ML14178A256
| ML14178A256 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 07/15/1992 |
| From: | Hunt M, Jape F, Casey Smith NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML14178A254 | List: |
| References | |
| 50-261-92-19, GL-89-10, NUDOCS 9208110103 | |
| Download: ML14178A256 (13) | |
See also: IR 05000261/1992019
Text
REG(
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTA STREET, N.W.
ATLANTA, GEORGIA 30323
Report No.: 50-261/92-19
Licensee: Carolina Power and Light Company
P. O. Box 1551
Raleigh, NC 27602
Docket No.: 50-261
License No.: DPR-23
Facility Name: H. B. Robinson
Inspection Conducted: June 15-19, 1992
Inspectors:
SM. Hunt
Date Signed
Approved by:
/9-
2
F. Jape, Chief
Date Signed
Test Programs Section
Engineering Branch
Division of Reactor Safety
SUMMARY
Scope:
This special, announced inspection was conducted to examine the licensee's
program for self-assessment of problems, followup of a previous inspection item,
and to review results of testing as related to Generic Letter 89-10.
Results:
The requirements of T.S. Section 6.5.1.6, concerning oversight activities of the
PNSC with regard to reviews of LERs, ACRs and SCRs, were verified as having
been satisfied. The licensee performs root cause analysis of LERs, ACRs, and
SCRs using investigative techniques such as Change Analysis, Barrier Analysis, and
Events and Causal Factors Charts. These root cause analyses were found to be
.
generally acceptable with a few exceptions. The developed corrective action plans
were consistent with the identified root causes and the corrective
9206110103 920722
ADOCK OBOOO261
0
2
action program monitored implementation of the corrective actions to assure
completion. A previously identified item related to GL 89-10, NRC Report No. 50
261/91-201 was closed. The review of the calculation for MOV FW-V2-6A,
related to the Generic letter 89-10 MOV Program identified one violation, 50
261/92-19-01, of Criterion III to Appendix B 10 CFR 50. (paragraph 4a.)
REPORT DETAILS
1.
Persons Contacted
Licensee Employees
- R. Barnett, Manager, Outages and Modification
- C. Baucom, Project Specialist, Regulatory Compliance
- W. Biggs, Manager, Nuclear Engineering Department Site Unit
- S. Billings, Technical Aide, Regulatory Compliance
- R. Chambers, Plant General Manager
- S. Farmer, Manager, Engineering Programs
- W. Gainey, Jr., Manager, Plant Support
- M. Grantham, Nuclear Engineering Department, HESS/Mechanical
- J. Harrison, Manager, Regulatory Compliance
- P. Musser, Manger, Engineering and Technical Support, Nuclear
Assurance Department
- J. Pearson, Nuclear Engineering Department, HESS/Mechanical
- D. Stadler, Onsite Licensing Engineer/Nuclear Licensing
NRC Resident Inspectors
- L. Garner, Senior Resident Inspector
- C. Ogle, Resident Inspector
- Attended exit interview
2.
Action on Previous Inspection Findings (92702)
Closed, VIO, 50-261/91-201
During an inspection of the licensee's Generic Letter (GL) 89-10, "Safety
Related Motor-Operated Valve Testing and Surveillance Program," a violation
was identified. Report No. 50-261/91-201, was issued on July 25, 1991,
and the notice of violation was forwarded in a letter issued October 4,
1991. The violation concerned the lack of documentation of corrective
actions taken for MOV FW-V2-6A, Feedwater Isolation Valve, which had a
galled valve stem. As a result of the violation the licensee has revised its
corrective action program to require complete documentation. In addition
the operations procedure for shutdown of the unit was revised to require
cycling of the three feedwater block valves during cool down to prevent
thermal binding of these valves which was determined to be a contributor to
the opening difficulties on valve FW-V2-6A. Additionally, during the recent
outage the valve stem was replaced. This violation is closed.
2
3.
