
Gary W. Johnston,
Senior Project Engineer
Glenn M. Tracy, Acting Deputy Director
Division of Reactor Safety
Attachment:
Partial List of Persons Contacted
List of Inspection Procedures
Used
List of Items Opened
List of Licensee Procedures
and documents Reviewed During This
Inspection
9609i90289
960912
ADOCK 05000397
8
-2-
EXECUTIVE SUMMARY
Washington Nuclear Project-2
NRC Inspection Report 50-397/96-16
From July 31 through August 9, 1996, a special inspection was conducted to review two
recent events, an early criticality occurring outside the estimated critical rod position
tolerance band during startup on June 27, 1996, and a power excursion on July 20,1996,
which resulted from the manipulation of the adjustable speed drive system by a
nonlicensed person.
The inspectors made the following conclusions.
~Oeretione
The operating crew failed to follow procedures when distracting activities were
conducted during the approach to criticality (Section 01.2).
Multiple procedural violations occurred due to an erroneous control room operator
log entry and a failure to use the correct strip chart paper on one of the source
range monitor recorders (Section 01.2).
The shift manager and the control room supervisor failed to use a conservative
decision making process,
did not exercise good command, control, and
communications,
and failed to elevate the estimated critical position problem to
upper operations management
(Section 01.2).
~En ineerin
The reluctance of the station nuclear engineers to bring forward to management
their initial concerns related to the accuracy of the estimated critical position
calculation demonstrated
a lack of conservative decision making (Section E1.2).
The continuing effort to verify the adequacy of the estimated critical position
calculations were thwarted by inadequate training in the use of the associated
software.
Therefore, combined with a lack of proceduralization for estimated
critical position calculations, the station nuclear engineers efforts merely continued
to reinforce the error in usage of the software (Section E1.2).
The lack of a reviewed and approved procedure for conducting an estimated critical
position calculation contributed to the early criticality event (Section E1.2).
The lack of an adequate
procedure with sufficient controls resulted in a nonlicensed
individual operating the reactor recirculation system and affecting reactivity
(Section 01.3).
-3-
Effectiveness of Licensee Controls and Evaluations
~
The licensee's investigations were too narrowly focused and did not identify all
facts involved in the events (Section 01.4).
4-
Re ort Details
On June 27, 1996, the reactor was restarted
and subsequently
shut down due to an error
in calculating the estimated critical rod position.
The reactor was restarted
on June 29,
and the licensee recommenced
power ascension
and testing of the digital feedwater and
adjustable speed drive modifications.
On July 20, while at 68 percent power, the plant
experienced
a short duration 15 percent power transient due to an error associated with
adjustable speed drive testing.
01
Conduct of Operations
01.1
General Comments
71707
Using Inspection Procedure 71707, the inspectors reviewed the operational aspects
of two recent events: an early criticality occurrence;
and Reactor Recirculation
Control Pump-1A speed transient during adjustable speed drive testing.
In the
former event, conduct of operations lacked the rigor of command and control
required by procedures during the approach to criticality and concerns related to
deviation from the estimated criticality band were not elevated to the appropriate
management
levels.
In the latter event,
a nonlicensed person inadvertently caused
a reactivit
erroneousl
im utin
data into the ad'ustable
speed drive
y
y
y
p
9
I
system.
01.2 Criticalit Achieved Prior to Minimum Estimated Critical Position
a.
Ins ection Sco
e 71707
On Thursday, June 27, 1996, reactor criticality was achieved prior to the minimum
expected point of criticality calculated by the station nuclear engineer.
The
inspectors reviewed control room logs, facility procedures,
data logs, and conducted
interviews with key licensee personnel.
The inspectors also reviewed the actions of
the control room shift operators to ascertain whether their actions contributed to
the estimated critical position event.
b.
Observations
and Findin s
0 erations Command
Control
and Communications
The inspectors observed that during the estimated critical position event,
a number
of barriers to effective command, control, and communications were either
breached
or circumvented:
(1)
Initial communications between the backshift shift manager and the onshift
station nuclear engineer indicated
a low confidence level in the estimated
critical position calculation.
While the shift manager questioned the results
-5-'f
the calculation, particularly with respect to its difference from the
estimated critical position calculated for the prior initial criticality of June 14,
1996, no compensatory
measures
of monitoring were deemed necessary.
