ML14181A997
| ML14181A997 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 03/12/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML14181A994 | List: |
| References | |
| 50-261-98-01, 50-261-98-1, NUDOCS 9803230024 | |
| Download: ML14181A997 (32) | |
See also: IR 05000261/1998001
Text
U.S. NUCLEAR REGULATORY COMMISSION
REGION II
Docket No:
50-261
License No:
Report No:
50-261/98-01
Licensee:
Carolina Power & Light (CP&L)
Facility:
H. B. Robinson Unit 2
Location:
3581 West Entrance Road
Hartsville, SC 29550
Dates:
January 4 - February 14, 1998
Inspectors:
B. Desai, Senior Resident Inspector
E. Girard, Region II Inspector
(Section E1.2)
M. Holbrook, NRC Contractor (Section E1.2)
J. Coley, Region II Inspector
(Section M3.1)
J. Lenahan. Region II Inspector
(Section E8.1)
M. Miller, Region II Inspector
(Section E8.1)
F. Jape, Region II Inspector
(Section E1.1)
Accompanying Personnel: T. Scarborough, NRR
Approved by:
M. Shymlock. Chief, Projects Branch 4
Division of Reactor Projects
Enclosure 2
9803230024 980312
ADOCK 05000261
G
EXECUTIVE SUMMARY
H. B. Robinson Power Plant, Unit 2
NRC Integrated Inspection Report 50-261/98-01
This integrated inspection included aspects of licensee operations,
maintenance, engineering, and plant support. The report covers a six-week
period of resident inspection; in addition, it includes the results of
inspections by Region II based reactor safety inspectors, including those
related to the Motor Operated Valve (MOV) program as well as Architectural
Engineering (A/E) followup inspection.
Operations
The conduct of operations was professional, risk informed, and safety
conscious (Section 01.1).
Portions of the Service Water (SW) System and the Control Room Emergency
Filtration (CREFS) System were walked down and no negative observations
were noted (Section 01.2).
Activities related to control of water level of the Ultimate Heat Sink
(Lake Robinson) were appropriately conducted and met Technical
Specification (TS) and UFASR requirements (Section 01.3).
Nuclear Assessment Section and Plant Nuclear Safety Committee continued
to provide strong oversight (Section 07.1).
Maintenance
Maintenance and surveillance activities were performed satisfactorily.
The inspector noted good controls of housekeeping and good supervisor
oversight of work activities (Section 1M1.1 and M2.2).
The process followed by the licensee in handling a pin hole leak in
Component Cooling Water Line 10-AC-41 until a code repair could be
performed was correct and thorough (Section M2.1).
Changes made to maintenance and operation procedures which implemented
the more restrictive requirements delineated in the new Improved
Technical Specifications were found to be effective and thorough
(Section M.3).
Engineering
The inspector concluded that the modification program including
procedures, records, and post modification testing practices to be
technically and administratively adequate (Section E1.2).
Based on the NRC inspections of the licensee's implementation of
GL 89-10 and on the licensee's commitments in its letter dated
February 20, 1998, the NRC is closing its review of the GL 89-10 program
at Robinson (Section E1.2).
.2
The I&C technician exhibited good questioning attitude when he noticed
that the Ronan I/P had behaved differently than what he had remembered
from a previous conversation with the responsible engineer. A non-cited
violation was identified for not adequately considering the failure
modes associated with the installation of new Ronan I/P.
Upon
identification, the licensee immediately stopped the installation
activity and reconsidered the merits of the ESR as written (Section
E3.1).
A violation was identified with three examples of failure to update the
UFSAR in accordance with 10 CFR 50.71 (e). A non-cited violation was
also identified for the six examples of licensee identified UFSAR
discrepancies discussed in IR 50-261/97-201 (Section E8.1).
Overall, the inspector noted that the predictive maintenance program
continued to be on a positive trend (Section E2.1).
Plant Support
The inspectors concluded that radiation control and security practices
were proper (Section R1.1 and S1.1).
Overall licensee performance in 1997 in the Radiological protection area
was excellent (Section R1.2).
Report Details
Summary of Plant Status
Robinson Unit 2 operated at full power for the entire report period without
any significant problems.
I. Operations
01
Conduct of Operations
01.1 General Comments (71707)
The inspectors conducted frequent control room tours to verify proper
staffing, operator attentiveness and communications, and adherence to
approved procedures. The inspectors attended daily operation turnovers,
management reviews, and plan-of-the-day meetings to maintain awareness
of overall plant operations. Operator logs were reviewed to verify
operational safety and compliance with Technical Specifications (TSs).
Instrumentation, computer indications, and safety system lineups were
periodically reviewed from the Control Room to assess operability.
Frequent plant tours were conducted to observe equipment status and
housekeeping. Condition Reports (CRs) were routinely reviewed to assure
that potential safety concerns and equipment problems were reported and
resolved. Good plant equipment material conditions and housekeeping
continued to be observed throughout the report period.
