ML17334A661
| ML17334A661 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 02/20/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML17334A659 | List: |
| References | |
| 50-315-97-25, 50-316-97-25, NUDOCS 9802260166 | |
| Download: ML17334A661 (32) | |
See also: IR 05000315/1997025
Text
U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Docket No:
License No:
50-315, 50-316
Report No:
50-315/97025(DRP); 50-316/97025(DRP)
Licensee:
500 Circle Drive
Buchanan, Ml 49107-1395
Facility:
Donald C. Cook Nuclear Generating Plant
Location:
'1 Cook Place
Bridgman, Ml 49106
Dates:
December 28, 1997 through January 31, 1998
Inspectors:
B. L. Bartlett, Senior Resident Inspector
B. J. Fuller, Resident Inspector
E. R. Schweibinz, Project Engineer
Approved by:
Bruce L. Burgess, Chief
Reactor Projects Branch 6
9802260ihb
980220
ADOCK 050003i5
6
EXECUTIVE SUMMARY
D. C. Cook Units 1 and 2
NRC Inspection Report No. 50-315/97025(DRP); 50-316/97025(DRP)
This inspection included aspects of licensee operations, maintenance,
engineering, and plant
support.
The report covers a five-week period of resident inspection and includes the followup to
issues identified during previous inspection reports.
~Oerations
~
The inspectors concluded that cold weather preparations had been properly implemented.
The inspectors also concluded that it was prudent to procure a stand-by boiler in order to
ensure that safety-related equipment and other plant spaces were not adversely affected
by cold weather in the event the plant heating boiler was unable to operate
(Section 01.2).
Maintenance
The inspectors concluded that the work activities observed were performed in a quality
manner with procedures present and in use.
The high quality of the Instrumentation and
Control (l&C)technician work on the Solid State Protection System (SSPS) relay repairs
was especially noteworthy. An exception to this good performance was a personnel error
by an l&Ctechnician on another activity which resulted in an invalid reactor trip signal.
In
addition, the inspectors determined that the required report regarding the inadvertent trip
signal was not going to be made to the NRC until after the inspectors questioned the lack
of a report (Section M1.1).
o
The questionable material condition of the SSPS master relay dust covers resulted in
both trains of SSPS being declared inoperable for operability under seismic conditions.
A
violation was identified when the licensee failed to make a timely report to the NRC
concerning
an unanalyzed condition that significantly compromised plant safety.
An
Unresolved Item was opened to track the review of additional data to determine when
licensee personnel were aware of the degraded condition of the SSPS relays
(Section M2.1).
Encnineering
o
Performance testing of the containment hydrogen skimming system continued during this
inspection report period. Computer modeling and engineering assessments
were being
performed in an effort to determine the as-found operability and to ensure the system
would be returned to an operable condition. An unresolved item on the as-found
condition of the hydrogen skimming system remained open.
The inspectors concluded
that the Licensee Event Report (LER) issued contained inappropriate statements
(Section E1.1).
During a review of selected LERs, the inspectors identified one event that was improperly
retracted and another event that was not reported in a timely manner.
A third LER had an
insufficient basis for retraction until the licensee performed additional calculations in
response to inspector questions.
One violation for failure to submit a timely LER was
issued (Section E8.1).
Plant Su
ort
~
No discrepancies
were noted.
Re ort Details
Summa
of Plant Status
Unit 1 remained in Mode 5, Cold Shutdown, during this inspection period. The unplanned outage
was in response
to NRC and licensee concerns with the operability of the containment
recirculation sump and other engineering issues.
Unit 2 remained in Mode 5, Cold Shutdown, during this inspection period. The unplanned outage
was in response
to NRC and licensee concerns with the operability of the containment
recirculation sump and other engineering issues.
I. 0 erations
01
Conduct of Operations
01.1
General Comments
71707
Using the referenced inspection procedure, the inspectors conducted frequent reviews of
ongoing plant operations.
The inspectors observed licensed operators closely monitoring
the panels and maintaining appropriate awareness
of plant conditions, although both
units were in Mode 5 for an extended period. Specific events and noteworthy
observations are detailed in the sections below.
01.2
Cold Weather Pre
aration
Both Units
a.
Ins ection Sco
e 71714
The inspectors verified that the licensee had implemented the procedures necessary to
prepare the plant for cold weather.