Cautionary actions of Step 4.2.5 in Procedure
PPM 1.3.59, "Reactivity
Management," were discussed
between the station nuclear engineer and the
shift manager.
This step required that, "Ifcriticality occurs before the
Minimum Allowable Critical Position, stop control rod withdrawal, notify the
control room supervisor.
The control room supervisor should direct the
control room operator to drive control rods in the reverse order."
Additionally, the control room supervisor erroneously assumed,
by review of
the station nuclear engineer's estimated critical position memorandum which
was addressed
to the reactor/fuels engineering manager, that the
memorandum had been reviewed by the manager when, in fact, it had not.
The results of subsequent
discussions
between the onshift station nuclear
engineer and the estimated critical position calculation (PowerPlex) software
custodian, also a station nuclear engineer, at 0200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br />, indicated that the
offshift station nuclear engineer doubted the estimated critical position
calculation results.
However, the onshift station nuclear engineer did not
consider it necessary
to elevate the lack of confidence in the estimated
critical position calculation up to his management
nor did he share the
information with the shift manager or the control room supervisor.
At 0533 hours0.00617 days <br />0.148 hours <br />8.812831e-4 weeks <br />2.028065e-4 months <br />, Intermediate Range Monitor IRM-Det-1C appeared
to fail and
was declared inoperable by the operators.
At the direction of the shift
manager, the control room supervisor (also the reactivity manager) diverted
his attention to filling out the limited condition of operation log paperwork.
This activity and his following preparations for turnover to the oncoming
control room supervisor consumed the majority of his attention and focus
until the time of turnover.
These activities were in noncompliance with the
startup Procedure
PPM 3.1.2, Revision 31, Step 4.2.2 caution, which stated,
that, "During the approach to criticality, avoid activities that can distract the
operator at the controls and the control room supervisor."
This was an
apparent violation of Technical Specification 6.8.1a (50-397/9616-03).
At the time of turnover, the shift technical advisor advised both control room
supervisors that the plant was close to going critical. The shift technical
advisor returned to the front panel and the reactor operator and shift
technical advisor continued to withdraw control rods.
The control room
supervisors did not stop their turnover and focus on the approach to
criticality. This was another example of a failure to adhere to the Reactivity
Management Procedure.
Procedure
PPM 1.3.1, Revision 26, "Conduct of
Operation," Step 4.6.1, stated, "During periods when reactivity
manipulations are in progress or plant activities which could affect reactivity
occur, the control room supervisor shall assume the responsibility of
Reactivity Manager.
Responsibilities include ensuring
a conservative
-6-
approach to operations involving core reactivity changes."
Procedure
1.3.1, Revision 26, Step 4.6.2 (r) further stated, "Shift turnover of the
control room staff is inappropriate when criticality is imminent." This was
another example of apparent violation (50-397/9616-03).
(5)
During the control room supervisor turnover walkthrough of the control
board panels, the oncoming control room supervisor suddenly observed the
condition of the plant and the source range monitor neutron level. The
reactor operator stated that criticality was imminent. The control room
supervisor discussed
the condition with the station nuclear engineer and
was informed of the deviation from the expected estimated critical position
margin.
The control room supervisor briefly informed the shift foreman.
The shift foreman acknowledged the control room supervisor but did not
internalize the information and act upon it. Instead, the shift foreman
focused upon conducting the shift briefing while the control room supervisor
returned to the at-the-controls area to oversee control rod withdrawals.
Procedure
PPM 1.3.1, Revision 26, Step 4.6.2b stated, "The Reactor
Operator at H13-P603 shall not be distracted by control room activities such
as shift turnover, shift brief, or surveillances."
The conduct of the shift brief
during the approach to criticality was another example of failure to follow
procedures
and apparent violation of Technical Specification 6.8.1a
(50-397/961 6-03).
(6)
In followup interviews, it was determined that the shift technical advisor
who had previously reviewed the estimated critical position calculation did
not internalize the information and, as such, did not serve as a barrier to
identify and inform the control room supervisor of the out-of-tolerance
estimated critical position/criticality conditions.