In general, the conduct of operations was risk informed, professional,
and safety-conscious.
01.2 System Walkdown
a. Inspection Scope (71707)
The inspector walked down portions of the Service Water (SW) and the
Control Room Emergency Filtration System (CREFS).
b. Observations and Findings
The portions of the SW system walkdown included the intake. Emergency
Diesel Generator (EDG) supplies, CREFS supplies, a sample of the turbine
building loads, and the Component Cooling Water (CCW) supply. For the
portions walked down, the inspector did not observe any abnormalities.
The inspector did note that at the intake, the licensee had installed a
temporary portable sump pump to periodically pump down the manhole that
houses certain SW pump/valve cables. Due to rain water buildup in the
manhole, the licensee had noted lower than expected normal cable
insulation resistance. The sump pump was installed using a work request
(WR).
The inspector questioned the licensee if this should have been
installed as a temporary modification. The system engineer provided a
list of attributes associ.ated with the pump installation justifying that
a temporary modification was not warranted. The inspector reviewed the
requirements for temporary modification and concluded that licensee
actions were appropriate.
2 .
c. Conclusions
The inspector concluded that for the portions walked down, the SW and
the CREFS system were appropriately aligned. No negative observations
were noted.
01.3 Ultimate Heat Sink
a. Inspection Scope (71707)
The .inspector reviewed licensee actions related to heavy rainfall and
its affect on the Ultimate Heat Sink (UHS).
b. Observations and Findings
Recent heavy rainfall in the area has caused levels in Lake Robinson to
be higher than normal.
Lake Robinson is the UHS for the Robinson
Nuclear Plant. TS 3.7.8 requires a lake level of greater than 218 feet.
Normally, the lake level is maintained at approximately 221 feet. With
the recent heavy rainfall in the area, the lake level had risen above
the normal level.
The licensee controlled the lake level by releasing
water from the Robinson spilling way lake level to below 222 feet on
several occasions (ref. UFSAR section 2.4). The control of the lake
level is performed by the fossil unit (Unit 1).
The inspector monitored
this activity on one occasion and did not identify any problems
associated with the coordination and control of lake level to meet TS
and design requirements. Notwithstanding, the licensee is reviewing
possible ways to better coordinate control of the UHS. This includes
the possibility of having a Unit 2 verses Unit 1 procedure to control
and coordinate maintenance of the lake level.
c. Conclusion
The inspector concluded that the licensee appropriately controlled
activities to maintain the water level in Lake Robinson within required
limits.
07
Quality Assurance In Operations
07.1 Plant Nuclear Safety Committee and Nuclear Assessment Section Oversight
a. Inspection Scope (40500)
The inspector evaluated certain activities of the Plant Nuclear Safety
Committee (PNSC) and Nuclear Assessment Section (NAS) to determine
whether the onsite review functions were conducted in accordance with TS
and other regulatory requirements.
b. Observations and Findings
The inspector periodically attended PNSC meetings during the report
period. The presentations were thorough and the presenters readily
responded to all questions. The committee members asked probing
questions and were well prepared. The committee members displayed
understanding of the issues and potential risks.
Further, the inspector
reviewed NAS audits and concluded that they were appropriately focused
to identify and enhance safety.
c. Conclusions
The inspector concluded that the onsite review functions of the PNSC
were conducted in accordance with TSs. The PNSC meetings attended by
the inspector were well coordinated and meetings topics were thoroughly
discussed and evaluated. NAS continued to provide strong oversight of
licensee activities.
08
Miscellaneous Operations Issues (92901, 92702)
08.1 (Closed) Violation 50-261/97-06-01, Inadequate safeguards procedures
that allowed ESF train being out-of-service without invoking a Technical
Specification action statement: The corrective actions presented in the
licensee's response, dated June 24, 1997, were reviewed and verified by
the inspector. The NRC accepted the response by letter, dated June 26,
1997. This violation is closed.
II. Maintenance
M1
Conduct of Maintenance
M1.1 General Comments
a. Inspection Scope (61726 and 62707)
The inspector reviewed/observed all or portions of the following
maintenance related work requests/job orders (WRs/JOs) and/or
surveillances and reviewed the associated documentation:
WR/JO 97-AEIR1 and PM-420, Votes Testing
OST 401-1, EDG "A"
Slow Speed Start
WR/JO AAQM-001, CCW Heat Exchanger Cleaning
OST-302-1, Service Water System Component Test
WR/JO AAFI1, Replace Installed Fisher I/P with Ronan I/P
EST-124, Response Time Testing of Reactor Coolant System RTDs
b. Observations and Findings
The inspector observed that these activities were performed by personnel
Procedures were present at the work location and being followed.
4
Procedures provided sufficient detail and guidance for the intended
activities. Activities were properly authorized and coordinated with
operations prior to starting. Test and maintenance equipment in use was
properly calibrated, procedure prerequisites were met, and system
restoration was completed.
c. Conclusions
The inspector concluded that routine and corrective maintenance and
surveillance activities were performed satisfactorily.
