ML14178A194
| ML14178A194 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 02/25/1992 |
| From: | Christensen H, Garner L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML14178A192 | List: |
| References | |
| 50-261-92-02, 50-261-92-2, IEB-80-11, NUDOCS 9203100039 | |
| Download: ML14178A194 (13) | |
See also: IR 05000261/1992002
Text
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UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION-II
<
101 MARIETTA STREET,N.WV.
ATLANTA, GEORGIA 30323
Report No.:
50-261/92-02
Licensee:
Carolina Power and Light Company
P. 0. Box 1551
Raleigh, NC 27602
Docket No.:
50-261
License No.: DPR-23
Facility Name: H. B. Robinson
Inspection Conducted:
January 14, 1992 -
February 7, 1992
Lead Inspector:
L'.
. Gar er, Senior Resident kfispector
Date Signed
Other Inspector: M. . Hunt, Reactor Engineer
K. . Jury, Resident Inspector
)
Approved
v:
. 0. hristensen, Section Chief*
Da e igned
Divisi n of Reactor Projects
SUMMARY
Scope:
This routine, announced inspection was conducted in the areas of operational
safety verification, surveillance observation, maintenance observation,
self-assessment, and follow-up.
Results:
A violation was identified for failure to adequately perform stroke timing of
two containment isolation valves in the manner required by the test procedure
(paragraph 3).
A non-cited violation was identified involving the failure of existing
procedures to contain adequate instructi.ons to completely test the motor
driven auxiliary feedwater subsystem initiation circuitry as required by
Technical Specification 4.8.5 and Table 4.8-1 (paragraph 3).
920a3100039 920225
ADOCK 05000261
2
Preliminary Individual Plant Examination results identified two scenarios as
contributing approximately 71 percent of the plant's 1.7E-03 total core damage
frequency.
The scenarios were a reactor coolant pump seal loss of coolant
accident (LOCA)
induced by the loss of all component cooling water pumps and
an inter-system LOCA initiated by leakage from the primary system into the
residual heat removal system (paragraph 2).
Two lower pistons were replaced in the emergency diesel generator A's engine
after the vendor notified the licensee via a 10 CFR Part 21 notification that
the pistons could contain a manufacturing flaw (paragraph 2).
A Nuclear Assessment Department audit of the Environmental
and Radiation
Control Unit identified poor radiological work practices and an ineffective
self-audit program as potential issues (paragraph 2).
In 1991 the site experienced the lowest person-rem site exposure, smallest
quantity of radwaste shipped, and smallest contaminated area since these items
have been trended (paragraph 2).
Design Basis Documentation discrepancies were being adequately evaluated and
appropriately addressed (paragraph 2).
Addressing corrective actions to preclude recurrence on non-significant
Nuclear Engineering Department
(NED)
Adverse Condition Reports (ACR)
was a
good practice. However,
an 1991 third quarter NED ACR trend report indicated
that procedures and methods may not establish conditions that effectively
prevent the occurrence of an adverse condition (paragraph 5).
REPORT DETAILS
1. Persons Contacted
- R. Barnett, Manager, Outages and Modifications
- D. Bauer, Regulatory Compliance Coordinator, Regulatory Compliance
- R. Chambers, Plant General Manager, Robinson Nuclear Project
D. Crook, Senior Specialist, Regulatory Compliance.
J. Curley, Manager, HBR Engineering Support Section, Nuclear Engineering
Department
- C. Dietz, Vice President, Robinson Nuclear Project
- J. Dobbs, Manager, Nuclear Assessment Department Site Unit
R. Femal, Shift Supervisor, Operations
- W. Flanagan, Jr., Manager, Operations
- W. Gainey, Manager, Plant Support
- J. Kloosterman, Manager, Regulatory Compliance
D. Knight, Shift Supervisor, Operations
A. McCauley, Manager -
Electrical Systems, Technical Support
R. Moore, Shift Supervisor, Operations
- M. Olinger, Senior Engineer, Nuclear Engineering Department Site Unit
R. Oliver, Manager - Risk Assessment, Nuclear Engineering Department
A. Padgett, Manager, Environmental and Radiation Control
M. Page, Manager, Technical Support
D. Seagle, Shift Supervisor, Operations
M. Scott, Manager - Support Systems, Technical Support
- R. Smith, Manager, Maintenance
W. Stover, Shift Supervisor, Operations
D. Winters, Shift Supervisor, Operations
Other licensee employees contacted included technicians, operators,
engineers, mechanics, security force members, and office personnel.
