ML17177A443
| ML17177A443 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 05/18/1992 |
| From: | Knop R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML17177A442 | List: |
| References | |
| 50-237-92-09, 50-237-92-9, 50-249-92-09, 50-249-92-9, NUDOCS 9206020123 | |
| Download: ML17177A443 (17) | |
See also: IR 05000237/1992009
Text
U. S. NUCLEAR REGULATORY COMMISSION
REGION II I
Report Nos.
50-237/92009(DRP); 50-249/92009(DRP)
Docket Nos.
50-237; 50-249
License Nos.
Licensee:
Commonwealth Edison Company
Facility Name:
Dresden Nuclear Power Station, Units 2 and 3
Inspection At:
Dresden Site, Morris, IL
Inspection Conducted:
March 23 through May 1, 1992
Inspectors:
Approved By:
Inspection Summary
W. Rogers
M. Peck
K. Shembarger
A. Madison
P. Lougheed
!ff c fic-<;t
R. C. Knop, Ch1ef,
Projects Section 18
s /; .'{/ci z_
Date
Inspection from March 23 through May 1. 1992 <Report Nos. 50-237/92009CDRP);
50-249/92009CDRP)).
Areas Inspected: A special unannounced safety inspection was conducted by
resident, headquarters, and regional inspectors to review previously
identified items regarding (1) the failure to close of a reactor recirculation
discharge valve, (2) flooding concerns on the diesel generator cooling water
pumps, and (3) recurrences of previous violations.
Results:
Five apparent violations were identified:
Inoperability of the low
pressure coolant injection system (Section 3), Failure to have procedures as
required by Part 21, (Section 4), Deficiencies in the corporate quality
assurance control of the motor operated valve program (Section 5),
Inoperability of the Unit 2/3 diesel generator under cribhouse flooding
conditions (Section 6), and Deficiencies in the corrective action process
resulting in failure to take corrective actions on a condition adverse to
quality and in repeats of previous violations (Sections 6-9) .
9206020123 920519
POR
ADOCK 05000237
O
1.
DETAILS
Persons Contacted
Conrnonwealth Edison
1K. Graesser, General Manager, Boiling Water Reactor (BWR) Operations
1C. Schroeder, Station Manager
28. Adams, Regulatory Assurance, Engineering and Construction (ENC)
1 *2s. Berg, Assistant Superintendent - Production
1 *2E. Carrol, Regulatory Assurance
2C. Collins, Site Engineer, Nuclear Engineering Department (NED)
1L. Gerner, Technical Superintendent *
2D. Hoffman, Nuclear Quality Programs
1D. Karjala, Performance Improvement Director
1J. Kish, On-Site Nuclear Safety
1J. Kowtowski, Production Superintendent
1W. Morgan, Corporate Nuclear Operations
2H. Mulderink, BWR Motor Operated Valve Coordinator, NED
1K. Peterman, Procedure Manager
i. 2R. Radtke, Regulatory Assurance Supervisor
2R. Ralph, Assistant Technical Staff Supervisor
2R. Rybak, Mechanical and Structural Design Supervisor, NED
2T. Schuester, Nuclear Licensing Supervisor
2G. Smith, Assistant Superintendent - Operations
1M. Strait, Technical Staff Supervisor
2D. Taylor, Regulatory Assurance Supervisor, ENC
18. Viehl, NED Site Supervisor
U.S. Nuclear Regulatory Commission
18. Burgess, Chief, Section lB
1
2A. Hsia, Acting Chief, Section 18
2W. Rogers, Senior Resident Inspector
i.zK. Shembarger, Reactor Engineer
1Present at the exit held by A. Madison on April 16, 1992.
2Present at the exit held by P. Lougheed on May 1, 1992.
The inspectors also talked with and interviewed other licensee employees
throughout the course of the inspection period.
2.
Action on Previously Identified Items
A.
(Closed) Unresolved Item (237/90021-0l(DRP)) "Reactor
Recirculation Discharge Valve 2-202-SA Failure to Close."
As
discussed in inspection report 237/91036(DRP), this unresolved
item bad three issues which required review:
(1) Completion and
review of the licensee's 10 CFR Part 21 evaluation, (2) further
NRC review of the licensee's condition adverse to quality system,
and (3) further~NRC *review"o~f theTfCensee's takuhtional
2
-*
B.
controls. These reviews have been completed and the conclusions
are discussed in Sections 4, 5, and 10 of this report. Therefore,
this item is considered closed.
(Closed) Unresolved Item (237/92005-01)
"Failure to Take Adequate
Corrective Actions to Prevent Recurrence of Previous Violations."
The causes of the recurrences of these violations are discussed in
Section 1 of this report.
Based on the conclusions reached there,
this item is considered closed.