Actions:
(1)
While the shift manager questioned the estimated critical position calculation
results when presented
the estimated critical position memorandum,
cautionary action statements
in the startup procedure gave the shift manager
confidence in continuing with control rod withdrawal. As discussed
above,
this action resulted in a failure to follow procedures.
(2)
The shift manager directed the control room supervisor to attend to other
duties which detracted from his reactivity manager duties and
responsibilities.
This was an apparent failure to comply with
Procedure
PPM 3.1.2.
(3)
The offgoing/oncoming control room supervisor turnover detracted from the
focus on the approach to criticality and an apparent failure to comply with
Procedure
PPM 1.3.1.
-7-
(4)
The offgoing/oncoming control room supervisors failed to discontinue their
turnover when informed that the reactor was close to criticality.
(5)
The shift manager failed to advise his upper operations management that the
early criticality was approaching
and the minimum estimated critical position
criteria would not be met.
(6)
The shift manager made
a cognizant decision to conduct
a shift briefing
while the control room supervisor and the reactor operator/shift technical
advisor continued to pull control rods.
This action was an apparent
noncompliance with procedures
as discussed
above in item (5) of the
command, control, and communication section.
(7)
The control room supervisor directed the reactor operator to pull control rod 18-47 from step 16 to 26 while the crew monitored the increasing
source range monitor count rate.
The reactor was then declared critical at
0705 hours0.00816 days <br />0.196 hours <br />0.00117 weeks <br />2.682525e-4 months <br />, June 27, 1996, with a source range monitor Channel A
detector reading 7x10'ps,
a reactor period of 345 seconds,
and a reactor
coolant system temperature of 211 degrees f.
In followup reviews of the
control room operator logs, the inspectors determined this data was not
accurately entered into the control room operator log as required by Startup
Procedure
PPM 3.1.2, Revision 31, Step 4.2.7 which stated, "Enter the
following here and in the control room operator log at the time of criticality .
.. Time, Neutron Level, Period, Control Rod Number, Control Rod Position,
Coolant Temperature."
The control room operator log was annotated
as
having a neutron level of 5000 cps.
This erroneous entry represented
a
failure to follow procedures
and a violation of Technical Specification 6.8.1a
(50-397/961 6-01 ).
(8)
Additionally, the inspectors identified that the control room operators had
erroneously installed and not identified one of the source range monitor
recorder's strip chart paper which was of the wrong type.
The strip chart
paper had a linear scale when the appropriate type was one with a
logarithmic scale.
Procedure
PPM 3.1.10, "Operating Data Logs,"
Revision 11, stated, "The purpose of recorder charts is to provide operations
and management
personnel with a permanent record of trends exhibited by
specific plant parameters."
The source range monitor recorder with the
incorrect strip chart paper would not serve the purpose of an accurate
historical permanent record.
This was a failure to follow procedures
and a
violation of Technical Specification 6.8.1a (50-397/9616-02).
c.
Conclusions
The inspectors concluded that apparent violations of regulatory requirements
existed in the operations area of responsibility.
The operating crew failed to follow
procedures when distracting activities were conducted during the approach to
-8-'riticality.
The shift manager and the control room supervisor failed to use a
conservative decision making process,
did not exercise good command, control, and
communications,
and failed to elevate the estimated critical position problem to
upper operations management.
Additionally, procedural violations occurred due to
an erroneous control room operator log entry and a failure to use the correct strip
chart paper on one of the source range monitor recorders.
01.3
Ad ustable
S eed Drive Power Excursion Event
Ins ection Sco
e 71707
On July 20, 1996, while performing tests on the adjustable speed drive system to
ascertain whether or not electronic resonance
of control circuitry might occur at
different pump speeds,
a contract engineer in the adjustable speed drive room
inadvertently caused
a 15 percent change in power from 68 to 53 percent and a
return to the initial 68 percent power level.
The event occurred while efforts were
being taken to serve as compensatory
measures
should an electronic resonance
condition appear during adjustment of the reactor recirculation pumps.
The inspectors reviewed the adjustable speed drive event and conducted interviews
with key facility personnel.
Test procedures
and other control processes
were also
reviewed.
b.