- Attended exit interview on February 12, 1992.
Acronyms and initialisms used throughout this report are listed in the
last paragraph.
2. Operational Safety Verification (71707)
The inspectors evaluated licensee activities to confirm that the facility
was
being operated safely and in conformance with regulatory
requirements.
These activities were confirmed by direct observation,
facility tours, interviews and discussions with licensee personnel and
management, verification of safety system status, and review of facility
records.
2
To verify equipment operability and compliance with TS,
the inspectors
reviewed shift logs, Operation's records, data sheets, instrument traces,
and records of equipment malfunctions.
Through work observations and
discussions with Operations staff members,
the inspectors verified the
staff was knowledgeable of plant conditions, responded properly to
alarms,
adhered to procedures and applicable administrative controls
except as discussed below, cognizant of in-progress surveillance and
maintenance activities, and aware of inoperable equipment status.
The
inspectors performed channel verifications and reviewed component status
and safety-related parameters to verify conformance with TS.
Shift
changes were observed,
verifying that system status continuity was
maintained and that proper control room staffing existed.
Access to the
control room was controlled and operations personnel carried out their
assigned duties in an effective manner.
Control room demeanor and
communications were appropriate.
Plant tours and perimeter walkdowns were conducted to verify equipment
operability, assess the general condition of plant: equipment,
and to
verify that radiological controls, fire protection controls, physical
protection controls,
and equipment tagging procedures were properly
implemented.
Preliminary IPE Results
On January 13,
1992, the licensee informed Region II of preliminary IPE
results which indicated a total CDF of 1.7E-03 per reactor year.
On
January 14,
the licensee reviewed these preliminary results in more
detail with NRR and Region II. Two scenarios, transient-induced LOCA and
interfacing system LOCA, account for 76.4 percent of the total CDF.
The major contributor to the transient-induced LOCA CDF was a common mode
failure of the CCW pumps which cascades into a loss of RCP seals and a
loss of ECCS. A loss of CCW would result in the direct loss of cooling
to the RCP thermal barrier coolers and loss of cooling to the charging
pump coupling oil coolers.
The failure of the charging pumps would
result in the loss of RCP seal injection cooling water, i. e., a loss of
all cooling to the
seals resulting in RCP
seal
failures.
Simultaneously CCW cooling to the SI and RHR seal coolers would also be
interrupted causing failure of the ECCS pumps.
This slow moving daisy
chain of events could be stopped by manually supplying cooling water to
the charging pump coupling oil cooler.
Temporary instructions were
issued on January 15,
1992,
to provide procedural steps for connecting
fire water from a nearby hose station to the existing couplings on the
charging pump coupling oil coolers.
The inspectors verified that'these
temporary instructions were adequate,
had been disseminated to the
operating shifts, and all the equipment necessary for the temporary
cooling connection had been pre-staged.
3
- The primary contributor to the interfacing system LOCA was leakage
through the normally closed RHR-750 and 751 valves, the RHR system
suction valves from the RCS hot leg, which initiates an unisolable LOCA
in the low pressure RHR system.
This coolant leakage could also steam
bind the ECCS pumps thus preventing the them from performing their safety
function.
The relative large contributor of this scenario to the CDF
resulted from the probability of the valves' double disk gate failures
due to not periodically leak testing these valves.
Leak testing every
refueling interval would result in the calculated interfacing system LOCA
contribution to the total CDF being reduced by approximately two orders
of magnitude. The licensee plans to leak test these valves during RO 14
which is scheduled to begin on March 27, 1992.