Evaluation of Licensee Self-Assessment Capability (40500)
T. S. Section 6.5.1.6 describes the responsibilities of the PNSC and
specifies reviews to be performed by this onsite organization. Violations of
the TS are required to be investigated and a report prepared that evaluates
the event and provides recommendations to prevent recurrence. The PNSC
reviews these reports and evaluates the adequacy of the developed
corrective action plan. Additionally, station problems documented as ACRs,
or SCRS, are reviewed by the PNSC to ensure that an adequate root cause
analysis has been performed and an effective corrective action plan has been
developed. The inspectors performed an evaluation of the licensee's
self-assessment capability by reviewing monthly PNSC meeting minutes
covering a period from March 15, 1991 to March 20, 1992. Selected action
items dispositioned by the PNSC at these meetings were independently
reviewed using the investigative techniques delineated in the "NRC HPIP
Procedure and Module Manual."
a.
Licensee Event Reports
The inspectors reviewed the events and their root causes documented
on the following LERs and evaluated the proposed corrective actions
to determine their adequacy.
o
LER No.91-004, Rod Control System Urgent Failure
o
LER No.91-007, Failure to Perform Surveillance Test
0
LER No.91-009, Over Temperature Delta Temperature Channel
Inoperable due to Summator Module Lag Constants
0
LER No.91-013, Diesel Driven Fire Pump inoperable
0
LER No.92-002, Failure to Test all Circuits Associated with
the Auxiliary Feedwater Auto-Start
0
LER No.92-004, T.S. Violation During ILRT
The root causes for these plant events ranged from random hardware
failure to human errors and procedural deficiencies. The inspectors
determined that the root causes identified by the licensee were for the
most part correct with exception of the following examples. The root
cause for the event documented on LER 91-007 was given as human
error. Application of the guidance delineated in the NRC HPIP manual
identified the root cause to be inadequate communications which
3
resulted in less than adequate shift-turnover. The licensee's
developed corrective plan was considered adequate, however, in that
administrative controls were established to ensure adequate communi
cation and work control during shift changes. Additionally, plant
personnel were indoctrinated on the use of the new administrative
controls. Another example of inadequate root cause analysis was
identified on LER 91-013. The licensee identified the root cause as
failure of the design engineering program to replace existing 1000
120OF thermostats with 120 0-140aF thermostats. The inspectors
discovered, however, that the primary causal factor was failure of the
CAP to initiate corrective action for a station problem that was
identified in October 1989 and which was documented on WR/JO
89-AJMF1. Discussions with licensee's personnel revealed that the
CAP was in a process of transition at the time the deficiency was
identified. This probably was the reason why it failed to initiate
corrective action for an identified and documented station problem. A
definitive evaluation of this root cause can not be made, however,
because of its indeterminate status. The licensee's implemented
corrective action for LER 91-013 was considered adequate based on
change out of the thermostats and an increase in the power rating of
the associated heater.
b.
Adverse Condition Reports/Significant Condition Reports
ACRs and SCRs dispositioned by the PNSC during regular monthly
meetings were independently reviewed by the inspectors for root
causes in order to evaluate the licensees self-assessment capability.
Objective evidence reviewed during this effort are listed as follows:
ACR No.91-009
ACR No.91-034
ACR No.91-283
ACR No.91-286
ACR No. 92-20
4
The inspectors determined that the licensee used various investigative
techniques during the performance of root cause analyses. Among
these were Barrier Analysis, Changes Analysis, and Events and Causal
Factor Charts. Based on review of the above ACRs/SCRs the
inspectors concluded that the root causes identified by the licensee
were generally correct. ACR No.91-286 was a typical example and
involved failure of the narrow range OTDT RTD instrumentation circuit
to meet TS requirement of 0.75 seconds time delay. This event was
also reported to the NRC on LER 91-009-01. The licensee correctly
identified the root causes documented on ACR No.91-286.
Additionally, the inspectors reviewed the close-out package and
verified that the developed corrective action plans for the three
primary causal factors specified in the LER had been completed in
accordance with the Licensee's commitments. SCR No.88-022
further demonstrated the licensee's capability to perform effective
self-assessments. This SCR involved an event wherein the reactor
vessel cavity boron concentration was inadvertently diluted to less
than 1950 ppm during a RFO with fuel off loaded. The inspectors
used the guidance of the NRC HPIP Module and determined the near
root causes to be in the functional areas of training, procedures,
supervision, and communications. The licensee's developed
corrective action plans for the seven causal factors identified in the
Event and Causal Factor Chart fell within the functional areas
identified by the NRC HPIP module.