3.
Reactor Recirculation Discharge Valve Failure to Close Resulting in the
Low Pressure Coolant Injection Svstem Being Inoperable
A.
Background:
The low pressure coolant injection (LPCI) system is
one of several emergency core cooling systems (ECCS) that operate
following a loss of-coolant accident (LOCA).
It takes suction
from the torus and returns water to the reactor vessel through the
reactor recirculation injection line.
In order to ensure that
LPCI water is injected into the reactor vessel, the appropriate
reactor recirculation loop discharge valve is required to close.
B.
Descriptjon of Event:
As discussed in special inspection report
50-237/91036(0RP), during the fall 1990 Unit 2 refueling outage,
an incorrect torque switch setting was established for the 2A
reactor recirculation discharge valve (2-202-5A) motor operator
due to misinterpretation of the zero point on a liberty
Technologies Valve Operation Test Evaluation System (VOTES) trace.
After completion of refueling activities, Unit 2 resumed operation
on January 4, 1991.
On August 6, 1991, operators attempted to close valve 2-202-5A
prior to restarting the reactor recirculation pump following
maintenance.
The operators then discovered that the discharge
valve torque switch was incorrectly set. The torque switch
setting was corrected on August 10, 1991.
C.
Analysis of Root Cause:
As discussed in special inspection report
50-237/91036(DRP), the root cause of the valve being inoperable
was the misinterpretation of the VOTES trace, due to valve stem
anomalies that were not apparent to the valve testers. Because of
limitations in the VOTES testing program, including the
methodology, training and software, the person performing the
valve testing was not equipped to properly evaluate the reactor
recirculation discharge valve data.
0.
Assessment of Safety Significance: If a LOCA had occurred between
January 4, 1991, and August 10, 1991, the 2A reactor recirculation
discharge valve would not have closed.
Because the valve was
unable to close, the LPCI system may not have been able to achieve
its intended safety function. Technical Specification 3.5.A.5
allows continued operation for a maximum of seven days, with LPCI
inoperable fo~-any reason, __ This seven-day period-was exceeded.
3
This is considered an apparent violation of Technical
Specification 3.5.A.5 (237/92009-0l(DRP)).
4.
10 CFR Part 21 Procedural Deficiencies
A.
Background:
During the review of the failure of the reactor
recirculation valve, the NRC identified an unresolved item
regarding the licensee's Part 21 evaluation process.
10 CFR Part 21.21 requires a licensee to have procedures to*
evaluate deviations and to ensure that appropriate licensee
management is made aware of any defects found during the
evaluations, and th~t appropriate notifications are made once a
defect is identified.
B.
Description of Inspection Findings: Three separate screenings for
Part 21 applicability were performed on the reactor recirculation
valve failure.
On August 10, 1991, the Operations Engineer
initially screened the valve failure to close and concluded a Part
21 evaluation was not required.
On August 12, 1991, either the
technical staff supervisor, or an assistant technical staff
supervisor, completed a reportability screening of the event and
again concluded a Part 21 evaluation was not required. On
September 4, 1991, the On-Site Review Committee reviewed the event
investigation and proposed corrective actions. Their
investigation concluded the incorrect torque switch setting was
the result of an inappropriate zeroing of the VOTES trace and that
a Part 21 evaluation was not required.
In all three reviews, the
only documentation for Part 21 screening was a checklist box
marked "no."
On January 14, 1992, the NRC completed a special inspection which
determined that the root cause of the event was VOTES process
limitations. The NRC further concluded that the licensee's
initial LER was incorrect when it concluded that an improved
version of the VOTES software would have identified the condition.
The licensee began a Part 21 evaluation, and on February 5, 1992,
they determined that a defect existed associated with the VOTES
process and issued a Part 21 ~otification.
C.
Analysis of Root Cause:
The inspectors identified the following
causes as contributing to the licensee's failure to identify that
a Part 21 evaluation was required:
Technical staff management appeared to be aware of Part 21
requirements when it came to hardware issues. However, they did
not have the necessary procedural guidance or background to
address Part 21 issues in the areas of software, methodology, or
training .
4
Dresden Administrative Procedure (DAP) 2-8, "Deviation Reporting,*
provided the basic definition of a deviation and noted that
existence of a deviation would require further evaluation by the
corporate engineering staff. However, it did not provide
sufficient guidance to allow site personnel to identify software,
methodology, or training Part 21 issues.
The licensee did not have a formal training program on Part 21
requirements for onsHe personnel.. Training was mainly on-the-job
exposure to non-conforming conditions, usually hardware
discrepancies.
D.
Assessment of Safetv Sjgnificance:
Because of the above factors, while there were three screenings,
no evaluation for Part 21 applicability was performed until after
the NRC inspection questioned the acceptability of not performing
such an evaluation.