The inspectors verified
that at the end of the report period the outage schedule was being
modified to include RHR-750 and 751 valve leak testing.
Implementation of the above described procedure change and leak testing
of the RHR-750 and 751 valves will reduce the total CDF from 1.7E-03 to
4.9E-04 per reactor year.
The inspectors will continue to follow-up on
the proposed actions as part of the routine inspection program.
B Inverter Malfunction
On January 23,
1992,
the B inverter which supplies power to protection
and control circuits on instrument busses #3 and #8 was observed to have
an erratic output voltage, i. e. indicated
voltage swings of
approximately 15 volts.
Operations declared the inverter inoperable and
placed the instrument busses on their alternate supply. The inverter was
repaired by replacing the capacitors on one of the inverter's circuit
cards.
The inverter was then placed back in operation and observed for
approximately one hour prior to returning it to service.
The inspectors
observed the inverter's removal from service, troubleshooting activities,
and its return to service.
The inspectors observed that operating
personnel
took prudent precautionary measures
such
as placing the
feedwater control system in manual during the time the instrumentation
busses were being transferred.
Part 21 Involving Pistons With Manufacturing Flaws
On January 29,
1992, Coltec Industries Inc. notified the licensee that a
Part 21 report was being issued concerning the potential for piston
cracking due to a manufacturing defect.
The Part 21 Report was
apparently only applicable to HBR since vendor records indicated that HBR
is the only nuclear plant which has a Fairbanks Morse EDG with rotating
pistons. The vendor recently determined that circumferential cracking of
diesel generator pistons in and immediately below the piston ring grooves
was traceable to one foundry which had supplied castings to Coltec
Industries Inc..
The foundry has now corrected the manufacturing process
problem.
Review of records revealed that four lower pistons supplied for
HBR's EDGs in 1989 and 1991 may have come from the suspect castings. The
licensee determined that two of the suspect pistons were stored in the
warehouse; however, the other two pistons had been installed during the
-4
last refueling outage in cylinders #2 and #8 of the A EDG
.
Further
review of records was unable to determine if
these pistons had been
manufactured from the suspect castings.
Thus, on February 3, 1992, the
licensee determined that it was prudent to replace the two pistons in the
A EDG.
The inspectors observed the piston replacement on that day.
Visual inspection of the removed pistons, as well as, the two pistons in
the warehouse revealed no indications of cracking. The licensee plans to
return all four potentially suspect pistons to the vendor for testing.
Containment Isolation Valve Leakage
On January 13,
1992,
primary sample valve PS-956F,
a normally closed
outboard CIV from RCS loops 2 and 3,.failed to meet its stroke time test
requirement of 10 seconds.
This test was being conducted following
maintenance on the valve's'position indicating lights.
The penetration
was apparently isolated as required by TS 3.6.3 by closing and removing
power to the PS-956E valve, which is the inboard CIV.
During repair
efforts on January 14, it
was determined that valve PS-956E had not fully
closed and was leaking at the 'rate of approximately 830 cc/min with RCS
system pressure at 2235 psig. It was also determined that valve PS-956F
was leaking and while not measured, its leakage rate appeared to be less
than the 830 cc/min through valve PS-956E. These two valves have IVSW
injected between them.-to-,enha.nce ,the.i.r contai-nment .i.solation..,ability.
Since the valve acceptance leak rate per EST-004,
Isolation Valve Seal
Water, is at 46 psig, the determination of whether or not the valve was
exceeding its allowable leak rate based upon the 2235 psig leak rate
could not be determined by the SS.
As a result the SS initiated an
Expert Operability Determination.request-at--5:00 p.m. on-January 14.
On January 15, the actuator spring for PS-956E was adjusted in an effort
to reduce the leak rate to allow work on PS-956F.
This evolution
ultimately reduced.the leak rate to 13 cc/min at 2235 psig at 11:00 a.m..
Subsequent attempts
to. adjust the spring actuator on
PS-956F was
unsuccessful in further reducing the leak rate.
