ML14176A752
| ML14176A752 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 07/31/1989 |
| From: | Dance H, Garner L, Jury K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML14176A749 | List: |
| References | |
| 50-261-89-12, NUDOCS 8908230088 | |
| Download: ML14176A752 (10) | |
See also: IR 05000261/1989012
Text
rPkREou 4
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTA ST., N.W.
ATLANTA, GEORGIA 30323
Report No.:
50-261/89-12
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:
June 11 - July 10, 1989
Inspector: /4--
7/3 /
L. W. Garner, Senior Reside t Inspector
D te Si nd
K. R. yry, Resient Insplector
D te Sig d
Approved by:
3/
H. C. Dance, Section Chief
Da e Signed
Division of Reactor Projects
SUMMARY
Scope:
This routine, announced inspection was conducted in the areas of operational
safety verification, surveillance observation, maintenance observation, onsite
followup of events,
operator requalification exam failure control,
and
follow-up on previous inspection items.
Results:
One violation with two examples was issued involving inadequate design controls
for OT delta T and OP delta T setpoints.
This is indicative of a lack
of attention to details for control of reactor protection system setpoints,
paragraphs 5.a and 7.
The licensee is in the process of verifying that manual isolation of SW to
non-safety related components can be accomplished in a timely manner if
required, paragraph 5.c.
The licensee is in the process of determining why two spent fuel assemblies
have become unlatched from a fuel handling tool during spent fuel pool fuel
movements, paragraph 5.
Procedure controls and licensee's sensitivity regarding removal of licensed
operations from shift duties upon failure to pass requalification exams are
considered adequate to preclude this situation from occurring, paragraph 6.
PDR ADCr:: 05000261
u
0I
REPORT DETAILS
1.
Persons Contacted
R. Barnett, Maintenance Supervisor, Electrical
- J. Benjamin, Engineering Supervisor, Plant Systems
C. Bethea, Manager Training
H. Bryon, Instructor
R. Chambers, Engineering Supervisor, Performance
- D. Crook, Senior Specialist, Regulatory Compliance
J. Curley, Director, Regulatory Compliance
- C. Dietz, Manager, Robinson Nuclear Project
R. Femal, Shift Foreman, Operations
W. Flanagan, Manager, Design Engineering
W. Gainey, Supervisor, Operations Support
E. Harris, Director, Onsite Nuclear Safety
D. Knight, Shift Foreman, Operations
D. McCaskill, Shift Foreman, Operations
- A. McCauley, Principle Engineer, Onsite Nuclear Safety
R. Moore, Shift Foreman, Operations
- R. Morgan, Plant General Manager
D. Myers, Shift Foreman, Operations
D. Nelson, Maintenance Supervisor, Mechanical
M. Page, Acting Manager, Technical Support
- D. Quick, Manager, Maintenance
- D. Sayre, Senior Specialist, Regulatory Compliance
D. Seagle, Shift Foreman, Operations
- J. Sheppard, Manager, Operations
R. Smith, Manager, Environmental and Radiation Control
- R. Steele, Operations Coordinator
- H. Young, Director, Quality Assurance/Quality Control
Other licensee employees contacted included technicians,
operators,
mechanics, security force members, and office personnel.
- Attended exit interview on July 20, 1989
Acronyms and initialisms used throughout this report are listed in the
last paragraph of the inspection report.
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.
To verify equipment operability and compliance with TS,
the inspectors
reviewed shift logs, operations' records, data sheets, instrument traces,
and records of equipment malfunctions.
Through work observations and
2
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, cognizant of
in-process 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 routinely 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 attentive and
professional manner.
Plant tours -and perimeter walkdowns were conducted to verify equipment
operability, assess the general condition of plant equipment,
and to
verify that radiological, fire protection, physical protection, and
equipment tagging procedures were properly implemented.
No violations or deviations were identified.
3. 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, test
instrumentation was properly calibrated, and the tests were completed at
the required frequency and 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-10 (Revision 9)
Power Range Calorimetric During Power
Operation
OST-401 (Revision 22)
Emergency Diesels
OST-402 (Revision 9)
Fuel Oil System Flow Test
No violations or deviations were identified.
4.