When the evaluation was performed, a
deviation was identified. This deviation was further evaluated
and a Part 21 defect was discovered.
The initial inadequate
screenings could have resulted in this defect not being identified
and reported.
The failure to have appropriate procedures addressing how to
identify and evaluate a deviation is considered an apparent
violation of 10 CFR 21.21 (237/92009-02(DRP)).
5.
Coroorate QA Program
A.
Background:
The corporate based nuclear engineering department
(NED) implemented a motor operated valve (MOY) testing program to
meet the requirements of Generic Letter 85-03 and NRC Generic
Letter 89-10.
They contracted with an architect/engineering (A/E)
firm to reestablish the design basis for MOVs, under the A/Es
accepted quality assurance (QA) program.
The A/E provided to NED
calculated valve thrust values to be used with the VOTES software.
NED then transmitted these thrust values to the site where they
were used to determine required MOY torque switch settings.
Following the reactor recirculation valve failure, the NED MOY
group reanalyzed all 39 safety-related VOTES tests performed
during the December 1990 refueling outage. During this
reanalysis, five valves were identified to have thrust values
outside the design specification. One valve was reported to the
NRC as being non-conforming in a response to Generic Letter 89-10.
For the other four valves, the design specification was revised
and retransmitted to the station to be incorporated into the next
testing effort.
B.
Description of Inspection Findings:
During follow up inspection
efforts as to the root cause of the reactor recirculation valve
failure _to c.lose,, NRC found that the licensee did not have any
5
c.
procedures cover;ng their act;vit;es on the MOY program.
NRC
ident;fied that NED rout;nely altered controlled calculat;ons ;n
order to reflect actual f;eld data such as torque sw;tch
tolerance, valve lubrication h;story, measurement equipment
uncertainty and spr;ng pack capacity, or to remove design marg;n
in order to bring non-conforming valves into conformance.
These
adjustments were performed by pen and ink markups, without
justificat;on of the new values and without any formal review of
either the assumptions or the final conclusions.
In some cases,
the final thrust window values issued by NED to the station were
outside of the thrust windows established by the Bechtel
calculations.
NED stated that these adjustments were acceptable
because the new maximum values were still below the valve or
actuator calculated structural limit. Specific examples of
adjustments include valves 2-1001-018, 3-1301-1, 3-1001-05A, and
3-3702; however numerous other examples exist.
NRC also identified that, due to the method by wh;ch the l;censee
incorporated the;r quality assurance program, NED d;d not have an
overall procedure addressing handl;ng of condit;ons adverse to
quality.
Instead the governing procedure for the work be;ng done
was to provide instructions for dealing with nonconforming
conditions. However, in the case of the MOY program, because no
procedure governed program activities, there was no guidance
w;thin NED on how to deal with conditions adverse to quality.
Therefore, when NED reevaluated safety-related VOTES tests results
following discovery of the incorrect 2-202-5A torque sw;tch
setting and d;scovered that 17 of the 37 requ;red rezeroing and
that 4 valves had thrust values outside the target window, there
was no mechanism to document these find;ngs, to ensure correct;ve
actions, or to prevent recurrence.
The four valves mentioned
above were:
2~1001-2A, 2-1001-28, 2-1501-5C, and 2-1501-32A.
At NRC's request, the licensee's nuclear quality programs group
(NQP) performed an audit specifically of the MOY program within
engineering during the week of March 23-27, 1992.
The audit
results confirmed that the same deficiencies existed within the
NED MOY program as identified by the inspectors and an audit
action item was generated.
Analysis of Root Cause:
NED considered the A/E thrust values to
constitute the design basis. Therefore, NED did not prepare any
procedures to address QA requirements for the.MOY program, and did
not believe that generic procedures, such as QE 51.D "Controlled
Analysis Originated by Nuclear Engineering Department", appl;ed to
the;r activit;es.
NED did not consider the alterat;ons being done
to the controlled MOY calculations to constitute des;gn changes,
because the des;gn basis was not be;ng altered.
NED also fa;led to recogn;ze that plac;ng correct;ve act;on
requirements ;n a program procedure rather than a gener;c
corrective act;c;ms procedure could result in sHuaHons**where
6
....
i
D.
procedural guidance on handling of conditions adverse to quality
did not exist.
These attitudes appeared to be due to a deficiency in the
licensee's QA manual, which implements the NRC-approved QA Topical
Report.
The procedures in the QA manual applied only to the
operating stations, and to *construction* activities. Activities
affecting quality performed by corporate engineering for operating
units were not addressed in the QA manual.
The licensee
considered the activities of the corporate engineering staff to
fall under the *construction* portion of the QA manual and
overlooked the transition from engineering support of construction
activities to engineering support of operational activities.