On January 16
calculation 92-C-0001 was generated by Technical Support to correlate the
leakage at 2235 psig to-46 -psig-to determine the-.signific-ance of the leak
and penetration operabilit y.- -The -c-al culation demonsvtrated that -the .13
cc/min leakage rate was low enough to meet operability requirements;
however,
it
concluded that from 5:00 p.m.
on January 14 through 11:00
a.m.
on January 15, -"the observed leakage- was -such that -the. -penetration
could not be considered -operable...".
This condition was reported as
required by 10 CFR 50.73 (a) (z) (ii) in LER 92-001.
While reviewing the history associated with. the valve failures, the
inspectors determined that problems with primary sample valve stroking
had occurred in December 1991.
At that time,
it
was identified that
valves PS-956F and 956G coul-d not be successfully timed from the local
panel
due to dual indication while closing and the loss of closed
indication, respectively.
The operators then locally timed the valves'
stem travel. as remote-timing was -not possible.
Discussions with the SS
which managed the evolution revealed that Operations felt local timing of
5
valve stem travel
was
adequate to comply with their procedural
requirements.
However, OST-701, Inservice Inspection Valve Test, section
6.3, required that. the valve stroke times be measured from control switch
actuation (or other actuating signal) to the time the valve reaches the
required position as determined by the valve position indicating lights.
This methodology is utilized by the licensee to meet ASME Section XI
stroke testing requirements and to demonstrate valve operability.
By
only timing valve stem travel, the licensee did not verify necessary.
valve disk movement via an approved or accepted methodology.
Failure to
follow the requirements of OST-701 in testi.ng these sample valves is a
violation: Failure To Test Primary Sample Valves In Accordance With
Procedures, 92-02-01.
Several concerns resulted from this evolution. The first deals with the
SS and operators being unaware what effect the December 1991 testing
methodology utilized had on assuring valve operability.
The second
concern is that while indication of inadequate CIV operation was
available (i. e.,
dual indication for PS-956F),
operations did not take
the proper actions to verify that the valve could perform its containment
isolation function nor was analysis performed to determine the cause of
the dual indication
(WR
initiation was the only action taken).
Additionally,
after the concern with the testing methodology was
identified on ACR 92-009, the timing methodology to be utilized during
future valve stroke tests and what actions to take if
a test anomaly
occurs,
was not formally nor uniformly disseminated to all operating
shifts. These concerns were discussed with the Operations Manager.
The licensee intends to leave the penetration in the configuration of
PS-956E and 956F closed with power removed until RO 14,
at which time
repairs can be performed.
Other sample points are being utilized for
monitoring purposes.
The inspectors reviewed the licensee's actions,
Operability Determination, and LER,
and do not have significant technical
concerns with the actions taken
nor
the penetration's current
configuration.
E&RC Performance
In January 1992,
NAD completed an E&RC assessment.
The draft report
identified several potential findings, the most significant of which
involved field observations of poor radiological work practices and an
ineffective self-audit program.
At the end of the report period, site
management was in the process of determining the scope of the potential
issues such that corrective actions could be implemented.
Inspection of
the licensee's response to the potential issues will be conducted by both
the residents and regional specialists as part of the routine inspection
program.
6
Review of E&RC goals for 1991 revealed a continuing improving trend in
certain key performance indicators.
Of specific note were three items
which indicated the best performance on record.
These were: 1) the
lowest site exposure,
193 person-rem;
2) the smallest quantity of
radwaste shipped,
2289 cubic feet; and 3) the smallest amount of
contaminated floor space excluding the CV, 1340 square feet.
DBD Status
The inspectors reviewed the status of the DBD effort. The licensee plans
to complete the initial DBD writing effort by the spring of this year.
The inspectors reviewed DBD discrepancies to verify that they had been
evaluated for operability concerns and completed and proposed resolutions
were adequate.
The inspectors verified that there were not operability
concerns in the open DBD discrepancies and that documentation was
generally sufficient to verify that
the discrepancy was
being
satisfactorily resolved.