The inspectors reviewed logs and Work
Procedure R0014700, "Plant Winterization," and conducted plant walkdowns.
b.
Observations and Findin s
With both units shutdown, the licensee had to rely on the single non-safety-related
plant
heating boiler (PHB) for facility heating.
Licensee personnel obtained a commercial oil-
fired steam boiler (referred to as the alternate heating boiler) as a backup source of heat
to ensure availability of steam for plant heating requirements.
The'alternate heating
boiler was maintained in hot standby, but isolated from the plant heating system, as a
ready source of heating steam for the facility.
. The alternate heating boiler was placed into service on January 15, 1998, and again on
January 30, 1998, when the PHB was removed from service to repair system leaks.
The
alternate heating boiler maintained the facility at an acceptable temperature.
The
inspectors performed walk downs of the turbine and auxiliary buildings during the period
when the alternate heating boiler was in service to verify the absence of cold weather
problems.
0
)I
During a walkdown of the auxiliary building, the inspectors identified an opening to the
refueling water storage tank yard in the Unit 2 exterior pipeway near the Nos.
1 and 4
feed regulating valves.
The opening appeared
to be for outage related services and
allowed a significant inflowof cold air, but it did not pose an immediate threat of cold
weather damage to any equipment.
The inspectors informed the operations shift
manager, and the opening was sealed.
c.
Conclusions
The inspectors concluded that cold weather preparations had been properly implemented.
The inspectors also concluded that it was prudent to procure a stand-by boiler to ensure
that safety-related equipment and other plant spaces were not adversely affected by cold
weather in the event the plant heating boiler was unable to operate.
02
Operational Status of Facilities and Equipment
02.1
En ineered Safet
Feature S stem Walkdowns
Both Units
In addition to routine plant inspections, the inspectors used Inspection Procedure 71707
to walkdown selected portions of the essential service water system.
No operability
concerns were identified.
II. Maintenance
M1
Conduct of Maintenance
M1.1
General Comments
Ins ection Sco
e 62707 and 61726
Portions of the following maintenance job orders, action requests,
and surveillance
activities were observed or reviewed by the inspectors:
I
~
C043418, Perform airflow tests on containment hydrogen skimmer system
2-HV-CEQ-1
C043419, Perform airflow tests on containment hydrogen skimmer system
2-HV-CEQ-2
C043563, Test and repair relays on Unit 1 Train A and 8 Solid State Protection
System (SSPS)
C043565, Test and repair relays on Unit 2 Train A and 8 SSPS
C043697, Inspect diverter valves in the Unit 1 West Essential Service Water
pump strainer
J,
Revision 5, Train "A" Reactor Protection System (RPS) and Engineered Safety
Features (ESF) Reactor Trip Breaker and SSPS Automatic Trip/Actuation Logic
Functional Test
Observations and Findin s
Overall, the inspectors noted that work was performed in a quality manner with the work
packages present and in active use.
The high quality of the l&C technician work on the
SSPS relay repairs was especially noteworthy. The technicians were noted to carefully
review the procedure steps and equipment before performing each step and to closely
observe each other to avoid errors.
Additional detail concerning this job evolution is
contained in Section M2.1.
During the performance of 4030.STP.410,
an l&C technician inadvertently missed a step
which resulted in a reactor trip signal.
Licensee personnel performed a followup
investigation and took appropriate corrective action.
The failure to follow a **procedure
step-by-step,
as required by plant procedures, was a violation of 10 CFR Part 50,
Appendix B, Criterion 5, "Instructions, Procedures,
and Drawings." This non-repetitive,
licensee-identified and corrected violation is being treated as a Non-Cited Violation
consistent with Section VII.B.1 of the NRC Enforcement Policy (50-316/97025-04).
On January 8, 1998, licensee personnel called the NRC operations center to report the
inadvertent reactor protection system trip signal (reference NRC Event 33505).
On
January 21, 1998, licensee personnel called the operations center to retract Event
No. 33505 based upon additional review of the event.
On February 2, 1998, the
inspectors informed the licensee that after an NRC review of the circumstances
surrounding the event that the event was required to be reported to the NRC. Licensee
personnel stated that an LER would be issued within the 30-day requirement of
Conclusions
The inspectors concluded that the work activities observed were performed in a quality
manner with procedures present and in use.