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
3
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:
WR/JO 89-AFRK1
A Rod Drive MG Set Repair
WR/JO 89-AEZK1
Replacement of A EDG Injector Nozzles
WR/JO 89-AFDT1
Replacement of A EDG Injector Pumps 6 & 8
WR/JO 89-AFTP1
Replacement of A EDG 6 & 8 Thermocouples
The WR's concerning work performed on A EDG was performed in response to a
high differential temperature (>300'F) between cylinders. As a result, on
May 5, 1989, the injector nozzles were replaced on cylinders 6, 7, 10 and
12. After performance of this work, there was still a high differential
temperature between cylinders 6 and 8.
After consultation with the
vendor, four calibrated injection pumps were replaced (WR
89-AFDT1);
however, this work did not correct the temperature problem. At that time,
maintenance decided to replace the cylinders' thermocouples (WR 89-AFTP1);
however, again this did not correct the problem.
As a result, the vendor
is going to perform special diagnostic work on the EDG,
as well as
supplying new injector pumps to try and determine the root.cause and a
permanent solution to the temperature problem.
This diagnostic testing
and evaluation is scheduled during July 1989.
During witnessing of WR 89-AFDT1,
the inspector identified a problem.
The licensee considers replacement of a EDG injector pump to be a "skill of
the craft" work activity and as such, does not require the work performed
- to be proceduralized.
According to instructions contained on the WR, the
work was to be performed per the vendor's technical manual.
The
technicians performing the work had neither the WR nor the technical manual
at the job location. This in and of itself is not a problem; however, the
vendor's technical manual contains specific torquing requirements in their
instructions for replacing an injector.
Per Plant Program procedure
PLP-013,
Rev. 3, Maintenance Program, Section 4.0, torquing of components
which have specific torque requirements is not defined as a "skill of the
craft" work activity. Section 5.0 of this procedures states that approved
procedures, instructions, or check Tists be utilized when the activity is
beyond the skill of the craft.
There were no instructions contained in
the WR for specific torquing to be performed, nor were the required torque
values documented.
However,
upon review of the completed WR,
the
inspector saw evidence that the torque requirements may have been met as
evidenced by a completed Torque Wrench Certificate of Calibration for this
work. Additionally, the nozzle replacements performed under WR 89-AEZK1
also had torque requirements that were not documented.
In this instance
however, there was a procedure specified to be utilized which did contain
the required torque values as specified in the technical manual.
This WR
also had a completed Torque Wrench Certificate of Calibration attached,
4
thus providing some assurance the torquing was performed.
As part of the
licensee's Maintenance Procedure Upgrade Program scheduled for completion
at the end of 1990,
these type of discrepancies are expected to be
addressed. The adequacy of the Procedure Upgrade Program and initiation
of new procedures when required is considered an URI:
Ensure Adequacy of
Upgraded Maintenance Procedures, 89-12-01.
No violations or deviations were identified.
5. Onsite Followup of Events (93702)
a.
Error in OT delta T and OP delta T Setpoints
On April 10,
1989,
the licensee determined that the full power
delta T value used in the reactor protection circuitry was different
than the value assumed in certain transient analysis.
This
inconsistency introduced non-conservations in both the OT delta T and
OP delta T protection circuit setpoints as defined in TS 2.3.1.1.2.
Evaluation of the significance of this error indicated that there was
sufficient conservatism in the K1 constant of the OT delta T circuit
to more than compensate. for this error.
Hence, the OT delta T
protective funcation if required would have occurred at or before the
time assumed in the transient analysis. However, this was determined
not to be the case for the OP delta temerpature circuit.
The error
was found to result in a non-conservative OP delta T setpoint which
could exceed the TS value by nearly five percent.
Thus, if required
the OP delta T protective circuit would not have responsed as rapidly
as required by TS to limit certain postulated transients.
However,
this error in OP delta T is deemed to have only minor safety
significance in that the OP delta T function serves only as a backup
function to other protection circuits and a five percent error in the
OP delta T setpoint is not of a sufficient magnitude to constitute a
significant degradation of the function.
The above described event is documented in LER 89-007, issued May 17,
1989.
As discussed in the LER,
the inconsistency existed from
February 25, 1989, beginning of cycle 13, to the time of discovery in
April 1989.
The major contributor appears to be a breakdown in
communications between various organizations responsible for design
controls.
During refueling 13, plant modification 959,
RCS Bypass
RTDs,
was implemented to place the RCS hot and cold leg RTDs in the
main coolant piping. During this process,, the licensee failed to
maintain adequate design control to assure that a parameter utilized
in an analysis and defined in TS 2.3.1.1.2.e was correctly translated
into procedures and instructions. This is a violation of 10 CFR 50
Appendix B Criterion III:
Inadequate Design Controls Results in
Non-conservative Reactor Protection System Setpoint, 89-12-02.