In those instances in which the documentation
was not sufficient, the inspectors were supplied additional information
to demonstrate that the item was being satisfactorily addressed.
All DBD
discrepancies should be resolved by the end of the year; however,
implementation of these resolutions may take several years. As they were
identified, the most significant resolutions were being scheduled.
The
licensee had no plans to incorporate references to the resolutions in
their revised DBDs.
Since these documents contain valuable information,
the inspectors noted that including resolution references would be
beneficial.
The licensee indicated that they would consider the benefits
of providing this information in the DBDs.
M-1087 10 CFR 50.59 Review
The inspectors review the 10 CFR 50.59 safety review package for M-1087,
RHR Minimum Flow Recirculation.
The inspectors observed that the
modification package, which was still in approval routing, indicated that
no unreviewed safety question existed. However, on page C42, the safety
review package stated that "Avoiding an Unreviewed Safety Question is
contingent upon a successful Large Break LOCA calculation by our fuel
vendor that accommodates the change in delivered LHSI."
The inspectors
discussed with
the NED HBR
Engineering Support Manager
the
appropriateness
of signing the modification approval
sheet as no
unreviewed safety question existed when the safety analysis contained a
contingency.
Further discussion with the author of the safety analysis
and cognizant engineer revealed that sufficient preliminary results had
been obtained from their Fuel Section to justify that an unreviewed
safety question did not exist and that the wording in the analysis had
been poor.
The engineering manager indicated that it
was not their
practice to sign-off unreviewed safety question determinations which
contained contingencies.
The inspectors had
no further concerns
regarding the reviewed documentation.
One violation was identified.
7
I0
3. Monthly Surveillance Observation (61726)
The inspectors observed certain safety-related surveillance activities on
systems and components to ascertain that these activities were conducted
in accordance with license requirements.
For the surveillance test
procedures listed below, the inspectors determined that precautions and
LCOs were. adhered to, the required administrative approvals and tagouts
were obtained prior to test initiation, testing was accomplished by
qualified personnel in accordance with an approved test procedure, the
tests were completed at the required frequency,
and that the tests
conformed to TS requirements.
Upon test completion,
the inspectors
verified the recorded test data was complete,
accurate,
and met TS
requirements, test discrepancies were properly documented and rectified,
and that the systems were properly returned to service.
Specifically,
the
inspectors witnessed/reviewed portions of the following test
activities:
OST-207
Motor Driven Auxiliary Feedwater Pump Flow Test
OST-401
Emergency Diesels.(Slow Speed Start)
Inadequate AFW Test Procedures
On January 28, 1992, the licensee identified that existing surveillance
test procedures did not completely verify that MDAFW
pump actuation
circuitry would function properly.
Specifically, testing one set of
normally closed contacts and associated interconnecting wiring in each of
the MDAFW pump actuation circuits had not been incorporated in the test
procedures.
These contacts must remain closed and the wiring integrity
must be maintained for the MDAFW
pumps to automatically start when
low-low S/G water level or tripping of the main feedwater pumps is
detected.
TS 4.8.5 and TS Table 4.8-1 items a and e. requires periodic
testing to verify that these conditions will automatically start the
MDAFW pumps.
In addition, the untested portion of the circuitry also
included the ATWS AFW actuation feature; however, testing of this feature
is not required by TS.
This testing deficiency resulted from inadequate
overlap of tests which were performed during plant shutdowns when
portions of the actuation circuitry is normally bypassed.
Review of
plant records revealed that both MDAFW pumps had successfully started on
August 30,
1991 when a low-low S/G water level condition had occurred
after a condensate pump shaft failure resulted in reduced feedwater flow
(see
IR 91-19).
This event was sufficient to demonstrate within the
required TS frequency (refueling interval) that the untested portion of
the circuitry would function properly.
The licensee plans to implement
revised test procedures during RO 14 to fully test the MDAFW actuation
circuitry.
Similar deficiencies in TS required surveillance activities
have been previously identified.
As discussed in their responses dated
July 9, 1990,
and April 3, 1991, to Notice of Violations, a review of the
programmatic and procedural adequacy of the TS surveillance program will
be completed during 1992.