The high quality of the l&C technician work
on the SSPS relay repairs was especially noteworthy.
An exception to this good
performance was a personnel error by an l&C technician on another activity which
resulted in an invalid reactor trip signal.
In addition, the inspectors determined that the
.required report regarding the inadvertent trip signal was not going to be made to the NRC
until after the inspectors questioned the lack of a report.
Maintenance and Material Condition of Facilities and Equipment
De raded Covers on Solid State Protection S stem Master Rela s
Both Units
Ins ection Sco
e 62707
On January 8, 1998, the licensee identified degraded covers on some master relays in
the SSPS.
The inspectors observed the licensee's investigation and short-term
corrective actions.
Licensee documents reviewed included:
~
- 1 IHP.Special Procedure (SP).SSPSA,
Revision 0, "Unit 1 Train "A", SSPS
Master Relay Replacement"
o
- 1 IHP.SP.SSPSB,
Revision 0, "Unit 1 Train "B", SSPS Master Relay
Replacement"
~
- 2 IHP.SP.SSPSA,
Revision 0, "Unit 2 Train "A", SSPS Master Relay
Replacement"
~
CR 98-0069, Covers for SSPS master relays are damaged
~
Engineering Technical Direction Memo (ENTDM)98-011, "Wiring Direction for The
Replacement of SSPS Master Relays"
~
Letter from Westinghouse
to American Electric Power (AEP-88-248), dated
May 10, 1988, Evaluation of master relay cover removal
Observations and Findin s
The SSPS master relays were installed with dust covers to keep dirt and dust from
accumulating inside the relay.
During a routine surveillance of the SSPS on January 8,
1998, an l8C technician noted that some SSPS master relay dust cover ears were
damaged or broken off. Past practice during surveillance testing of this portion of the
SSPS involved removal of the relay dust covers to allow manual actuation of the relay
contacts.
Removal of the cover required that retaining ears on the base of the relay be
pried open to release the cover.
Repeated
prying of the ears resulted in some ears being
damaged or tom off.
Condition Report (CR) 98-0069 was written to document the degraded covers.
The
prompt operability determination in the CR stated that normal operation of the relays was
not affected.
The operability determination also stated that a seismic event would likely
dislodge the covers from the relay assembly.
The operability determination concluded
that due to the light weight of the covers, it was unlikely for a dislodged cover to
spuriously actuate relays in the vicinityor to prevent relays from actuating.
Based on the
prompt operability determination, the SSPS was initiallydeclared operable with the
damaged relay covers.
The licensee's
CR preliminary reportability review also
determined that the degradation of the covers was not reportable under 10 CFR 50.72.
On January 11, 1998, the licensee performed a pull test to determine the seismic
acceptability of the covers.
A total of 38 covers (15 of 48 in Unit 1, and 23 of 48 in Unit 2)
failed the pull test.
The inspectors questioned the operability status of the SSPS.
Management and l8C personnel had determined that the relays were inoperable;
however, the relays were not logged as inoperable in the operations department
equipment out-of-service logs.
After pull testing was completed, operations department personnel declared the SSPS
operable pending engineering review of the test data.
The inspectors determined that the
job order used to test the relay covers included appropriate acceptance
criteria and that
the operations department should have declared the relays which did not meet the
acceptance
criteria inoperable.
After the inspectors questioned operations management
about the pull test results, the licensee concluded that the prompt operability
determination of the original condition report was inadequate.
Both trains of SSPS in
each unit were subsequently declared inoperable.
Based on the results of the pull testing, on January 15, 1998, the licensee determined
that the degraded SSPS relay covers were an unanalyzed condition that significantly
compromised plant safety, and a four-hour non-emergency report was made to the NRC.
10 CFR 50.72(b)(2)(i) required, in part, that the licensee report within four hours any
event found while shutdown, that, had it been found while the reactor was in operation,
would have resulted in the nuclear power plant being in an unanalyzed condition that
significantly compromised plant safety.'ecause
the pull test job order contained
appropriate. acceptance
criteria, the inspectors determined that the delay from discovering
the condition, on January 11, 1998, to reporting the condition, on January 15, 1998,
constituted a violation of 10 CFR 50.72.
(50-315/97025-01(DRP);
50-316/97025-01(DR
P)).