Some inadequate root cause analyses were identified by the
inspectors. Typical of this small sample was ACR No.91-034. This
ACR involved an event related to inadequately revised calibration
procedures required per plant modification M-959. Licensee
management determined the root cause to be indeterminate. The
inspectors, however, identified the near root causes as inadequate
design control. Specifically, the postmodification/calibration test
requirements and test acceptance criteria were not adequately
specified in the plant modification package. The immediate corrective
action of revising the loop calibration procedures to be technically
adequate was necessary but not sufficient to prevent recurrence of a
similar problem. Discussions with the senior resident inspector
revealed that the licensee's response to an NOV involving a civil
penalty more adequately addressed the required corrective actions for
the deficiency documented on ACR 91-034.
The inspectors also selected ACR No.92-186 for review of problem
assessment activities. The ACR was written to identify a tripping
condition which occurred while the A Emergency Diesel Generator
(EDG) was undergoing an over-speed trip test. A similar condition had
occurred on a B EDG while it was undergoing post maintenance
overspeed trip testing. In each instance, the same condition, fuel rack
unlatched, was found. The A EDG was operational at the time the
trip occurred, while the B EDG was still in post maintenance test
status.
As a result of the EDG A trip, an investigation team was organized to
determine the cause of the fuel rack unlatched condition. The team
consisted of knowledgeable engineering personnel, an operations
person and a maintenance supervisor and craftsman. Additional
personnel provided special assistance when needed.
The team conducted root causes analysis which included a description
of the event, equipment failure/conditions affecting the event, a
chronological description, and summarized the factors that influenced
human behavior. A list of proposed corrective actions to preclude
recurrence was prepared. Corrective actions for contributing factors
and improvements based on the investigation were recommended.
The team used various causal factors check sheets to assess each
aspect of the event. These check sheets were applicable to any
investigation and contained a group of questions that could be applied
to any situation.
The inspectors concluded that the investigation was thorough and the
method of reaching the solution was acceptable. The recommendation
seemed to fit the findings of the investigating team.
c.
Conclusion
The inspectors concluded that the licensee management generally
performed an adequate root cause analysis for LERs. ACRs and SCRs.
The developed corrective action plans were also consistent with the
identified root causes to ensure implementation of effective corrective
actions. Additionally, the corrective action program monitors
implemented corrective action plans to verify completion of corrective
actions. The inspectors attended the PNSC monthly meeting on
June 17, 1992 to observe the depth of review of overall plant
performance. The meeting was well conducted with a prepared
agenda. The agenda items were presented to the committee members
in a clear and understandable manner, and the committee reviews
were thorough and in depth.
6
Within this area no violations or deviations were identified.
4.
Generic Letter 89-10. Safety-related Motor-Operated Valve Testing and
Surveillance (TI 2515/109)
The inspectors performed a limited review of the licensee's Motor Operated
Valve (MOV) program.
a.
Differential Pressure Testing (DP)
The testing of MOVs, either static or under DP conditions is performed
using VOTES diagnostic equipment. The acceptance criteria is
furnished to the testing personnel by the licensee's Nuclear
Engineering Department (NED). The traces produced by the
diagnostic equipment are screened by on-site personnel to verify the
required thrust is developed and is within the thrust window. The
traces are also reviewed for acceptable motor current, packing load,
and verification that the maximum thrust at torque switch trip (TST) is
below the maximum allowable thrust. Once these items are verified,
the trace data is forwarded to NED for detailed analysis. The
inspectors were advised that if during the NED review a discrepancy is
found, the site is notified and corrective action is initiated.