- Assessment of Safety Significance:
NRC determined that the
litensee failed to incorporate the motor Qperated valve program
activities into procedures specific to that activity. This is
contrary to the requirements of 10 CFR Part 50, Appendix B,
Criterion V, *instructions, Procedures, and Drawings* which
requires that activities affecting quality be prescribed by
documented procedures, instructions, or drawings appropriate* to
the circumstance (237/249-92009-03a(DRP)).
Because no program procedures existed, and because the licensee
concluded that the overall program requirements specified in
procedure QE 51.D did not apply, when changes to the design were
made the licensee did. not ensure that the requirements of
10 CFR Part 50, Appendix B, Criterion III, "Design Control,* were
met.
Criterion III requires that design changes be subject to the
design control measures conunensurate with those applied to the
original design and be approved by the organization that_performed
the original design (237/249-92009-03b(DRP)).
Because neither program specific nor generic procedures contained
information on dealing with a condition adverse to quality, when
valve thrust values outside the design specification were found,
they were not identified as a condition adverse to quality, and
corrective actions were not promptly taken. This is contrary to
the requirements of 10 CFR Part 50, Appendix B, Criterion XVI,
"Corrective Action" which requires that measures be established to
assure that conditions adverse to quality are promptly identified
and corrected, and, for significant conditions adverse to quality,
that the cause of the condition is determined and corrective
action taken to preclude repetition (237/249-92009-03c(DRP)).
The violations were considered to constitute a serious deficiency
in the licensee's control of their quality assurance program in
the area of motor operated valve testing. This is considered an
apparent violation of 10 CFR Part 50, Appendix B, Criterion II
"Quality Assurance Program", with the previously discussed
failures to meet the requirements of Criteria Ill, V, and XVI as
examples (237 /24~-9~0_0~-0-~ (QRPJ) ...
7
6.
Ojesel Generator Flooding
A.
B.
Background:
The emergency diesel generators provide electrical
power to safety related components in the event of a loss of
offsite power.
The diesels are cooled by water supplied via the
diesel generator cooling water pumps, which are located in the
cribhouse below ground level. If a cooling water pump fails,. then
the diesel generator will overheat within three to ten minutes,
depending on load.
The diesel generator cooling water pumps are located at the bottom
of the cribhouse at elevation 495' adjacent to the circulating
water pumps. Approximately 30 seconds following a circulating
water pump seal failure, the diesel generator cooling water pumps
would be submerged.
Based upon a calculation which does not take
into account volume displaced by piping and equipment, the maximum
time between the seal failure and complete submergence of the
diesel generator cooling water junction boxes and transfer switch
is approximately three minutes, providing power to the circulating
water pump motors is not terminated before then.
If power to the
circulating water pump motors is removed before the three minutes,
or if offsite power is lost concurrently with the seal failure,
then the diesel generator cooling water pumps themselves would be
submerged.
The remainder of the cribhouse would flood only to the
river level.
As the junction boxes and transfer switch are
located above the maximum river level, they would not be affected .
As Jong as the pumps, and.the power feeds at the pump motor were
~ubmergence qualified, the pumps would be available to provide
cooling water to the emergency diesel generators.
Oescriotjon of Events:
In June 1972 Quad Cities had an internal
flooding event due to circulating water boot seal failure.
In
August 1972, the Atomic Energy Commission (AEC) required the
licensee to evaluate the potential for a similar failure at
Dresden.
The event to be evaluated wa~ the catastrophic failure
of a circulating water pump discharge expansion joint. The
circulating pump motors were to be assumed unaffected by the
flood, becasue they were above the maximum flood level, and to
continue to operate, resulting in the continued discharge of water
into the cribhouse pit at a predicted flow rate of approximately
250,000 gallons per minute.
The loss of circulating water would
cause a loss of condenser vacuum, which would result in turbine
and reactor trip.
Upon the reactor trip, offstte power was
assumed to be lost, in accordance with design basis scenarios. In
subsequent correspondence, the licensee noted that the diesel
generator cooling water pumps were susceptible to internal
flooding.
The licensee committed to install submersible pumps
with canned motors so that the pumps would not be lost during the
internal flood. A maximum flood level to 517' elevation was
discussed in the licensee's response to the AEC questions.
The
pumps were installed in 1973; however, electrical junction boxes
located at aeero~i!Tla!e]y tb~e:_510~' elevation were not waterproofed.
8
c.
In November 1986, because of 10 CFR Part 50, Appendix R concerns,
a transfer switch was installed for the 2/3 diesel generator
cooling water pump.
This transfer switch would select an
appropriate electrical power feed for the pump in the event of a
fire.
As part of this modification, the cable to the pump motor
at elevation 492' was replaced.