Replacement of the unacceptable
covers began on January 19, 1998.
Extensive
planning and coordination between the engineering, maintenance
and operations
departments were evident.
The inspectors observed portions of the planning process
and replacement of selected relays.
The I&Ctechnicians were observed to be careful,
methodical, and thorough in the performance of the relay replacement and subsequent
testing.
Preliminary data identified by the licensee showed that the SSPS relays were known to be
degraded since at least May 10, 1988, and adequate corrective actions were not taken in
a timely manner.
In May 1988, the SSPS vendor recommended
that the dust covers be
replaced or epoxied in place to prevent plastic fragments from the cover from falling into
the relay or another relay below it. The recommendation also stated that ifthe cover
began backing offthe relay during a seismic event, chattering of the normally closed
contacts could be induced.
The licensee did not followthe recommendations
of the
vendor to replace or epoxy the covers although the seismic operability of the relays was
in doubt.
Licensee and NRC persohnel are continuing to identify and review
documentation related to this issue.
Pending the results of the additional review this will
remain an Unresolved Item (50-315/97025-03(DRP); 50-316/97025-03(DRP)).
c.
Conclusions
The questionable material condition of the SSPS master relay dust covers resulted in
both trains of SSPS being declared inoperable for operability under seismic conditions.
A
violation was identified when the licensee failed to make a timely report to the NRC
concerning an unanalyzed condition that significantly compromised plant safety.
An
Unresolved Item was opened to track the review of additional data to determine when
licensee personnel were aware of the degraded condition of the SSPS relays.
0
III. En ineerin
E1
Conduct of Engineering
E1.1
0 en
Unresolved Item 50-316/97024-03
Inadvertentl
Plu
ed H dro en
Skimmer Suction Line Unit 2
a.
Ins ection Sco
e 62707
On November 26, 1997, licensee maintenance personnel identified a blockage of the
Steam Generator (S/G) No. 3 enclosure Train B hydrogen skimmer suction piping.
Subsequently,
engineering personnel initiated an assessment
of the as-found operability
and began efforts to restore the system to an operable status.
Licensee procedures and
documentation reviewed included:
~
Job Order C0043514, Investigate and repair motor operated damper 1-VMO-101
cycling at 45 degrees
~
LER 50-316/97-009-00, Blockage of containment air recirculation inlet line
~
CR 98-0033, During containment hydrogen skimmer tests on Unit 1, flow rates
were found below the acceptance
criteria.
b.
Observations and Findin s
As part of the corrective action for a blocked hydrogen skimmer line (referred to as the
CEQ system) the licensee committed to perform flow testing of both trains in both Unit 1
and Unit 2. On January 4, 1998, the licensee tested the Unit 1 CEQ fans and identified
that flows from various compartments
in both trains were lower than required by the
Updated Final Safety Analysis Report (UFSAR). Subsequently,
licensee personnel
determined that the low flow on the No.
1 CEQ system was due to a mis-positioned
shutoff damper, 1-VMO-101.
Damper 1-VMO-101 was improperly re-installed during maintenance
such that instead of
going from full closed to full open on an actuation signal it was partially open at all times.
In addition to reducing flow through the skimmer line, the mis-positioned damper also
created an ice condenser bypass flow path.
Unresolved Item (50-316/97024-03(DRP)) was initiallyopened pending the results of'a
proposed surveillance test.
During this inspection period, the licensee determined that
flows through the safety system were less than stated in the UFSAR. However, the CEQ
flows were discussed twice in the UFSAR.
In Chapter 5, "Containment," higher, normal
values were listed and in Chapter 14, "Accident Analysis," lower values were listed. The
lower values reflected more realistic assumptions.
The values identified during the
surveillance test were between the two values listed in the UFSAR.
The licensee determined that while the flowwas degraded, constituting a potentially non-
conforming condition, a specific operability decision could yet be made without further
analysis.
The initial flow tests performed during pre-operational testing were performed
with the ice condenser empty of ice. This test condition eliminated any concerns with ice
melt during testing.
Additional flow tests and computer modeling were required because
with the ice condenser filled with ice, test conditions identical to the preoperational tests
could not be established.
c.