See
paragraph 7 for a second example of a Criterion III violation.
5
These two examples indicate a lack of attention to detail involving a
lack of thorough and effective review process for control of.reactor
protection system setpoints.
During the exit, this item was
discussed with licensee management.
b. Unlatching of a Spent Fuel Assembly
During loading of spent fuel assemblies into the DSC on June 23,
1989,
one of the assemblies (K-18)
became partially unlatched from
the fuel handling tool in similar fashion to the occurrence described
in IR 89-09. The sequence of events occurred as follows:
(1) The spent fuel handling tool was latched on assembly K-18 (which
was the fifth assembly being loaded into the DSC)
and proper
latching was verified by Operations and Engineering personnel.
(2) Operations attempted to insert the assembly into the DSC at
which time the assembly "bumped"
the canister key-way causing
the tool cable to slacken (approximately 6").
Simultaneously,
one of four tool latching fingers became unlatched from the
assembly nozzle.
(3) After discussions between Operations and Engineering, it
was
decided that the logical procession was to attempt to set the
assembly into its respective canister position.
While
0IIIact)
t
howeve durig
t
the
han dlito
re-attempting to load the assembly into the canister, a similar
impact resulted; however, during this impact the handling tool
became completely disengaged from the top nozzle and three of
the four tool fingers latched onto the hold down springs.
(4) The situation was discussed and the decision was made to place
the assembly in the nearest old fuel rack location to take
advantage of the large funnel-shaped lead in provided in the
older fuel racks.
The assembly was successfully placed in this
location.
(5)
The
remaining canister locations were successfully loaded
utilizing a newly refurbished handling tool without encountering
further difficulties.
The fuel handling tool being utilized when the unlatchings occurred is the
same tool that was being utilized during the spent fuel assembly drop of
April 26, 1989. After that event, the licensee performed what they
believed to be adequate corrective action and root cause determination.
There was no failure mechanism determined for the April 1989 event based
on corrective actions and attempted root cause determination; the handling
tool was returned to service. Subsequent to the June events described
above, the tool was removed from service, tagged out, and will be sent to
Westinghouse for complete failure analysis including handling tool
6
dimensional checks.
Follow-up for the earlier April event was identified
in IR 89-09 as IFI 89-09-06. For administrative purposes, this item is
considered closed. The inspectors will monitor the root cause analysis
performed-for both events.
This item is an URI:
Review Root Cause
Analysis Performed on Fuel Handling Tool, 89-12-03.
C.
No Justification Exists for Reliance on Manual Isolation of SW to the
Turbine Building Under Certain Postulated Accident Conditions.
While conducting a review of the SW system,
ONS determined that
certain postulated single failures would result in the failure of the
turbine building SW automatic function during accident situations.
These single failures would also result in only two of the four SW
pumps delivering SW-to both safety and non-safety related components.
Existing analysis assumes that the two SW pumps are providing flow to
only the safety related components,
e.g., the turbine building SW
supply valves are isolated.
However,
the original design of the
plant was for manual isolation of the turbine building SW valves by
operation of the remote control switches on the RTGB.
The automatic
isolation was included within the last five years as an operator aid.
As such, the automatic control circuitry was not designed to function
under all potential single failures.
No documentation has been located which provides the time frame
within which the manual isolation must occur.
Failure to.manually
isolate the SW to the turbine building components could eventually
result in degradation or loss of some safety functions.
Thus the
failure to have the isolation time requirement defined placed the
plant in a potential
However, existing
emergency procedures do require manual isolation of the turbine
building SW supply isolation valves if low SW header pressure occurs.
These steps are to be performed early on in the procedure.
Based
upon the small chance that an accident will occur with the exact
component failures required to initiate the postulated scenarios, and
engineering judgement that existing procedural requirements are
sufficient compensatory measures if such scenarios were to occur, the
licensee believes operation can safely continue until this issue is
resolved.
The licensee has taken steps to inform operating crews of
this issue.
At the close of the report period, the licensee was
actively pursuing verification via analysis and/or possible testing
of the acceptability of reliance on manual isolation of the SW to the
turbine building. Concurrently the licensee is developing possible
modifications to upgrade the automatic isolation function to preclude
single failures. This issue is considered an URI:
Review Resolution
of Single Failure Impact on SW System Performance, 89-12-04.
No violations or deviations were identified.