M2
Maintenance and Material Condition of Facilities and Equipment
M2.1 Review of the Through Wall Flaw Evaluation and Operability Determination
Processes for a Pin Hole Leak in Component Cooling Water (CCW) Line
10-AC-41
a. Inspection Scope(62700)
The inspector observed.a reported pin hole leak in the moderate energy
CCW Line 10-AC-41 and reviewed documentation to determine if actions
taken by the licensee were in accordance with structural integrity
guidance given in NRC Generic Letter (GL) 90-05 for ASME Class 3 piping,
and guidance given for resolution of degraded and nonconforming
conditions and continued operability in NRC GL 91-18. Since this pipe
section could be isolated, code repair was successfully performed during
the week of February 9-13, 1998.
b. Observation and findings
The pin hole leak in the moderate energy CCW pipe 10-AC-41 was between
Valves CC-775 and CC-776. The subject pipe was the return pipe from the
Spent Fuel Pool Heat Exchanger. The thinned area was located downstream
of Valve CC-775, one-inch beyond the weld of the weld neck flange which
supports the valve. The circumferential position of the thinned area on
the pipe coincided with downstream edge of the butterfly valve disk and
was approximately centered at the 12 o'clock pipe position. The valve
was installed with the disk shaft horizonal. The positioning of the
valve make this location susceptible to erosion and would be exacerbated
by the increased turbulence resulting from throttling flow.
The inspector held discussions with cognizant engineers, reviewed the
ultrasonic test data for the remaining wall thickness of the affected
area on the 10-inch schedule 40 carbon steel pipe; verified that the
evaluation/calculation used to evaluate the structural integrity of CCW
Line 10-AC-41 with the through-wall flaw was acceptable based on the
criteria and methodology specified in GL 90-05 for ASME Code Class 3
moderate energy lines: and reviewed the operability determination (ESR
9800050) including the enclosed 10 CFR 50.59 Safety Evaluation
Unreviewed Safety Question Determination form. As a result of these.
reviews, the inspector concluded that the licensee's actions met.the
guidance given in GL 90-05 and GL 91-18 and the pin hole through-wall
5
flaw in CCW Line 10-AC-41, while undesirable, did not reduce the safety
margin for this system or endanger other equipment in the immediate
area.
c. Conclusion
The process followed by the licensee in handling the pin hole leak in
Component Cooling Water Line 10-AC-41 was correct and thorough. A code
repair of the leak was successfully performed.
M2.2 Containment Personnel Airlock Troubleshooting and Semiannual Airlock
Leakage Test
a. Inspection Scope (62707,61726)
The inspector observed portions of Special Procedure, SP-1418,
Containment Personnel Airlock Leakage Troubleshooting, and Engineering
Surveillance Test, EST-010, Containment Personnel Airlock Leakage Test
(semiannual). These tests were performed to comply with 10 CFR 50,
Appendix J, Option A, for Type B testing, and to fulfill the
requirements of TS Surveillance Requirement (SR) 3.6.2 for the
containment personnel airlock.
b. Observations and Findings
Previous airlock test results indicated higher than normal leakage
through the inner door of the personnel air lock. The licensee prepared
SP-1418 to provide a troubleshooting method for identifying the leakage
pathway for the personnel air lock. The inspector reviewed this
procedure and found it to be adequate. The technique utilized was to
pressurize the space between the inner and outer doors to about 20
pounds per square inch gauge (psig), and using a device called
Ultraprobe, detect the leakage location. The Ultraprobe is an
ultrasonic inspection system which detects noise due to the air leakage.
A briefing in the control room was held prior to beginning the
troubleshooting with all personnel involved. The inspector attended the
briefing and found it to be detailed and beneficial. Questions were
raised regarding personnel safety, Health Physics issues. TS compliance
and the test technique. These were all satisfactorily answered. The
test engineer was knowledgeable and very familiar with the
troubleshooting technique and the regulatory requirements.
Following the briefing, workmen proceeded to the job site. The
inspector witnessed the activities at the air lock. The craft personnel
were knowledgeable and experienced for their assigned tasks. The work
proceeded according to the approved procedure without exception.
Containment integrity was maintained throughout the test and no TS
action statements were entered. Communications were maintained between
0
the job site and the control room. In particular, good communications
6
was maintained between the two workmen inside the containment vessel and
those outside.
The licensee concluded that the leakage was located at a seal on the
inner door opening-closing mechanism and was within allowable limits.
Repairs have been scheduled for the upcoming refueling outage.
Following completion of the troubleshooting test, activities were
directed to performance of EST-010. The inspector reviewed EST-010, and
noted that the procedure had been updated with regard to the new
improved TS. This test procedure provided a method to satisfy the
testing requirement of 10 CFR 50 Appendix J, for Type B testing. This
test was required to be completed following maintenance on the air lock
and at least once every six months. Therefore, performing the test on
this date satisfied both requirements.