The inspectors reviewed the test traces with the licensee
representatives for the following valves:
Thrust at
Thrust
Valve ID
Thrust Range
Flow cutoff
at TST
Number
LBS
LBS
LBS
CC-735
8,883 - 12600
833
9359
CC-730
3351 - 12600
3427
5775
CC716B
3351 - 12600
3560
4236
RHR-744B
6669.6 - 126000
4461
9693
RHR-744A
6669.6 - 126000
6020
23,422
SI-870B
8099 - 12600
2552
8571
SI-870A
8099 - 12600
5616
9595
SI-864B
12856 - 21600
Static only
19085
CVC-350
1482 - 7200
Static only
. 2239
FW-V2-6B
38211 - 63000
Static only
44863
FW-V2-6C
38211 - 63000
Static only
38573
FW-V2-6A
38211 - 63000
Static only
43010
No problems were identified with these tests.
7
The inspectors reviewed the status of feedwater MOV FW-V2-6A to
examine the corrective actions taken to improve the operability of the
valve. The licensee had made changes to improve the operation of
the valve in the open direction. The TOLs had been tripping during
the opening stroke of FW-V2-6A. The tripping was determined to be
the result of thermal binding of the valve. Thermal binding occurs
when the valve is closed while at high temperature and allowed to
cool in the closed position which causes the seats to tighten on the
wedge gate disc. Then on the next attempt to open the valve after
cool down, a high thrust is required to unseat the valve. This high
thrust requirement causes the valve actuator motor to stall, causing
the TOLs to trip.
On June 15, 1991, an operability review determinated that thermal
binding was occurring. In January 1992, the licensee performed
calculations Nos. RNP-M/MECH-1398, 1399 and 1400 which
recommended a lighter spring pack to stay within the torque rating of
the actuators based on the postulated accident differential pressures
of 50 psid. The operating procedures were revised to require cycling
of the feedwater valves FW-2V-6A, B and C, during unit cool down.
Lighter spring packs were installed during the June 1992 outage in
the valve actuator of each valve to reduce the thrust at the end of the
close cycle. The lighter spring packs were installed as the result of
the calculated differential pressure across the valves of 50 psid. This
value was based on the postulated assumption that feedwater
regulating valves located down stream of each block valve will close
in seven seconds after a safety injection (SI) signal is received and the
reactor feedwater pump trips and coasts to a stop. The design basis
differential pressure report DP-027FW for the motor operated valves
(MOVs) in the feedwater system for the Robinson Nuclear Plant
acknowledged that if the feedwater regulating valves were in the
manual mode at the time the SI signal was received, the block valve
would see some, "substantial though indeterminate AP." This report
assumed that the flow control valve would close and cause minimum
flow across the block valve, but in any case did mention that the line
pressure is assumed to be 580 psig during accident conditions with
the feedwater regulating valves in the manual position. The
assumption held to in the evaluation is that the feedwater regulating
valves will always be closed first.
8
On June 15, 1992, Feedwater Regulating Valve FCV-478 was given a
close command but did not close sufficiently to reduce the differential
pressure across Block Valve FW-2V-6A. FW-2V-6A torqued out
before completely closing off the flow to Steam Generator A. Work
request WR/JO 92-AJEH2 was written to check the stroke and adjust
the positioner of FCV-478. The inspector inquired about the condition
of FW-V2-6A and why it did not close fully. The reason given was
the discharge pressure of the condensate pump was greater than the
50 psid. The inspector then questioned the licensee concerning the
basis for the assumption that the reactor feedwater pump tripping
would cause the DP across-the block valve to be less than 50 psid.
It appears that an unverified assumption was made by the licensee
that the condensate pump is also tripped when a safety injection
signal is received. The condensate pump trip is a manual action taken
by the operator, and is not initiated by an automatic trip signal.
The H. B. Robinson FSAR Table 15.1.5.2, ACTUATION SIGNALS
AND DELAYS FOR MSIV, SIS AND FEEDWATER SAFETY ACTIONS,
states that the main feedwater regulating valve closure occurs seven
seconds after the SI signal. When the licensee recalculated the
differential across the valve with the condensate pump still operating
the DP was calculated to be 480 psid at closing and 375 psid at
opening. The licensee immediately issued the necessary work
requests to reset the torque switches on these three valves to enable
the actuator to develop the required thrust without tripping the torque
switch before closure is accomplished.