7
6. Operator Requalification.Exam Failure Controls
As part of an
NRC
concern regarding licensed operators who fail
requalification exams performing shift duties, the inspectors reviewed and
evaluated the licensee's controls in this area.
Training Instruction,
TI-200, Rev. 24, Robinson Plant Operator Requalification Program, provides
a positive control mechanism for ensuring that an operator who fails a
requalification exam is prevented from performing shift duties.
This
instruction contains a methodology for notifying both the individual and
his/her supervision of the failure, and delineates the requalification and
accelerated testing necessary before the individual may return to shift
duties. Additionally, based on discussions with the licensed training
supervisor, this situation has not occurred at this facility in at least
the past eight years.
Thus, established procedural controls, historical
data, and the licensee's sensitivity in this area are considered adequate
to preclude this situation from occurring at HBR.
7. Action on Previous Inspection Findings (92701)
(Closed)
URI 261/88-03-01,
Review OT Delta T Safety Analysis Results.
This item, discussed in IR 88-03, involves discovery that an incorrect
loop transport time had been used in determination of the OT Delta T trip
setpoint.
This item was reported to the NRC by LER 88-002, issued
February 19,
1988,
and by supplement number 01 issued June 6, 1988.
At
the time of discovery, the licensee took immediate action to compensate
for the error by adjusting the K1 constant of the OT Delta T trip setpoint
to 1.09, a more conservative value than the 1.1565 value specified in TS 2.3.1.1.2.d. The LER supplement reported that the TS value of 1.1565 in
the OT delta T trip equation could not have maintained the MDNBR above the
licensed limit of 1.17 during a control rod drop transient initiated from
full power.
The calculated MDNBR for the rod drop transient was
determined to be 1.065.
This value is approximately 10 percent below the
desired 1.17 value.
However, during discussions with cognizant NRC staff
in both Region II
and NRR,
it
was determined that the reduction in
confidence level for prevention of fuel damage during the rod drop
transient had not been significantly affected.
During the evaluation of the safety significance of the incorrect transient
analysis, the contractor performing the evaluation for the licensee failed
to adjust all the parameters necessary to perform the analysis for the rod
drop transient.
This resulted in the licensee verbally -informing the NRC
that the analysis had demonstrated tht the MDNBR for the rod drop
transient had been greater than 1.17 for the incorrect loop transport
time. Subsequent reviews revealed the error and the NRC was notifed as
required. This recent breakdown in control of transient analysis as well
as the problem which initiated the event in 1975 constitute a violation of
NRC requirements. Specifically, these items contributed to the nonconser
vative OT delta T setpoints from 1975 until December 1988 and is the second
8
example of violation 89-12-02 associated with failure to establish
adequate design controls per 10 CFR 50 Appendix B, Criterion III as
discussed in paragraph 5.
With the issuance of the Notice of Violation,
URI 261/88-03-01,
LER 88-002 and supplement LER 88-002-01 are considered
closed.
8.
Exit Interview (30703)
The inspection scope and findings were summarized on July 20, 1989, 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. Proprietary information is not contained in this report.
Item Number
Description/Reference Paragraph
89-12-01
URI -
Ensure Adequacy of Upgraded Maintenance
Procedures, paragraph 4,
89-12-02
VIO -
Inadequate Design Controls Results in
Non-Conservative Reactor Protection System Setpoints,
paragraphs 5.a and 7.
89-12-03
URI - Review Root Cause Analysis Performed on Fuel
Fuel Handling Tool, paragraph 5.b.
89-12-04
URI - Review Resolution of Single Failure Impact on
SW System Performance, paragraph 5.c.
10. Acronyms and Initialisms
CFR
Code of Federal Regulations
Dedicated Shielded Canister
GL
Generic Letter
HBR
H. B. Robinson
IFI
Inspector Followup Item
IR
Inspection Report
LCO
Limiting Condition for Operation
LER
Licensee Event Report
Motor Generator
MDNBR
Minimum Departure from Nucleate-Boiling Ratio
MWt
Megawatts Thermal
NRC
Nuclear Regulatory Commission
Nuclear Reactor Regulation
Onsite Nuclear Safety
OST
Operations Surveillance Test
OP delta T
Overpower Delta Temperature
OT delta T
Overtemperature Delta Temperature
9
REV
Revision
Resistant Temperature Detector
Reactor Turbine Gauge Board
T
Temperature
TI
Training Instruction
TS
Technical Specification
Unresolved Item*
W/R
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.