The personnel responsible for the
modification overlooked the requirement for the pumps and their
power feeds to be submergence qualified. Therefore, when they
replaced the cable, they failed to reseal the motor connections
against water intrusion.
On several occasions in 1989, 1990, and 1991, contractor employees
of the licensee identified to the licensee that the 1972
convnitment to the AEC did not appear to have been met.
In June 1991, the power feeds to the Unit 3 diesel ge~erator
cooling water pump were sealed against water intrusion.
In December 1991, the licensee identified that the issue had not
been handled in accordance with their procedural requirements and
the power feeds to the Unit 2 and 2/3 diesel generator cooling
water pumps were sealed. This included resealing the Unit 2/3
power feed that was unsealed in 1986.
In January 1992, the licensee identified that the transfer switch
connections for the 2/3 pump had not been sealed in December.
They were promptly sealed.
In April 1992, while going through the li~ens~e's correspondence
on this issue, the NRC questioned whether the electric cables were
submergence qualified. The licensee determined that the cables
were submergence qualified through a document review and
Also in 1992, the licensee determined that the scenario of a
condenser boot seal failure with a subsequent loss of offsite
power was not within their design basis. They concluded that only
design basis accidents, defined in Chapter 14 of the final safety
analysis-report, coupled with a coincident loss of offsite power,
were within the design basis. They acknowledged that they had not
met an NRC commitment, but concluded that the diesel generators
would have been operable under all design basis scenarios. These
conclusions overlooked the 1986 transfer switch modification to
the Unit 2/3 cooling water pump, which resulted in the electrical
power feed going into the motor at the 495' elevation not being
resealed following cable replacement.
Analysis of Root Cause:
The initial failure to seal the power
feed boxes against water intrusion appeared to be due to an
oversight as to the maximum flood level which could occur
following a circulating water seal failure .
9
-*
The 1986 failure to reseal the Unit 2/3 power feed at the pump
motor following the transfer switch modification was due to
overlooking the requirement for submergence qualification during
the safety evaluation process. The safety evaluation failed to
address the potential for water intrusion into the diesel
generator cooling water pump resulting in loss of that pump with
subsequent loss of the associated diesel generator. Because of
the unsealed power feeds, the Unit 2/3 diesel generator would have
been inoperable under following a circulating water line break
with concurrent loss of offsite power.
The 1989-1991 failures to take corrective actions were due to the
licensee assuming that the junction boxes and cables were
submergence qualified, without verifying the assumption.
D.
Assessment of Safet~ Significance: Betwen November 1986 and
December 1991, a circulating.water pump seal failure coupled with
a concurrent loss of offsite power would have resulted in the Unit
2/3 diesel generator cooling water pump being inoperable due to
water intrusion into the electrical feeds.
The inoperability of
the diesel generator cooling water pump would have resulted in the
Unit 2/3 emergency diesel generators overheating when called upon.
If a single failure of another diesel generator was assumed to
occur, for any reason, then one unit could have been in a station
blackout condition. This would have severely affected the plant's
capability to achieve shutdown.
The root cause of these feeds
being unsealed was an inadequate safety evaluation which failed to
account for pump submergence. This is an apparent violation of
10 CFR 50.59 (237/249-92009-04(DRP).
Additionally, the failure to protect the diesel generator cooling
water pumps power feeds from *flooding, as previously corrunitted to,
is a significant condition adverse to quality. The existence of
this significant condition adverse to quality was first identified
to the licensee in November 1989.
However corrective actions were
not taken until late 1991 or early 1992.
The failure to promptly
identify and correct the condition adverse to quality is an
apparent violation of 10 CFR Part 50, Appendix B, Criterion XVI
"Corrective Actions" (237/249-92009-05a(DRP)).
7.
Failure to Follow Dresden Administrative Regyirements
A.
Background:
The Dresden Administrative Procedures (DAPs)
delineate requirements for performance of tasks throughout the
plant. They are called out in Technical Specification (T/S) 6.3
as required procedures, and must be followed in order to show
compliance with the technical specifications.
The NRC previously identified violations in the area of failing to
follow DAPs as follows:
10
- ,
1.
Between April 1 and August 30, 1990,
the licensee failed to
maintain the Control Rod Drive Accumulator High Water/Low
Pressure Alarm Log in accordance with the requirements of
OAP 9-12, "Procedural Adherence Deficiencies." Corrective
actions for this violation were to:
Issue a weekly list of new and revised Dresden Operating
Procedures and selected OAP procedures to all licensed
operators to ensure that they are aware of procedure changes
which altered their day-to-day routines.
Hold tailgate sessions on the ~equiremenis of OAP 9-12.
Develop a list and sununary of 66 DAPs, and-present them
during tailgate meetings to all station personnel.