Conclusions
Performance testing of the containment hydrogen skimming system continued during this
inspection report period. Computer modeling and engineering assessments
were being
performed to determine the as-found operability and to ensure the system would be
returned to an operable condition. An unresolved item on the as-found condition of the
hydrogen skimming system remained open.
EB
Miscellaneous Engineering Issues
E8.1
Review of Selected LERs
Both Units
a.
Ins ection Sco
e 92700
The inspectors reviewed selected LERs to verify that the reports were timely, complete,
and appropriate.
The LERs reviewed were:
LER 50-316/97003, Revision 0, Revision 1, and Revision 2
LER 50-316/97004, Revision 0, and Revision
1
LER 50-315/97024, Revision 0
b.
Observations and Findin s
b.1.
0 en
LER 50-315/97024:
Material Discovered in Containment Degrades Containment
Recirculation Sump and Results in Condition Outside Design Basis
On September
11, 1997, with the unit shutdown, an NRC inspector identified a fibrous
material, known as Fiberfrax, in the Unit 2 lower containment, a condition that was
outside the design basis of the plant. Subsequent
inspections by the licensee confirmed
this finding and resulted in the identification of additional fibrous material.
On September
17, 1997, the licensee reported this finding to the NRC in accordance with 10 CFR 50.72(b)(2)(i).
However, the licensee did not send in the LER until October 17, 1997.
Licensee personnel submitted the LER 30 days after the date this issue was determined
to be reportable instead of within 30 days of the date of discovery.
requires, in part, that the licensee shall submit a LER for any event of the type
described in this paragraph within 30 days after the discovery of the event.
10 CFR 50.73(a)(2)(ii)(B) describes,
in part, any event or condition that resulted in the
nuclear power plant being in a condition that was outside the design basis of the plant.
The failure to submit an LER to the NRC within 30 days of the date of discovery was a
violation of 10 CFR 50.73.
(50-315/97025-02(DRP); 50-316/97025-02(DRP)).
10
I
,]I
/I
l
0 en
LER 50-316/97003-03:
Performance of Dual Train Component Cooling Water
(CCW) Outage During Unit 2 1996 Refueling Outage Resulted in Condition Outside The
Plant's Design Basis
The licensee reported in Revisions 0, and
1 of the LER that during the Unit 2, 1996
refueling outage a dual train CCW train outage was planned and performed.
However,
certain sections of the UFSAR had not been properly considered and thus the plant had
possibly operated outside the design basis.
The licensee's failure to comply with the
UFSAR had been identified by NRC inspectors and documented in Inspection Report
No. 50-315/97201.
Subsequently
in Revision 2, the licensee retracted the LER based upon the
conclusion'hat
sufficient controls were in place to ensure the plant remained within the design basis.
The inspectors reviewed the licensee's basis for retracting the LER'and concluded that
the. basis was insufficient. The licensee had taken credit for the ability of maintenance
and operations department personnel to restore the CCW components within the spent
fuel pool (SFP) calculated time to boil upon a postulated loss of cooling.
The inspectors determined that taking credit for manual action lacked a sufficient
analyzed basis to support the conclusion that the CCW system could be restored in time
for the SFP to remain below the bulk temperature limit. The inspectors informed licensee
personnel of their conclusions and the licensee subsequently resubmitted
LER 50-316/97003, as Revision 3.
10 CFR 50.73(a)(2)(ii), required that the licensee report to the NRC conditions outside the
design basis.
The inspectors concluded that the licensee's withdrawal of the LER in
Revision 2 was not appropriate.
b.3.'losed
LER 50-316/97004-01:
Change to CCW Temperature Without Revision to
On August 26, 1997, with Unit 2 at 100 percent rated thermal power, the licensee
determined that the unit had operated outside its design basis during the Unit 2 1996
refueling outage.
It was determined that this event was reportable under
10 CFR 50.73(a)(2)(ii) as a condition outside the design basis.
During the refueling
outage the licensee had administratively lowered the maximum allowable CCW
temperature to 90'F. This had been done to support the thermal analysis of the SFP
during the Unit 2, 1996 refueling outage.
The UFSAR stated that the CCW system was
capable of safely handling the required safety loads at temperatures
up to 95'F.
Subsequently
in Revision 1, the licensee withdrew the LER based upon a determination
that the 90'F limit and associated
administrative controls had been reviewed as required
by 10 CFR 50.59, prior to the Unit 2 refueling outage, and that this condition did not
represent an Unreviewed Safety Question.