Discussion of the test was included in the briefing for the
troubleshooting activity discussed above. The inspector witnessed the
following portions of the test:
Installation and removal of the strong back on the inner door.
Achievement of the test pressure of 46.5 psig.
Completion of the stabilization period.
Completion of the test and evaluation of test results.
The test was conducted as prescribed by the procedure. Containment
integrity as well as communication was maintained wi.th the control room
throughout the test period. Upon completion of the test, test results
were reviewed and the acceptance criteria were met. Throughout the test
period, test activities were coordinated with reactor operators and
supervision. Special tools and instrumentation were.noted to satisfy
the test conditions and accuracy requirements as stated within the test
procedure.
c. Conclusions
The inspector concluded that maintenance and surveillance activities
were performed satisfactorily. Coordination between the various crafts
was observed to be excellent. Throughout the job, Health Physics and
security personnel were available to assist in getting the job done
without problems. Work control practices were observed to be
satisfactory. Materials and backup help were available either at the
job site or by phone call.
The procedures and-work instructions were
well prepared, were easy to follow, and were understood by the workmen.
M3
Maintenance Procedures and Documentation
M3.1 Verification that the New Improved Technical Specifications were
Properly Implemented into Maintenance/Operation Procedures
a. Inspection Scope (Technical Instruction 2515/130)(62700)
.7
Robinson Unit 2 has been operating with TS issued with the original
operation license on July 31, 1970, as amended from time to time. By
letter dated August 27, 1996, as supplemented by letters dated
December 18, 1996, January 17, February 18, March 27, April 4,
April 25, April 29, May 30, June 2, June 13, August 8, September 10,
October 2. (RNP RA/97-0216), October 2 (RA/97-0207), October 10, and
October 21, 1997, the licensee proposed to amend Appendix A of Operating
License No. DPR-23 to completely revise the TS. The proposed .amendment
(No. 176) was based upon NUREG-1431, "Standard Technical Specifications
- Westinghouse Plants," Revision 1 dated April 1995, and upon guidance
in the "NRC Final Policy Statement on TS Improvements for Nuclear Power
Reactors" (Final Policy Statement), Published on July 22, 1993 (58 FR
39132). The overall objective-of the proposed amendment, consistent
with the Final Policy Statement, was to rewrite, reformat, and
streamline completely the existing TS for Robinson.
Robinson implemented their new Improved Technical Specifications (ITS)
on November 13, 1997. During inspections documented in this report, the
inspector reviewed the marked-up TS conversion documents, the ITS, the
TS Bases, the Technical Requirements Manual (TRM), the Off-Site Dose
Manual (OSDM), the Core Operating Limits Report (COLR), self-assessments
performed by the licensee, verified TS requirement relocations, reviewed
procedures used to control the relocation of TS requirements, verified
implementation of more restrictive ITS requirements in plant
implementing procedures, and verified the accuracy of the licensee's
computer tracking system. The focus of the inspection however, was to
determine if new more restrictive ITS requirements were properly
implemented into maintenance surveillance test (MST) procedures. This
review also resulted in many operations surveillance test (OST)
procedures being reviewed because the ITS requirements were applicable
to both maintenance and operations.
b. Observations and Findings
Two self-assessment audits had been performed by the licensee to
determine the readiness for implementation of the ITS. Assessment No.
LIC/PR-97-04, dated April 14, 1997, verified the adequacy of more
restrictive requirements implemented in OSTs. Findings identified by
this assessment were effective in improving the ITS implementation
process. The inspection conducted by the inspector ran parallel to this
self-assessment except the inspector focused primarily on maintenance
activities. A sample of TS requirements listed as relocated in the
licensee conversion submittal to NRC were verified and reloc'ated to the
specified location. Thirty-four more restrictive ITS requirements were
verified to be properly implemented into maintenance and operation
procedures.. Procedure additions were accomplished in accordance with
the requirements of Administrative Procedure No. AP-022, Revision 28,
"Document Change Procedure."
Ninety-three implementing procedures were
reviewed by the inspector in the verification process. The more
restrictive changes were chosen from Sections 3.3, 3.4, 3.7, 3.8 and 3.9
of the TS. Correct implementing procedures for 15 more restrictive TS
requirements were also verified in the licensee's surveillance tracking
8
and scheduling system to ensure the accuracy of this system. In
addition, action items for Corrective Action Reports involving
inadequate procedures dated from September 1, 1997, through
January 31, 1998, were reviewed to determine if any examples of improper
implementation of ITS requirements had been detected during the field
use of revised procedures. This review did not reveal any inadequate
procedure discrepancy resulting from the implementation of the new TS.
c. Conclusion
Changes made to maintenance and operation procedures which implemented
the more restrictive requirements delineated in the new Improved
Technical Specifications were found to be effective and thorough.