Observations
and Findin s
The inspectors interviewed key operations personnel involved with validation and
verification of Procedure
PPM 8.3.339, "Test Instructions
- Reactor Recirculation
(RRC) Adjustable Speed Drive (ASD) and Reactor Digital Feedwater
(DFW) Control
Power Ascension Test Program," Revision 1. The inspectors noted that the test
procedure had been validated for those areas which could be modeled on the plant-
specific simulator, but was not validated for test areas related to in-plant local
actions such as the input of adjustable speed drive runback values at local panels
inside the adjustable speed drive room.
Further, discussions with operations
personnel indicated that directions for local actions inside the adjustable speed drive
room were discussed
during the pretest brief in the control room.
The procedure
did not contain specific steps by which compensatory measures
were to be taken
should electronic resonance
appear.
The only written guidance that existed was as
an embedded
action statement to a caution in Step 8.9 of the procedure.
As such,
the procedure was inadequate
in that it did not contain measures sufficient to
preclude manipulation of facility controls as expressly prohibited by
10 CFR 50.54(i), in that, a nonlicensed individual operated the reactor recirculation
control system and affected reactivity.
This was considered
an apparent violation
of Criterion V of 10 CFR 50, Appendix B (50-397/9616-05).
-9-
In addition, the plant operations committee review of the test procedure had
identified a potential problem with the procedure
in this area, but had not followed
through with sufficient actions to preclude the event from occurring.
As such, the
licensee had prior knowledge of the potential for the event but did not take
sufficient measures
to preclude its occurrence.
Further, the work control process
did not identify the test evolution as a potential problem because
it did not have
procedural steps to review and, therefore, did not review the task to a level of
depth which would have identified a problem with the procedure.
c.
Conclusions
The inspectors concluded that the procedure controlling adjustable speed drive
testing for local actions in the adjustable speed drive room was inadequate
and an
apparent violation. The inspectors
also concluded that controls and barriers which
would have precluded
a nonlicensed individual from manipulating reactivity were not
established.
As a result, a nonlicensed individual was placed in a position in which
he alone could, and did, cause
an effect on core reactivity.
01.4 Effectiveness of Licensee Controls and Evaluations
71707
a.
Ins ection Sco
e
The inspectors reviewed the licensee incident review board investigations, followup
problem evaluation requests,
and conducted discussions with key personnel
involved in the events and in the investigation of the events.
Observations
and Findin s
The inspectors noted that the licensee's incident review board findings did not
include a number of the discrepancies
identified by the inspectors.
For example:
(1) the erroneous control room operator log entry related to the initial criticality
count rate (5000 cps versus 70,000 cps); and (2) the fact that one of the source
range monitor recorders had the incorrect strip chart paper installed.
Additionally, the incident review boards did not identify some root causes which the
inspectors believed were very relevant.
For example, the estimated critical position
incident review board focused on the reactor engineering aspects
and did not
address operations command and control during the events and what barriers were
breached or circumvented.
The adjustable speed drive incident review board
investigation focused solely upon the contract engineer's
actions and did not
address what barriers in the procedural review process were circumvented.
For
example, why did the operations review process
and work control process not
function properly and preclude
a nonlicensed individual from even being in a position
in which he could affect the manipulation of reactivity?
-10-
c.
Conclusions
The inspectors concluded that the licensee's investigations were too narrowly
focused and did not identify all facts involved in the events.
05 Operator Training and Qualification (41500)
a.
Ins ection Sco
e
The inspectors reviewed licensed operator training related to the approach to
criticality. The review included both the initial licensed operator certification
program and the licensed operator requalification program, as well as interviews
with key training department personnel.
b.
Observations
and Findin
s
The inspectors noted that the initial licensed operator certification training included
the appropriate knowledge and abilities competencies
and these were linked to
simulator and classroom training sessions.
The licensed operator requalification
program was procedurally linked to lesson plans and learning objectives.
In the
requalification training, the training associated
with reactivity focused upon lessons
learned from industry events and simulator sessions
related to startup of the plant.
In the simulator requalification training, operators discussed
plant indications during
the approach to criticality and then continued with plant heatup operations.
As
such, not all reactor operators had the opportunity to pull control rods to criticality.
However, all operators were involved in training discussions.
Conclusions
The inspectors concluded that the operators had the prerequisite knowledge and
abilities related to approach-to-criticality operations.