2.
On December 27, 1990, the licensee conunitted to perform
training on various DAPs as a corrective action to an
escalated violation on failure to perform a safety
evaluation required by 10 CFR 50.59.
The wording used in
the corrective actions to this violation was identical to
that of the corrective actions for the violation above.
B.
Description of Event: --The corrective actions to these violations
were ineffective as attested by the following recurrences:
1.
On March 7, 1992, the licensee performed Dresden Operating
Surveillance (DOS) 6600-03.
During performance of the
procedure, the 2/3 Diesel Generator vent fan and fuel oil
transfer pump failed to transfer to Unit 3 power when
expected.
However, the licensee failed to complete OAP Form
9-llA, Procedural Conunent Supplement, as required by OAP
9-11 "Procedure Usage and Adherence," when the failures
occurred.
2.
Also on March 7, 1992, during performance of the same
procedure as above, several steps were performed locally by
station personnel.
The test leader, stationed in the
control room, signed off on these steps without directly
observing the steps being performed and without including
the initials of the persons actually performing the steps as
required by OAP 9-11.
3.
On April 1, 1992, the standby liquid control storage tank
air sparge inlet valve 3-1101-36 was identified to be open
and unlocked although DOP 0040-M4 required it to be locked
closed and independently verified in the locked closed
position.
When the valve was manipulated on March 20, 1992,
personnel did not use either an approved procedure or outage
checklist, nor was an operator in continuous attendance, as
required by OAP 7-14, "Control and Criteria for Locked
Equipment and Valves."
11
- 1* ,, *
c.
Analysis of Root Cause:
During the inspections, personnel
involved in the repeated violations repeatedly stated that they
were not aware of the administrative requirements, and none of the
personnel recalled training on administrative procedures.
The
following reasons for the lack of awareness were identified:
The OAP su11111aries provided to the departments to be covered
in the tailgate meetings were brief, and not all departments
covered all the DAPs when the tailgates were held.
Some
department heads stated that OAP training was partially
given, to those present at the time, and other department
heads did not remember the extent of the training sessions.
Documentation did not exist for a substantial portion of
station.personnel that should have received OAP training.
Previously identified problems with procedural inadequacies
and revision timeliness have discouraged personnel from
using the DAPs.
D.
Assessment of Safety Significance: While the individual
recurrences of the failure to follow OAP requirements were
relatively minor, the underlying cause, personnel not being aware
of administrative requirements; is more significant.
Additionally, the fact that corrective actions to previously
issued violations, including one at Severity Level Ill, did not
preclude repetition of personnel being unaware of the requirements
gives rise to the concern that more serious recurrences could
happen.
The failure to adequately implement corrective actions
for this violation is considered an example of an apparent
violation of 10 CFR Part 50, Appendix 8, Criterion XVI,
"Corrective Action" (237/249-92005-05b(DRP)).
8.
Failure to Report Inadvertent Engineered Safety Features Actuations
A.
Background:
The NRC also previously issued violations for failing
to make required notifications in accordance with the requirements
of lOCFR 50.72. These previous violations were:
1.
On December 8, 1990, the licensee failed to report an
unplanned. engineered safety feature (ESF) actuation in
accordance with the requirements of 10 CFR 50.72. The
corrective action was to issue a memorandum to operations
personnel defining an ESF actuation as any unplanned or
unknown occurrence involving the actuation of an ESF train,
which resulted in the completion of the desired
repositioning of any piece of equipment.
2.
On July 4, 1991, the licensee again failed to report an
unplanned ESF actuation. The NRC issued a violation of
10 CFR Part 50, Appendix B, Criterion XVI, "Corrective
12
B.
Action,* for the ineffective corrective actions from the
previous violation.
In response to this violation, the
licensee committed to:
Provide training to the shift engineers (SE) and shift
control room engineers (SCRE).
Clarify the guidance provided to operations personnel.
The clarified guidance defined an ESF actuation as:
- unplanned actuation of ESF systems*or components
thereof (e.g., valve movement, pump starts) are
expected to be*reported regardless of what caused the
actuation, even if the actuation was unnecessary or
- was not directly initiated by ESF actuation signals*.
Develop a flow chart to aid in determining
reportability requirements and provide it by early
first quarter, 1992.
Place a copy of NUREG 1022, Licensee Event Reporting
System, in the control room to aid in reportability
determinations.
Description of Event:
The corrective actions to these violations
were insufficient to prevent further failures to make required
notifications as noted below:
1.