M8
Miscellaneous Maintenance Issues (92902)
M8.1 (Closed) Violation 50-261/97-09-02:
Failure to properly calibrate OPDT
channels. The inspector verified that the corrective actions described
in the licensee's response, dated October 24, 1997, and accepted by the
NRC on November 6, 1997 to be completed. This issue and the corrective
actions were presented in LER 50-261/97-07-00, which was closed in NRC
Inspection Report 50-261/97-09. Changes were made to the Loop
Calibration procedures and the event was reviewed with appropriate plant
personnel to prevent recurrence. This violation is closed.
III. Engineering
El
Conduct of Engineering
E1.1 Onsite Engineering-Design Changes and Modifications
a. Inspection Scope (37551)
The inspector reviewed the licensee's plans and activities related to
design changes and modifications for the upcoming refueling outage
(RFO-18). Three modification packages were reviewed to access:
Design Control of the original design basis,
Review process of the modification packages,
Interface controls, and
Quality of the design packages.
b. Observations and Findings
The three Engineering Service Request (ESR) selected for review were as
follows:
ESR 9600113, Differential Pressure Transmittal Reorientation,
9
ESR 9700366, Safety Injection Pump Net Positive Suction Head
(NPSH) Improvement, and
ESR 97-00671, Emergency Core Cooling System (ECCS) Sump Screen
Replacement.
Each of these.packages was prepared using the licensee's procedure EGR
NGGC-005, Engineering Service Request. This procedure provides detailed
instruction regarding all aspects of a design change or modification.
ESR-9600113
The objective of this .modification was to improve the task of filling
and venting transmitters identified in Condition Report (CR) 95-0262.
The modification concentrated on minimizing the potential for gas or air
accumulation in the sensing lines and transmitter diaphragm.
The original installations of these Rosemount transmitters were per the
manufactures recommendation, but venting proved to be difficult and time
consuming. A mock setup, at the facility, demonstrated that this change
would be more effective and took less time in completing the venting
process. The problem was recognized by the licensee for several years
and previous corrective actions were not fully effective. Therefore the
licensee was determined to complete this modification of reorienting the
transmitters such that they will easier vent air or gases.
The licensee issued LER 50-261/95-009-00 and -01 which discussed this
problem as it related to the cold leg accumulator level. Also, the NRC
issued violation 50-261/95-30-01.regarding this problem. These reports
have been previously discussed in inspection Report 50-261/97-04.
ESR 96-00113 contains detailed instruction for all transmitters to be
reoriented. A 10 CFR 50.59 safety evaluation was included in the
package, which concluded that no unreviewed safety question resulted as
a result of this change. The safety evaluation also determined that the
UFSAR would require updating as a result of this change. Specifically,
several valves, shown in Fig. 6.3.2-2 of the UFSAR will be removed per
this modification and new supports for some transmitters would be added.
These changes are scheduled for the next UFSAR change submittal.
The package was reviewed by all necessary disciplines. This included
ALARA considerations, electrical design considerations, environmental
qualification, maintenance, outage planning, operations and the system
engineers. The quality of the package was satisfactory for the stated
objective.
ESR 9700366
In inspection report 50-261/97-201 a concern with net positive suction
head (NPSH) requirements for the Safety Injection (SI) system was
identified. Plant engineering personnel promptly began an evaluation to
determine the ability of the SI system to fulfill its intended system
.
10
functions. Upon discovery that a potential discrepancy existed in the
NPSH design calculation, the refueling water storage tank water level
was increased from the original value. This action provided sufficient
NPSH but did not provide a comfortable margin. Investigations continued
and one of the corrective actions presented in LER 50-261/97-008-00 was
to modify the SI system piping to gain additional NPSH and a comfortable
margin.
ESR 97-00366 is an emergent project for implementation in Refueling
Outage, RFO-18 and is intended to increase the NPSH by reconfiguring the
suction piping. The new piping configuration has been determined to
provide about 4 feet of additional head pressure for the "B"
SI pump and
5 feet of additional head pressure for the "C"
SI pump. The "A"
SI pump
was considered to have acceptable NPSH.
The modification package provides detailed instruction for the piping
changes. The replacement and reconfiguration design, material, and
construction requirements applicable to the system have been identified
in the modification package. The package requires a pressure test in
accordance with inservice inspection requirements.
ESR-9600671
During refueling outage 17, (RFO-17), the emergency core cooling system
(ECCS) sump screen in the containment building were noted to be in a
degraded condition. Prior to plant restart the sump screens were
repaired. The licensee issued LER 50-261/96-005-00 and CR 96-02152 to
describe the issue.
This modification was to replace the existing carbon steel ECCS sump
screens with stainless steel material.
The replacement screens, as
described in the modification package, will meet or exceed the total
flow area and wire mesh size as the original screens.
The inspector reviewed the modification package and determined that it
met the functional requirements and complied with the design inputs of
the original installation. It was designated as a configuration change
(CC) per EGR-NGGC-0005, Engineering Service Requests. No category A
drawings are affected and no change to the UFSAR will be required by the
CC.