However, the inspectors also
concluded that because
of decreased
opportunities for plant startups and because
of a greater number of licensed operators
in the program, the experience level of
operators may not have been as high as it had been in the past.
III. En ineerin
E1
Conduct of Engineering
E1.1
General Comments
37551
Using Inspection Procedure 37551, the inspectors reviewed the involvement of
station nuclear engineers
in the early criticality and the Reactor Recirculation Control
Pump-1A speed transient during adjustable speed drive testing.
In the former event,
concerns of the station nuclear engineers regarding the results of calculations of the
estimated critical rod position and the deviation from the estimated criticality band
-11-
were not elevated to the appropriate management
levels.
In the latter event,
a
nonlicensed person,
a contractor engineer, inadvertently caused
a reactivity
transient by erroneously imputing data into the adjustable speed drive system.
The
inspectors summarized other observations
related to these events in Sections 01.2
and 01.3 of this report.
E1.2 Criticalit
Achieved Prior to Minimum Estimated Critical Position
a.
Ins ection Sco
e
On Thursday, June 27, 1996, reactor criticality was achieved prior to the minimum
expected point of criticality calculated by the station nuclear engineer.
The
inspectors reviewed control room logs, facility procedures,
data logs, and conducted
interviews with key licensee personnel.
b.
Observations
and Findin s
The estimated critical position calculated for the June 27, 1996, reactor criticality
was performed utilizing the incorrect software flag for the plant condition.
The
PowerPlex software required that a specific set of conditions be input to solve for
the desired parameters.
The software accomplished this with a series of flags that
were set by the user.
The flag setting that caused the estimated critical position
calculation to be incorrect was for Xenon dependence.
Normally the flag default
is -1, which covers most cases for online use.
For the case of a plant that had been
shutdown for a short period of time, where the Xenon concentration had not
decayed fully the flag should have been set to 0.
Discussions with the software custodian, who also conducted the station nuclear
engineers training on the PowerPlex software, indicated to the inspectors that he
had questioned the vendor about the use of the -1 option.
However, the use of the
option for the case of a post-trip return to criticality was not a topic discussed
and
the custodian was left with the impression that the default value of -1 would be
used in almost all possible situations.
Therefore, when the estimated critical
position calculation was done the option of 0 was not selected.
When subsequent
calculations were conducted, the other station nuclear engineers performed the
calculation in the same method.
On June 12, 1996, the first estimated critical position following the refueling outage
was performed.
This estimated critical position was done using charts provided by
the fuel vendor, and no anomalies were found.
The estimated critical position was
validated using the PowerPlex software, which agreed with the manual calculation.
The wrong Xenon dependence
flag was set for this calculation, however, the
software automatically depletes Xenon for the first run after a refueling.
As a
result, the station nuclear engineers
did not receive prior indication of the error in
use of the software.
-1 2-
The reactor scrammed on June 24, 1996, at around 0800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br />.
On the following
day, June 25, 1996, an initial estimated critical position was calculated estimating
0000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> that night.
The station nuclear engineer recalculated
an estimated
critical position that night for a 0600 critical time on June 26, 1996.
The station
nuclear engineer that performed the calculation noted that the result was the same
as the previous calculation, which was unexpected.
The station nuclear engineer
was concerned about the calculation and expressed
his concerns to the oncoming
station nuclear engineer for the day shift. Additional runs of the software were
done, which resulted in the same results.
A manual calculation was done which
came up with a different result, that later turned out to be more accurate.
The
manual calculation, however, was discounted because
it did not include factors that
the PowerPlex software included, particularly the fact that. one rod was inoperable
and would stay in the core.
Runs on the software of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the trip and 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> gave the same results.
The station nuclear engineers
ascribed that the reason for this conclusion was that
the Xenon had decayed out of the core by 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the scram.
This was not
the case,
as subsequent
analysis showed that the amount of Xenon would have
been nearly equal to equilibrium at the time of the scram, and the 36-hour figure
was around half of the equilibrium value.
The engineers continued to analyze the
differences, requesting assistance
from the fuels engineering group to do other
calculations.
The differences between the calculations of the two groups amounted
to about 6 mK, but was discounted.