The inspectors reviewed the LER and determined that the design basis of the CCW
system included the capability of the CCW system to safely shutdown/maintain shutdown
the unit at a maximum initial temperature of 95'F. The licensee's analysis for the Unit 2
11
,'l
f(l
Ij
refueling outage showed that the CCW temperature was required to be maintained lower
than the UFSAR upper temperature limit in order for the SFP to be maintained within the
design basis.
This issue was reported in Inspection Report No. 50-315/97021; 50-316/97201 as
Unresolved Item 50-315/97201-23; 50-316/97201-23.
The NRC Architect Engineering
inspection team was concerned that the licensee's
10 CFR 50.59 evaluation failed to
properly recognize the safety consequence
that the CCW system could not perform its
design function to remove the required SFP heat load at its maximum operating system
temperature of 95'F, as stated in the UFSAR. Thus, the licensee was operating
outside the design basis of the CCW system and the issue was reportable under
10 CFR 50.73(a)(2)(ii) as a condition outside the design basis.
Following discussions with NRC inspectors on the 90'F limit versus the 95"F design
basis, the licensee reanalyzed the Unit 2 with more realistic assumptions.
The new
analysis demonstrated
that the SFP would be maintained within the design basis at a
CCW temperature of 95'F.
The inspectors concluded that the licensee's retraction of LER 50-316/97004 was
appropriate; however, the supporting evidence was insufficient until NRC inspectors
questioned the basis for the retraction.
The issues identified during the review of the LER
regarding the licensee's
10 CFR 50.59 evaluation of the CCW system will be tracked
under Unresolved Item 50-315/97201-23; 50-316/97201-23; therefore, this LER is closed.
c.
Conclusions
During a review of selected LERs the inspectors identified one event that was improperly
retracted and another event that was not reported in a timely manner.
A third LER had an
insufficient basis for retraction until the licensee performed additional calculations in
response
to inspector questions.
One violation for an untimely report was issued.
E8.2
Closed
LER 50-316/97009:
Blockage of containment air recirculation inlet line results in
a condition outside the design bases.
This issue is discussed
in Inspection Report
No. 50-315/97024; 50-316/97024 and above in Section E1.1 as Unresolved
Item 50-316/97024-03(DRP).
The LER did not present any new information, nor did it
include any additional licensee commitments; therefore, this LER is closed.
However, the
inspectors identified several portions of the LER which were not fullyjustified and
appeared
to be contrary to the design bases.
~
As part of the assessment
of the safety significance, the LER stated, "An
evaluation of the impact of the blockage has concluded that there would have
been adequate flow through the steam generator (S/G) enclosure to preclude
excessive accumulation of hydrogen." To support this assessment
of the safety
significance, the LER included a statement thai hydrogen would not be generated
inside the S/G compartments
and that the hydrogen generated
in the lower
compartment would be well mixed, thus having little potential for significant
amounts of hydrogen to accumulate
in the S/G compartment.
The LER also
stated that the opposite train of the hydrogen skimmer system was available
except for short periods of maintenance
and would have been capable of
providing the required flow.
12
t 'I
0
The design basis of the hydrogen skimming system as stated in the UFSAR assumed
that hydrogen would be generated
inside the S/G compartments.
Licensee personnel
informed the inspectors that the assumptions contained in the UFSAR were overly
conservative.
By using more realistic assumptions
and engineering judgement, the
licensee's engineering staff determined that lower flows through the skimmer system
would be adequate.
The inspectors considered that using different assumptions than those contained within
the UFSAR was inappropriate for evaluating the safety significance of the lower hydrogen
skimmer flow. The LER did not clearly identify that this evaluation was based on
engineering judgement and the use of assumptions different from those contained in the
design basis.
Also, the planned flow testing of the two hydrogen skimmer trains had not
been completed when the LER was issued.
Thus, the capability of either train to meet
the required flow had not been verified. These comments were provided to the
appropriate licensee staff for consideration and corrective actions, as necessary.
IV. Plant Su
ort
R1
Radiological Protection and Chemistry Controls (71750)
During the resident inspection activities, routine observations were conducted in the
areas of radiological protection and chemistry controls using Inspection Procedure 71750.
No discrepancies were noted.