The package has been reviewed and approved by the appropriate
engineering disciplines, and appropriate plant programs, such as ALARA,
Maintenance Rule, Quality Control, Containment Coatings, Fire Protection
and Maintenance Procedures. Operations and mechanical maintenance have
reviewed and accepted this CC.
The modification package invokes PLP-047, Foreign Material Exclusion
Area (FMEA) Program. PLP-047 prescribes a foreign material exclusion
area, FMEA, to be established based on nuclear safety and risk. Three
levels of control are defined. ESR-00671 states that a Level 2 FMEA is
required for this CC. The FMEA is established prior to beginning any
011
work activity. The inspector reviewed the licensee's plans for this CC
and agreed that a Level 2 FMEA is appropriate. The final step in the CC
is to complete an inspection in accordance with EST-139, Containment
Sump Inspection. This inspection satisfies the post-modification test
requirement.
c. Conclusions
The inspector concluded that the modification program including
procedures, records, and post modification testing practices were
technically and administratively adequate. The three packages reviewed,
by the inspector, were found to contain all of the necessary tools and
controls to be properly implemented. Each of these jobs relies heavily
on the Responsible Engineer (RE) properly performing assigned duties.
The RE is the single point of cbntact and is accountable for all change
activities through closeout.
E1.2 Implementation of Generic Letter (GL) 89-10."Safety-Related Motor
Operated Valve Testing and Surveillance"
a. Inspection Scope (Temporary Instruction 2515/109)
This inspection assessed the licensee's implementation of GL 89-10,
which was previously determined inadequate during NRC Inspection
50-261/96-12. This inspection identified two violations (VIOs) which
addressed the principal deficiencies found in the licensee's
implementation of GL 89-10. The first of these violations, identified
as VIO 50-261/96-12-05, involved inadequately justified design
assumptions and the use of incorrectly determined stem rejection loads
in calculating opening valve factors. The second, VIO 50-261/96-12-06,
involved inadequate evaluations of test results relative to
calibrations, test conditions, and anomalous test data. The interim'
status of the licensee's actions to resolve the violations and complete
implementation of GL 89-10 was reviewed during Inspection 50-261/97-12,
which found that the progress toward correcting the violations was
satisfactory.. The current inspection further evaluated the licensee's
corrective actions for both of the violations. The inspection also
examined several other topics, including the setup of the licensee's
butterfly valves, MOV operability evaluations, and concerns identified
by Inspection 50-261/97-12.
The inspection was conducted through reviews of documentation and
interviews with licensee personnel. In assessing the resolution of
violations and concerns, the. inspectors focused on a sample of valves
selected from a tabulation of MOV test information, valve factors,
capability margins, etc., which the licensee had prepared for its
GL 89-10 valves. The valve sample was as follows:
12
AFW-V2-16A
Auxiliary Feed Water Header Discharge to Steam
Generator "A"
RHR-744A
Residual Heat Removal to Reactor Coolant Cold Leg
Isolation
RC-536
Pressurizer Power Operated Relief Valve (PORV) Block
FCV-626
Thermal Barrier Outlet Isolation
SI-869
Loops "B"
and "C"
Hot Leg Injection Shutoff
SI-845C
Containment Spray Additive Tank Discharge Throttle
V6-16A
Service Water North Header Supply to Turbine Building
V6-16B
Service Water South Header Supply to Turbine Building
The inspectors reviewed the test packages, calculations, and engineering
evaluations for the above MOVs. Other documents reviewed included:
Standard Procedure EGR-NGGC-0203, "Motor-Operated Valve
Performance Predication, Actuator Settings, and Diagnostic Test
Data Reconciliation," Revision 4
Standard Procedure EGR-NGGC-0101, "Electrical Calculation of Motor
Output Torque for AC and DC Motor Operated Valves (MOVs),"
Revision 2
ESR-9700330, "Determination of MOV Valve Factors," Revision 1
ESR-9700328, "Determination of MOV Rate-of-Loading Factors,"
Revision 1
ESR-9700331, "Determination of MOV Stem Factors," Revision 1
Altran Technical Report No. 97111-TR-01, "Summary Report of MOV
Differential Pressure Test Evaluations," Revision 1
Calculation RNP-M/MECH-1245, "Set Up Calculation for MOV FW-V2
6B," Revision 8
Calculation RNP-M/MECH 1452, "Evaluation of Static and Dynamic
Test Data for AFW-V2-16A," Revision 2
Calculation RNP-M/MECH-1237, "Set Up Calculation for MOV RHR
744A," Revision 3
Generic Letter 89-10 Site Improvement Plan, Revision 0
Capability Table (tabulation of MOV test information, valve
factors, capability margins, etc.), Revision 3
b. Observations and Findings
1. Justifications for Assumptions (VIO 50-261/96-12-05)
Violation 50-261/96-12-05 identified that the valve factors (VFs), rate
of loading (ROL), and stem friction coefficients (COFs) assumed in
setting and capability calculations had not been adequately justified.