The difference between the June 12, 1996,
estimated critical position and the last estimated critical position that was calculated
was 16 mK. The engineers became convinced that Samarium poisoning accounted
for the difference.
This would explain some changes
in the expected estimated
critical position, however, not to the magnitude that occurred.
On June 27, 1996, the station nuclear engineer gave the estimated critical position
to the shift manager.
The shift manager questioned the difference between the
estimated critical position of June 12 and the new estimated critical position.
The
station nuclear engineer convinced the shift manager that sufficient independent
work was done to justify the estimated critical position.
The onshift control room
supervisor was not informed of this disparity.
The rest of the event process
is
described in Section 01.2 of this report.
The inspectors reviewed Procedure
PPM 1.3.59, "Reactivity Management Program,"
Revision,1, and noted that step 2.6.3a of the procedure stated that, "An estimated
ECP calculation should be performed by the SNE... and an acceptable
band of
delta K/K for criticality should be applied."
Procedure
PPVI 3.1.2, "Reactor
Startup," Revision 31, step 4.2.3 further required that the station nuclear engineer,
"...verify criticality occurs between the Minimum Allowable ECP Limit and the
Maximum Allowable ECP Limit." However, no approved plant procedure existed on
June 27, 1996, to perform a calculation for an estimated critical position and
applied tolerance band, nor were instructions provided within Procedure
PPM 3.1.2,
"Reactor Startup."
The inspectors noted that a draft procedure had existed for
-13-
some time but did not include sufficient instructions to preclude the miscalculation
of the estimated criticality point.
As a result, on June 27, 1996, the reactor was
taken critical approximately 4 mK earlier than the minimum allowable estimated
critical point limit. The lack of specified procedural guidance was considered
an
apparent violation of Technical Specification 6.8.1a (50-397/9616-04).
c.
Conclusions
The inspectors noted the apparent reluctance of the station nuclear engineers to
bring forward to management
their initial concerns related to the accuracy of the
estimated critical position calculation.
Further, their assurance
that sufficient
independent verification was done led the shift manager to believe the estimated
critical position was correct.
The continuing effort to verify the adequacy of the
estimated critical position calculations were thwarted by inadequate training in the
use of the PowerPlex software.
Therefore, combined with a lack of
proceduralization for estimated critical position calculations, the station nuclear
engineers efforts merely continued to reinforce the error in usage of the software.
E5
En ineer Trainin
and Qualification 41500
a.
Ins ection Sco
e
The inspectors reviewed station nuclear engineer training related to the approach to
criticality and the use of the PowerPlex software to calculate the estimated critical
position.
The review included initial training and subsequent
refresher training, as
well as interviews with the PowerPlex software custodian responsible for providing
the training on the use of the software.
Observations
and Findin
s
The inspectors noted that, for the initial training, the station nuclear engineers were
presented with all the functional attributes of the PowerPlex program.
The review
of the training indicated that the original training was sufficient and addressed
the
necessary
knowledge and ability factors.
However, subsequent
training on the
software did not include sufficient detail to ensure that changes
made to the
software by the vendor were addressed.
The vendor had recently provided new
software with changes to the flag settings related to time dependent
concentration.
The effect of the setting of the flag for the case of a cold
time-dependent
post-trip calculation for a critical rod position was not discussed
during training of the station nuclear engineers.
As a result, when the calculation
for the reactor startup of June 27, 1996, was performed, the error in selecting the
flag was repeated consistently.
This resulted in each confirmatory calculation being
performed with the same conditions set in the program.
-1 4-
c.
Conclusions
The inspectors concluded that the station nuclear engineers training on the use of
the PowerPlex software was flawed such that the personnel consistently performed
the estimated critical position calculations incorrectly.
V. Mana ement Meetin s
X1
Exit Meeting Summary
The inspectors presented
the inspection results to members of licensee management
at the
conclusion of the inspection on August 8, 1996.
The licensee acknowledged the findings
presented.
The inspectors asked the licensee whether any materials examined during the inspection
should be considered propriety.
No proprietary information was identified.