This violation will not be. subject to
enforcement action because the licensee's efforts in identifying and
8
correcting the violation meet the criteria specified in Section V.G. of
the Enforcement Policy.
This item is identified as NCV: Failure To
Provide Adequate Procedures For Performing TS Required AFW Surveillance
Activities, 92-02-02.
One NCV was identified.
4. Monthly Maintenance Observation (62703)
The inspectors observed safety-related maintenance activities on systems
and components to ascertain that these activities were conducted in
accordance with TS,
approved procedures,
and appropriate industry codes
and standards.
The inspectors determined that these activities did not
violate LCOs and that required redundant components were operable.
The
inspectors verified that required administrative, material,
testing,
radiological,
and fire prevention controls were adhered to.
In
particular, the inspectors observed/reviewed the following maintenance
activities:
PIC-002
D/P Electronic Transmitter (4-20ma)
CM-610
Emergency Diesel Piston, Ring Main Bearing And
Crankshaft Overhaul
CM-615
Emergency Diesel Generator A And B Injector
Nozzle And Injector Nozzle Adapter
WR/JO 92-ABBZ1
B Inverter Repair
WR/JO 92-ABJI1
EDG A Cylinder #2 And #8 Lower Piston
Replacement
During performance of CM-610
the inspectors observed that the lower
piston retainer plate to grove clearance was recorded as 0.00015, whereas
the measured value was 0.0015.
The correct value was subsequently
entered into the procedure.
The measured value met the 0.002 maximum
allowed clearance.
During performance of CM-615,
the inspectors noted that the procedure
which had been revised on January 15,
1992 as part of the maintenance.
procedure upgrade program contained a human factor weakness.
The
weakness involved placement of acceptance criteria in caution notes
without repeating the information in the subsequent procedure steps or on
data sheets.
Examples included the caution notes on page 16 of the
procedure which provided a +/- 25 degree acceptance criteria for
positioning the injector nozzle adapter collar studs and a requirement
that the injector nozzle holder be positioned with the word TOP in the up
position.
This item was discussed with the appropriate maintenance
personnel.
No violations or deviations were identified.
9
5. Self-Assessment (45000)
The inspectors reviewed the corporate NED corrective action program as it
pertained to HBR.
The inspection consisted of a review of all the 1991
ACRs associated with the HBR engineering project. The inspectors verified
that the ACRs
had been properly classified as significant or non
significant, were being addressed in a timely manner,
and corrective
actions appeared to be appropriate.
The inspectors noted that most
non-significant ACRs,
through not required by procedures, had corrective
actions specified to preclude recurrence.
This was considered to be a
good practice.
However, the 1991 Third Quarter Trend Report For Adverse
Condition Reports Generated By NED,
the latest report available involving
all three nuclear sites, identified a management attention item, i. e.,
"Evaluation of causal factors indicates that NED procedures and methods
may not establish conditions that effectively prevent the occurrence of
an adverse condition."
In response to this potential item, a committee
had been formed to perform a detailed review of the ACRs' casual factors
and determine what corrective actions,
if
any,
are required.
The
inspectors will follow-up on this item as part of the routine inspection
program.
No violations or deviations were identified.
6. Follow-up (92700, 92701, 92702)
(Open) URI 50-261/89-26-02, Cable Submergence Qualification. The
licensee discovered in 1989 that several
EQ cables would
become
submerged,
but felt that if the cables met the LOCA submergence testing
they were qualified.
The NRC staff did not accept this submergence
testing as meeting the intent of NUREG-0588,
paragraph 2.2(5).
The
licensee then chose to await further industry testing.
In May of 1991,
NUREG/CR-5655 was published which documented the testing of 12 different
cable products.
The licensee's list of cables that would be submerged contained 3
brand/types that were not identical or similar to those tested.
These
were identified during this inspection.
Later, discussions with the
licensee were held and the licensee is to submit additional data to the
NRC for acceptability studies.
This item remains open.