The inspectors examined the licensee's corrective actions for the
violation during Inspection 50-261/97-12 and verified completion of
several corrective actions specified by the licensee. However, the
Engineering Service Requests (ESRs) being prepared to establish and
justify the VFs, ROL, and COFs and the related corrections to the
licensee's calculations had not been completed at the time and their
adequacy could not be confirmed. During the current inspection, the
inspectors verified that the ESRs and calculations were complete and
reviewed the ESRs and selected calculations to determine if the VFs,
13
ROL, and COFs now being assumed were adequately justified. The findings
of the reviews are described below.
Valve Factors (Established and Justified in ESR 9700330)
The inspectors found that the VFs established for the majority of valve
groups were adequately justified and incorporated into the calculations.
However, several concerns were identified:
The licensee had not been able to test two 1500# Copes-Vulcan
14-inch parallel double-disc gate valves (RHR-750 and 751) to
determine their VF.
Further, there was no industry test data
currently-available for use in establishing their VF. In the
absence of directly applicable test data, the licensee
evaluated Electric Power Research Institute (EPRI) data from
tests of valves with similar disc and seat contact surfaces.
The maximum VF determined using this data was 0.61. Rather
than apply this value, the licensee elected to use higher, more
conservative VFs in its calculations. These valve factors were
based on evaluation of Robinson's overall gate valve test
results. The maximum flow isolation VF of 0.66 was selected
for closing and the maximum opening VF of 0.73 was selected for
opening. The opening VF was further increased to 0.78 to
account for Bernoulli effects. The inspectors found that the
selected valve factors were reasonable in comparison to general
industry results. However, they were concerned that the values
selected were not based on actual testing of this valve design
or a closely similar design. The licensee identified this
concern for resolution in its Generic Letter 89-10 Site
Improvement Plan. The plan indicated that the licensee would
participate in further industry efforts to obtain applicable VF
data for this valve design.
Robinson's PORV block valves (RC-535 and RC-536) were 1500#
Westinghouse 3-inch flex-wedge gate valves. The licensee was
not able to dynamically test these valves but obtained test
results from Comanche Peak testing of similar valves. Based on
the Comanche Peak blowdown test results, the licensee selected
an open valve factor of 0.60 and a close valve factor of 0.68.
These values appeared reasonable to the inspectors, based on
values typically applied in the industry. However, the
inspectors expressed concern that the licensee had not compared
the internals of its valves to the Comanche Peak valves to
assure similar blowdown performance. The licensee identified
this concern for resolution in its Generic Letter 89-10 Site
Improvement Plan, which indicated that it would either use the
EPRI Performance Prediction Methodology (PPM) or would compare
the internal configuration of the Comanche Peak valves to the
Robinson valves to establish the validity of the block valves
VFs.
14
Robinson's GL 89-10 program included five Velan 2-inch globe
valves. The.licensee was not able to dynamically test any of
these valves and selected a VF of 1.10 based on the results of
tests that Turkey Point performed on similar valves. The
inspectors found that the Turkey Point tests were performed
with cold water and were concerned that the test data might not
be directly applicable to Robinson's valves, which had high
fluid temperature operating.requirements. The inspectors'
concern was limited, as the licensee's valves had large thrust
capability margins. The licensee identified this concern for
resolution in its Generic Letter 89-10 Site Improvement Plan,
which indicated that additional industry information would be
obtained to support the VFs used for these globe valves. In
addition, the plan stated that the documented basis for the VFs
would be revised accordingly.
Fire Protection valves (FP-248, 249, 256, and 258) were 900#
Anchor/Darling 4-inch flex-wedge gate valves. The licensee
applied a 0.80 closing VF to these valves in thrust
calculations. No in-plant test results were available to
support this VF. The licensee had evaluated EPRI's prototype
test results and selected data points from four similar
Anchor/Darling valves to support the VF. Use of individual
EPRI prototype test results generally is not acceptable to the
NRC. However, in
this instance, the inspectors' concern was
limited, as the VF selected both satisfactorily bounded the
EPRI results and was conservatively high as compared to overall
industry VF determinations. Additionally, the licensee's
valves had large thrust capability margins (exceeding 40%).
The licensee's Generic Letter 89-10 Site Improvement Plan
addressed this concern and specified further efforts to obtain
applicable industry data for these valves.
Rate.of Loading (Established and Justified in ESR 9700328)
The inspectors found that the licensee used a non-standard but
conservative method for determining the rate of loading. The licensee
statistically analyzed the rate of loading values obtained from the
in-plant testing of Robinson's gate valves and determined that the mean
value was 5.6% with 2 standard deviations equal to 23.8%. The
inspectors found that the values established were satisfactory.
The licensee was not able to dynamically test any of Robinson's globe
valves to determine a rate of loading value to use in thrust
calculations.