On March 14, 1992, the high pressure coolant injection
(HPCI) suction valve (MOY 3-2301-6) unexpectedly opened
during the Unit 3 integrated leak r*te test when high
drywell pressure provided an ESF actuation signal to the
HPC I system. Shift operations management failed to
recognize the opening of the MOY as an unplanned ESF
actuation, and did not report the event within four hours as
required. After NRC inquiries into the event, operations
management determined a 10CFR50.72 report was appropriate.
The report was made on March 18, 1992.
2.
On April 19, 1992, the Low Pressure Coolant Injection (LPCI)
minimum flow valve (3-1501-13A) unexpectedly closed twice
when cycling of valves 3-1501-38A and 3-1501-388 occurred
during performance of DOS 1500-1," "LPCI Valve Operability
Test," on Unit 3. Shift operations management failed to
recognize the closings of the minimum flow valve as an
unplanned ESF actuation, and did not report the events per
10CFR50.72(b)(2)(ii). Later management review of the events
determined a 10CFR50.72 report was appropriate.
The report
was made on April 20, 1992.
13
..
C.
Analysis of Root Cayse:
These corrective actions were either not
effective or not implemented to preclude the third and fourth
failures to report an ESF actuation. Specifically:
For the third failure, the SE did not consider the event
reportable because the operation of the MOY was not spurious
and the intended function was accomplished, that is, the
valve went open.
For the fourth failure, the SE did not consider the event to
be reportable because only the minimum flow valve closed and
the SE thought that the LPCI suction and discharge val~es
were the only ESF components in that system.*
The flow chart was not available to_ the operations personnel
to aid in ascertaining reportability requirements.
Personnel involved in the development of the flow chart did
not consider it developed enough to be released to shift
personnel. A new target date of June 1, 1992, was selected
by the licensee to accomplish the corrective action.
However, NBC was not notified nor was a variance from the
violation response *commitment requested.
NUREG 1022 was* available in the control room; however, it
was not used in making the reportability determination.
The responsible SE and SCRE-during the MOY actuation did not
recall the NUREG 1022 training. Additionally, none of the
shift personnel interviewed had utilized NUREG 1022 to aid
in a reportability determination.
(The inspectors confirmed
training was provided.)
D.
Assessment of Safety Sjgnificance:
As discussed above, neither of
the individual recurrences was of major safety significance.
However the fact that two nearly identical repeats of previous
violations occurred within a five week period indicates that the
corrective actions to the previous violations failed to prevent
recurrence. This is another example of an apparent violation of
10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action"
(249/92005-05c(DRP)).
9.
Failure to Consider Systematic Evaluation Program Commitments in
Performing Safety Evaluations
A.
Background:
On February 8, 1991, a temporary alteration (T/A) on
the Unit 2 HPCI system provided an interface between class lE
electrical equipment and non-safety measuring and test equipment
(M&TE).
This was contrary to a systematic evaluation program
(SEP) commitment to incorporate the electrical isolation
philosophy of IEEE 384 and Regulatory Guide 1.75 for plant
modifications whenever practical (1985 CECo letter from B. Rybak
14
B.
to R. Bilbert {NRR)).
The corrective action to the violation was
to revise OAP I0-02, "IOCFR50.59 Safety Evaluation/ Screening" to
incorporate a safety evaluation screening review work sheet.
Description of Event:
Although the OAP was revised, it was not
used when on March I9, I992, M&TE to monitor voltage was installed
on the auxiliary compartment of ESF 4I60 VAC Bus 34-I under
temporary alteration {T/A) 111-7-92.
The T/A provided an indirect
interface between Class IE electrical equipment and non-safety
H&TE.
The IO CFR 50.59 safety evaluation did not address the
probability or the consequences of malfunctioning M&TE or the
bases for why the Class IE circuit would be protected following a
malfunction of the M&TE.
C.
Analysis of Root Cause:
Neither the safety evaluation preparer or
reviewer used the checklist or were aware of the convnitment to
The reasons for this are very similar to those in the
examples of failure to follow the DAPs.
D.
Assessment Of Safety Sjqnjfjcance:
As with the other examples,
the individual failure is not significant; however the failure of
the corrective action program to prevent recurrence is
significant. This is another example of an apparent violation of
IO CFR Part 50, Appendix B, Criterion XVI,
"Correcti~e Action"
{249/92005-05d{DRP)) .
IO.
Review of Inspection Report 50-237/9I036CDRPl
Upon review of inspection report 50-237/9I036{DRP) as part of ensuring
that all aspects of the unresolved item had been closed, the following
areas were found to either be misleading or to contain violations of
minor safety significance:
A.
Section 4.b: Failure to Document Personnel Oyalification
The inspection report stated that "the licensee had not maintained
records documenting station and corporate personnel qualification
to NOD-MA.I requirements." During further review, it was found
that .station records complied with the Convnonwealth Edison Nuclear
Quality Assurance Manual requirements.