(Closed) IFI 90-19-01,
Evaluate Licensee's Inspection Program
For
Openings In Reinforced Concrete Walls With Respect To Appendix "R" and
IEB 80-11 Reinforcements.
The inspector reviewed Operations Surveillance
Test Procedure OST-623, Fire Barrier Penetration
Seal
Inspection
(Refueling) which was completed January 26,
1991.
This inspection was
performed over a period of time from April until December of 1991,. and
identified 50 fire barrier penetration which by procedure were declared
The inspector reviewed the Final Report for Supplemental
Response to IE Bulletin 80-11,
dated March 1991 which documented the
inspections and verification or justification for the fill-in materials
for the block out configuration of the 50 questionable fire barrier
10
penetrations originally identified by OST-623.
The fill-in material for
these penetration was determined and evaluated.
External loading affects
were considered and evaluated for the walls containing these penetrations.
Additional rework was required for five HVAC penetration which contained
only one row of brick and twenty-four attachments.
After correction of
these 5 penetrations the analysis and evaluation confirmed that existing
brick in-fill penetration conform to the requirement of IEB 80-11.
Engineering evaluations90-104, Generic Evaluation of HVAC Fire Damper and
Fire Door Installation Discrepancies and 90-129,
Evaluation of Fire
Barrier Penetration Seals in Fire Zone 27 were reviewed by the inspector.
The purpose of these evaluations was to determine the adequacy of the
existing fire dampers and doors and the adequacy of the fire barrier
penetrations seals in fire zone 27.
There were no concerns identified by
the inspector. This item is closed.
No violations or deviations were identified.
7.
Exit Interview (30703)
The inspection scope and findings were summarized on February 12,
1992,
with those persons indicated'in paragraph 1. The inspectors described the
areas inspected and discussed in detail the inspection findings listed
below and in the summary. Dissenting comments were not received from the
licensee.
The licensee did not identify as proprietary any of the
materials provided to or reviewed by the inspectors during this
inspection.
Item Number
Description/Reference Paragraph
92-02-01
VIO -
Failure To Test Primary Sample Valves In
Accordance With Procedures (paragraph 2).
92-02-02
NCV -
Failure To Provide Adequate Procedures For
Performing
TS
Required
Surveillance
Activities (paragraph 3).
8.
List of Acronyms and Initialisms
a.m.
Ante Meridiem
ACR
Adverse Condition Report
American Society of Mechanical Engineers
Anticipated Transient Without Scram
cc/min
cubic centimeters/minute
Component Cooling Water
Core Damage Frequency
CFR
Code of Federal Regulations
ontainment Isolation Valve
corrective Maintenance
CCopnnColnWae
CV
Containment Vessel
.
Design Basis Documentation
D/P
Differential Pressure.
E&RC
Environmental and Radiation Control
Environmental Qualifications
EST
Engineering Surveillance Test
HBR
H. B. Robinson
Heating Ventilation Air Conditioning
ie.
That is
Inspection and Enforcement
IEB
Inspection and Enforcement Bulletin
IFI
Inspector Follow-up Item
IR
Inspection Report
Individual Plant Evaluation
IVSW
Isolation Valve Seal Water
LCO
Limiting Condition for Operation
LER
Licensee Event Report
Loss of Coolant Accident
LHSI
Low Head Safety Injection
M
Modification
Motor Driven Auxiliary Feed Water
NAD
Nuclear Assessment Department
Non-cited Violation
NED
Nuclear Engineering Department
Nuclear Reactor Regulation
OST
Operations Surveillance Test
p.m.
Post Meridiem
Process Instrument Calibration
PS
Primary Sample
Psig
Pounds per square inch
gage
Reactor Coolant Pump
Roentgen Equivalent Man
Refueling Outage
S/G
Safety Injection
Shift Supervisor
TS
Technical Specification
Unresolved Item*
Violation
Work Request
WR/JO
Work Request/Job Order
- Unresolved items are matters about which more information is required to
determine whether they are acceptable or may involve violations or
deviations.