ATTACHMENT
PARTIAL LIST OF PERSONS CONTACTED
Licensee
P. Bemis, Vice President for Nuclear Operations
L. Fernandez,
Licensing Manager
G. Smith, Plant General Manager
A. Langdon, Acting Operations Manager
J. Swailes, Engineering Director
D. Swank, Regulatory and Industrial Affairs Manager
R. Webring, Vice President Operations Support
M. Monopoli, Maintenance Manager
NRC
G. Tracy, Acting Deputy Director, Division of Reactor Safety
LIST OF INSPECTION PROCEDURES USED
IP 37551:
Onsite Engineering
IP 41500:
Training and Qualification
IP 71707:
Plant Operations
LIST OF ITEMS OPENED
~Oened
50-397/961 6-01
VI0
50-397/961 6-02
50-397/961 6-03
50-397/961 6-04
VIO
50-397/961 6-05
Failure to adhere to procedures
related to control room
operator log entries
Failure to adhere to procedures
related to retention of
permanent plant records
Failure to adhere to procedures
related to reactor startup
Failure to have a procedure for calculation of an estimated
critical position
Failure to have procedure appropriate to circumstances
resulting in a nonlicensed person operating the controls of the
reactor
LIST OF LICENSEE PROCEDURES AND DOCUMENTS
REVIEWED DURING THIS INSPECTION
PPM 3.1.2, Revision 31, "Reactor Startup"
PPM 3.1.10, Revision 11, "Operating Data Logs"
-2-
PPM 1.3.59, Revision 1, "Reactivity Management"
PPM 8.3.339, Revision 1, "Test Instructions - Reactor Recirculation Adjustable Speed Drive
and Reactor Digital Feedwater Control Power Ascension Test Program"
PPM 1.3.1 Revision
26, "Conduct of Operations"
Incident Review Board Report for PER 296-0576, "RRC-PUMP-1A Speed transient during
ASD testing," dated July 25, 1996
Incident Review Board Report for PER 296-0522, "CriticalityAchieved Prior to Minimum
Estimated Critical Position (ECP)," dated July 2, 1996
Interoffice Memorandum, Subject:
Estimated Critical Position (ECP) for BOC-12, dated
June 26, 1996
ENCLOSURE 3
Enforcement Policy: Section V,
"Predecisional Enforcement Conferences"
Erd'orccmcn( Policy S~~~g
Y.
PREOECISIOHAL
EHFORCEHEHT CONFERENCES
Whenever the
NRC has learned of the existence of a potential violation for
which
escalated
enforcement
action
appears
to
be
warranted,
or
recurring
nonconformance
on the part of a vendor,
the
HRC may provide
an opportunity for
a predecisional
enforcement conference with the licensee,
vendor, or other person
before taking enforcement
action.
The purpose of the conference
is to obtain
information that will assist
the
HRC in determining the appropriate
enforcement
action,
such
as:
(I)
a
common
understanding
of facts,
root causes
and missed
opportunities associated with the apparent violations, (2)
a common understanding
of corrective
action
taken or planned,
and
(3)
a
common understanding
of the
significance of issues
and the need for lasting comprehensive
corrective action.
If the
HRC concludes that it has sufficient information to make
an informed
enforcement decision,
a conference will not normally be held unless
the licensee
requests it.
However,
an opportunity for a conference will normally be provided
before issuing
an order based
on a violation of the rule on Deliberate Hisconduct
or
a civil penalty to
an unlicensed
person.
If a conference
is not held,
the
licensee
will normally
be
requested
to
provide
a written
response
to
an
inspection
report, if issued,
as
to
the
licensee's
views
on
the
apparent
violations
and their root causes
and
a description of planned
or
implemented
corrective action.
During the predecisional
enforcement
conference,
the licensee,
vendor, or
other persons will be given an opportunity to provide information consistent with
the
purpose
of the
conference,
including
an
explanation
to
the
HRC of the
immediate corrective actions (if any) that were taken following identification
of the potential
violation or nonconformance
and the long-term comprehensive
actions
that
were
taken
or will be
taken
to prevent
recurrence.
Licensees,
vendors,
or other
persons
will be told
when
a
meeting
is
a
predecisional
enforcement
conferences
A predecisiona1
enforcement conference is a meeting between the
NRC and the
licensee.
Conferences
are normally held in the regional offices
and are not
normally open to public observation.