The failure on the part of the licensee to consider the DP across the
three feedwater valves with the condensate pumps operating, and
setting the feedwater block valves to close at a pressure less than
actual is identified as violation 50-261/92-19-01: Inadequate design
control involving unverified assumptions related to D/P for Valves FW
2V-6A,B, and C.
10 CFR50, Appendix B, Criterion III states in part "... design control
measures shall provide for verifying or checking the adequacy of
designs ....
Contrary to the above, on June 15, 1992, Feedwater
Block Valve FW-2V-6A did not fully close due to the differential
pressure across-the valve having been calculated at a lower value than
existed in the system. The differential pressure had been calcualted
to be 50 psid across each of the three feedwater block valves and the
valves had been adjusted for closure at that pressure. Upon inquiry
by the NRC Inspectors, the licensee recalculated the differential
pressure to be 480 psid. The differential pressure was the result of
9
the condensate pump operating, which was assumed to trip upon
receipt of a safety injection signal allowing the valves to close under
the lower diffferential pressure.
b.
Schedule
The licensee was returning the unit to operation after the completion
of a refueling outage. The MOV testing scheduled during the outage
was completed as planned. The licensee differential pressure tested
23 MOVs and static tested 42 MOVs. Other planned maintenance
items such as the installation of VOTES sensors, electrical and
mechanical preventive maintenance, and the replacement of torque
switches was accomplished during the outage as scheduled, with the
exception of 4 torque switch replacements which were delayed due to
parts availability.
c.
Maintenance
The licensee has developed procedure TMM-032, TECHNICAL
SUPPORT MANAGEMENT MANUAL PROCEDURE; MOTOR
OPERATED VALVE PROGRAM, for the purpose of establishing,
implementing, and maintaining an overall program for motor operated
valves. This procedure references various maintenance documents as
guidelines for maintaining MOVS. The inspector reviewed a draft
revision of MMM-003, Appendix A, POST MAINTENANCE TESTING,
which defines the post maintenance testing required at the completion
of various MOV maintenance activities. The control of switch
settings is maintained under procedure CM-1 11, LIMITORQUE LIMIT
SWITCH AND TORQUE SWITCH MAINTENANCE. This appears to be
an adequate procedural control for accomplishing the MOV GL 89-10
program.
d.
Training
The licensee has a group of craftsmen that travel between the nuclear
plants and perform the testing of MOVs during outages. The test
data taken are initially reviewed by the site personnel and later by the
NED. Licensee representatives informed the inspector that some of
-the site instrumentation and control (I&C) personnel had been trained
in the use of the diagnostic equipment, but due to the heavy work
load during the recent outage, they were not involved in the testing
completed this outage. The licensee has scheduled classes for
training in the use and analyzing of the Votes equipment and traces
for selected site personnel. The inspector noted that a MOV training
10
for selected site personnel. The inspector noted that a MOV training
class was in session during this inspection.
5.
Exit Interview
The inspection scope and results were summarized on June 19, 1992, with
those persons indicated in paragraph 1. The inspectors described the areas
inspected and discussed in detail the inspection results listed below.
Proprietary information is not contained in this report. The violation 50
261/92-19-01, Inadequate design control involving unverified assumptions
related to DIP for Valves FW-2V-6A, B, and C, was discussed and no
dissenting comments were received.
Acronyms and Initialisms
ACR
Adverse Condition Report
Corrective Action Program
DP
Differential Pressure
GL
Generic Letter
Human Performance Investigation Process
Instrumentation and Control
LER
Licensee Event Report
Motor Operated Valve
NRC
Nuclear Regulatory Commission
Over-temperature delta-temperature
Parts Per Million
PNSC
Plant Nuclear Safety Committee
psid
Pounds Per Square Inch Differential
Refueling Outage
Resistance Temperature Device
Safety Injection
Significant Condition Report
TS
Technical Specification
Torque Switch Trip
Thermal Overload Limit
VOTES
Valve Operation Test & Evaluation System
WR/JO
Work Request/Job Order