Procedure NQA-I of this
manual requires that records of the implementation of
indoctrination and training are to be maintained in the form of
attendance sheets, training logs, or personnel records.
Review of
site personnel training records for site personnel showed that the
required courses had been taken and documentation was available.
This section of Inspection Report 50-237/9I036 should be amended
to read "the licensee had not maintained records documenting
corporate personnel qualification to NOD-MA.I requirements."
Corporate records were not available, as identified by the
licensee in Quality Assurance/ Nuclear Safety Audit CE-9I-04.
The
licensee determined that required training had been perfont!ed but_
-
I5
documentation was unavailable. The licensee conunitted to
reestablish corporate training records to show compliance with the
NOD requirements.
The inspectors verified that this convnitment
was met.
The failure to retain training records is a violation of
10 CFR Part 50, Appendix 8, Criterion II, "Quality Assurance
Program.* However, the violation was licensee identified, is of
minor safety significance, and corrective actions were initiated.
Therefore this violation meets the criteria for a noncited
violation as discussed in 10 CFR Part 2, Appendix C, Section V.G.l
(1991).
B.
Section 4.c:
Use of an non-safety-related consulting service
The inspection report identified that safety-related work
(consulting services as to the proper zero point to be used to
determine the as-left torque switch setting) was performed under a
non-safety-related contract. This should have been characterized
as a violation of the requirements of 10 CFR Part 50, Appendix 8,
Criterion IV "Procurement Document Control* which requires, in
part, that measures be established that applicable regulatory
requirements necessary to assure adequate quality are suitable
included or referenced in the documents for procurement of
_material, equipment, and services. However, no violation will be
cited because the designation of the contract as non-safety-
rel ated had no bearing on the torque switch settings, the
classification appeared to be an isolated example, and corrective
actions were initiated. Therefore the criteria for a noncited
violation under 10 CFR Part 2, Appendix C, Section V.A have been
fulfilled, and no violation will be .issued.
C.
Section 4.d: Test Procedyre Oyality
The inspection report stated that "Neither the work package nor
the testing procedure delineated any quantitative or qualitative
acceptance criteria related to the lOTES diagnostic, the as-found
or as-left torque switch settings or the valve thrust windows.*
Further review found this statement to be partially incorrect.
The test package did contain the target thrust values to be used,
along with a letter from corporate engineering specifying that
setting the thrust values outside the target values required
corporate approval. Although they were not specifically
identified as acceptance criteria, the personnel performing the
testing regarded them as such. Additionally, qualitative test
acceptance criteria (that the valve stroke open and closed) were
included. This section of Inspection Report 50-237/91036 should
be amended to read "Although the work package and the testing
procedure did not contain VOTES diagnostic quantitative or
qualitative acceptance criteria, the corporate engineering
approved target thrust windows were included with the package and
were considered to be acceptance criteria by the testing
personnel."
16
D
Section 4.d goes on to state that *The HOV coordinator's concerns
about selecting the zero point on the valve and use of industry
expert services were not documented.*
The HOV coordinator should have documented his concerns on form
- l
OAP 9-llA, Procedural Comment Supplement, as required by OAP 9-11
1
- procedure Usage and Adherence,* when the failures occurred.
He
failed to do so. This is an example of failure to follow Dresden
Administrative Procedures which occurred in December 1990.
The
failure of the HOV coordinator to follow the OAP requirements and
document the concerns with the recirculation discharge valve would
not have prevented the valve failure.
The violations identifying
problems with failing to follow DAPs were issued in November and
December 1990. Therefore, corr~ctive actions to these violations
would not have prevent~d this* violation. Based on this, the
criteria of 10 CFR Part 2, Appendix C, Section V.A were satisfied,
and no violation against 10 CFR Part 50, Appendix B, Criterion V
will be cited.
11.
Violations for Which a Notjce of Vjolation Will Not Be Issyed
The NRC uses the Notice of Violation as a standard method for
formalizing the existence of a violation of a legally binding
requirement.
However, because the NRC wants to encourage and support
licensee's initiatives for self-identification and correction of
problems, the NRC will not generally issue a Notice of Violation for a
violation that meets the requirements set forth in 10 CFR Part 2,
Appendix C, Sections V.A (1991).
Violations of regulatory requirements
identified during the inspection for which a Notice of Violation will
not be issued are discussed in Section 10, above.
11.
Exit Interview
The inspectors .met with licensee representatives (denoted in section 1)
throughout the inspection period. Exit meetings were held on April 16
and May 1, 1992, to summarize the scope and apparent findings of the
inspection activities. The inspectors also discussed the likely
informational content of the inspection report with regards to documents
or processes reviewed by the inspectors during the inspection. The
licensee did not identify any such documents or processes as
proprietary.
17