However,
a trial program is being conducted
to
open
approximately
25
percent
of all eligible
conferences
for public
observation,
i.e.,
every
fourth eligible
conference
involving one of three
categories
of licensees
(reactor, hospital,
and other materials licensees) will
be open to the public.
Conferences will not normally be open to the public if
the enforcement
action being contemplated:
(I) Would
be
taken
against
an individual, or if the action,
though not
taken
against
an
individual,
turns
on
whether
an
individual
has
committed
wrongdoing;
i;oforccmcnt I'olicy SLttcmcnt
(2)
Involves significant personnel
failures
where
the
HRC has
requested
that the individual(s)
involved be present
at the conference;
(3) Is based
on the findings of an
HRC Office of Investigations
report;
or
(4)
Involves
safeguards
information,
Privacy
Act
information,
or
information which could be considered
proprietary;
In addition,
conferences will not normally be open to the public if:
(5) The conference involves medical misadministrations or overexposures
and
the conference
cannot
be conducted without disclosing the exposed individual's
name;
or
(6) The conference will be conducted
by telephone
or the conference will
be conducted
at
a relatively small licensee's facility.
Notwithstanding meeting
any of these criteria,
a conference
may still be
open if the
conference
involves
issues
related
to
an
ongoing
adjudicatory
proceeding with one or more intervenors
or where the evidentiary basis for the
conference
is
a matter of public record,
such
as
an adjudicatory decision by the
Department of Labor.
In addition, with the approval of the Executive Director
for Operations,
conferences will not be open to the public where good cause
has
been
shown after balancing
the benefit of the public observation
against
the
potential
impact
on the agency's
enforcement
action in
a particular case.
As
soon
as it is determined
that
a conference
will be
open
to public
observation,
the
NRC will notify the licensee that the conference will be open
to public observation
as part of the agency's trial program.
Consistent with the
agency's
policy on
open
meetings,
"Staff Meetings
Open
to Public," published
September
20,
1994
(59 FR 48340),
the
NRC intends
to announce
open conferences
normally at least
10 working days in advance of conferences
through (I) notices
posted
in the Public Document
Room,
(2)
a toll-free telephone
recording at 800-
952-9674,
and
(3)
a toll-free electronic bulletin board
at 800-952-9676.
In
addition,
the
NRC will also
issue
a press
release
and notify appropriate
State
liaison officers that
a predecisional
enforcement
conference
has
been scheduled
and that it is open to public observation.
The public attending
open conferences
under the trial program may observe
but
not participate
in
the
conference.
It is
noted
that
the
purpose
of
conducting
open
conferences
under the trial program is not to maximize public
attendance,
but
rather
to
determine
whether
providing
the
public
with
opportunities
to
be
informed of
HRC activities is compatible with the
NRC's
ability to exercise
its regulatory
and safety responsibilities.
Therefore,
members
of the public will be
allowed
access
to the
NRC regional
offices to
attend
open enforcement
conferences
in accordance
with the "Standard
Operating
Procedures
For
Providing
Security
Support
For
NRC
Hearings
And Meetings,"
published
Hovember
1,
1991
(56 FR 56251).
These procedures
provide that visitot s
may
be subject
to personnel
screening,
that signs,
banners,
posters,
etc.,
not
larger than
18"
be permitted,
and that disruptive persons
may be removed.
md%/BR41 95
Enforcement Policy Statcmcnt
Members of the public attending
open conferences will be reminded that (I)
the apparent
violations discussed
at predecisional
enforcement
conferences
are
subject
to further review and
may
be subject to change prior to any resulting
enforcement action and {2) the statements
of views or expressions
of opinion made
by HRC employees
at predecisional
enforcement
conferences,
or the lack thereof,
are not intended to represent final determinations or beliefs.
Persons attending
open
conferences
will be provided
an opportunity to
submit written
comments
concerning
the trial program anonymously to the regional office.
These
comments
will be subsequently
forwarded to the Director of the Office of Enforcement for
review and consideration.
When needed
to protect the public health
and safety or common defense
and
security,
escalated
enforcement action,
such
as the issuance
of an immediately
effective
order,
will
be
taken
before
the
conference..
In
these
cases,
a
conference
may be held after the escalated
enforcement
action is taken.
NJREG/BR%195