S1
Conduct of Security and Safeguards Activities (71750)
During normal resident inspection activities, routine observations were conducted in the
areas of security and safeguards
activities using Inspection Procedure 71750.
No
discrepancies
were noted.
F1
Control of Fire Protection Activities (71750)
During normal resident inspection activities, routine observations were conducted in the
area of fire protection activities using Inspection Procedure 71750.
No discrepancies
were noted.
X1
Exit Meeting
The inspectors presented the inspection results to members of the licensee management
at the conclusion of the inspection on February 2, 1998.
The licensee had additional
comments on some of the findings presented.
No proprietary information was identified
by the licensee.
13
0
lI
PARTIALLIST OF PERSONS CONTACTED
Licensee
¹K. Baker, Manager, Production Engineering
¹T. Beilman, Scheduling Superintendent
¹A. Blind, Vice-President Engineering
¹J. Boesch, Maintenance Superintendent
¹D. Cooper, Plant Manager
¹S. Delong, Management Information
¹MB. Depuydt, Nuclear Licensing
¹S. Farlow, Supervisor l8C Engineering
¹M. Finissi, Supervisor, Electrical Systems
¹R. Gillespie, Operations Superintendent
¹D. Hafer, Manager, Plant Engineering
¹D. Landot, Plant Performance
¹D. Morey, Chemistry Superintendent
¹D. Noble, Radiation Protection Superintendent
¹R. Ptacek, Nuclear Licensing
¹F. Pisarsky, Supervisor, Mechanical Component Engineering
¹T. Postlewait, Manager, Design Engineering
¹P. Russell, Supervisor, Plant Protection
¹J. Sampson,
Site Vice-President
¹P. Schoepf, Supervisor, Safety-related Mechanical Systems
¹R. Stevens, Nuclear Licensing
¹J. Tyler, Manager, Plant Protection and Emergency Preparedness
¹L. VanGinhoven, Materials Management
¹A. Verteramo, Supervisor Reactor Engineering
¹S. Wolf, Performance Assurance
¹B. Burgess, Branch Chief, Region lil
¹Denotes those present at the February 2, 1998, exit meeting.
14
IP 61726
IP 71707
IP 92700
INSPECTION PROCEDURES USED
On-site Engineering
Surveillance Observations
Maintenance Observation
Plant Operations
Plant Support Activities
Onsite Review of LERs
ITEMS OPENED
50-315/97025-01
50-316/97025-01
50-315/97025-02
50-316/97025-02
50-315/97025-03
50-316/97025-03
50-316/97025-04
ITEMS OPENED, CLOSED, AND DISCUSSED
Failure to report an event outside the design basis in a
timely manner in accordance with 10 CFR 50.72
Failure to report an event outside the design basis in a
timely manner in accordance with 10 CFR 50.73
SSPS Relay Operability History
Failure to follow procedures
ITEMS CLOSED
50-316/97004-01
50-316/97009-00
50-315/97025-03
LER
Change to CCW temperature without revision to UFSAR
LER
Blockage of containment air recirculation inlet line results in
a condition outside the design bases
Failure to follow procedures
ITEMS UPDATED
50-316/97024-03
Inadvertently Plugged Hydrogen Skimmer Suction Line
15
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LIST OF ACRONYMS
AEP
bcc
CFR
CR
DCC
DPR)
EDT
ENTDM
IS,C
IHP
IR
JO
LER
Ml
NRC
PHB
SSPS
S/G
American Electric Power
blind carbon copy
carbon copy
Component Cooling Water
Code of Federal Regulations
Condition Report
Donald C. Cook
Division of Reactor Projects
Demonstration Power Reactor
Eastern Daylight Time
Engineering Technical Direction
Engineered Safety Feature
Essential Service Water
Instrumentation and Control
l8C Head Procedure
Inspection Report
Job Order
Licensee Event Report
Loss of Coolant Accident
Non-Cited Violation
Nuclear Regulatory Commission
Nuclear Reactor Regulator
Plant Heating Boiler
Public Document Room
Standard Cubic Feet per Minute
Safety Evaluation
Spent Fuel Pool
Shift Manager
Special Procedure
Senior Reactor Operator
Solid State Protection System
~
Surveillance Test Procedure
Updated Final Safety Analysis R
Unresolved Item
Memo
eport
16
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