Instead, the licensee analyzed the results from a limited
number of globe valves tested by EPRI. Based on this analysis, the
licensee established a mean value of 12.4% and 2 standard deviations
equal to 18.6% for use in its globe valve thrust calculations. The
inspectors found that the applied values were reasonable. As the globe
valves had large available thrust margins, the inspectors did not
identify any significant concern regarding the rate of loading values
selected. The inspectors noted that longer-term the licensee may
receive test results from its own testing or industry testing that will
permit better justification of the rate of loading selected for these
The'inspectors verified that the licensee properly employed the rate of
loading values from its analyses in its thrust calculations.
Stem Friction Coefficient (Established and Justified in ESR 9700331)
The licensee's calculations typically assumed and applied a 0.20 stem
friction coefficient (COF) value if the stem COF was measured under
static conditions and a more conservative 0.22 value if the COF could
not be measured under -any conditions. In isolated cases measured values
exceeded the 0.20 assumption and the measured values were used. The
inspectors found that the values used were reasonable based on COFs
obtained throughout the industry; however, the licensee's test results
did not support these values with a high degree of statistical
confidence. ESR 9700331 attributed this to inaccuracies in the torque
wrench method it had used in determining output torque, which caused the
calculated stem friction coefficients to be unrealistically high. The
licensee supported this assumption with the results of a study conducted
at its Harris plant. In that study, output torque obtained by the
torque wrench method was compared to more precise direct torque
measurements. The study determined that the measured torque values
obtained with the torque wrench method were on average 9% higher than
the direct torque measurements for Limitorque SMB-00 actuators and 25%
higher for SMB-1 actuators. The inspectors expressed concern that the
licensee had no precise direct torque measurements on Robinson's MOVs to
assure the adequacy of the COFs assumed. Robinson personnel stated that
testing was planned during Robinson's Refueling Outage (RO) 18
(March 1998) to verify this assumption. Further, in a letter to the NRC
dated February 20, 1998. the licensee specifically committed to perform
testing during RO 18 to provide more precise static and dynamic stem
factors to support the stem coefficient of friction assumptions
contained in ESR 9700331.
2. Use of Incorrect Stem Rejection Loads (VIO 50-261/96-12-05)
In addition to using inadequately justified assumptions, Violation
50-261/96-12-05 identified that the stem rejection loads used in
calculating opening valve factors were incorrect. NRC inspectors
examined the licensee's corrective actions for the violation during
Inspection 50-261/97-12 and verified completion of several corrective
actions specified by the licensee. However, the licensee had not
completed the revisions to correct the calculations. During the current
inspection, the inspectors verified that these actions were complete.
This verification was based on documented closure of the actions in
Condition Reports 96-3178 and -3179 and review of the calculation
examples listed previously. In addition, the inspectors reviewed the
calculation for a subsequent dynamic test performed January 8, 1998, on
valve AFW-V2-14C and verified that the reconciliation calculation
16
correctly determined and utilized the stem rejection load in determining
the open valve factor.
Inadequate Evaluation of Test Results (VIO 50-261/96-12-06)
Violation 50-261/96-12-06 identified that the licensee had not
adequately evaluated test results to assure that test requirements had
been satisfied. The licensee had not adjusted valve opening thrust
measurements for test measurement calibration errors, had failed to
recognize that test data indicated that the test conditions in some
tests were not as intended. Additionally, the licensee failed to
resolve significant anomalies exhibited in some test data. The
inspectors examined the licensee's corrective actions for the violation
during Inspection 50-261/97-12 and verified completion of the corrective
actions for this violation, except correction of calculations and
additional training of MOV personnel in interpretation of test data.
During the current inspection, the inspectors verified that the
calculations were complete and that they satisfactorily addressed the
inadequately evaluated test results. The verification was performed
through a review of the following calculation examples:
Calculation RNP-M/MECH-1245, Revision 8, "Set Up Calculation
for MOV FW-V2-6B" (revision addressed originally incorrect test
differential pressure)
Calculation RNP-M/MECH 1452, Revision 2, "Evaluation of Static
and Dynamic Test Data for AFW-V2-16A" (revision corrected the
maximum open force selected from test)
Calculation.RNP-M/MECH-1237, Revision 3, "Set Up Calculation
for MOV RHR-744A" (revision corrected actuator efficiency and
provided appropriate open calibration error)
Calculation RNP-M/MECH-1517, Revision 3, "Set Up Calculation
for MOV MS-V1-8C" (revision and provided appropriate open
calibration error)
Calculation RNP-M/MECH-1446, Revision 3, "Set Up Calculation
for MOV RC-536" (revision provided appropriate open calibration
error)
With regard to the additional training of MOV personnel specified as a
corrective action, the inspectors verified that the training was still
scheduled to be performed during RFO-18 (March 1998), was tracked in the
licensee's database, and was shown to be still open.
Butterfly Valves
The licensee.only had three butterfly valves in its GL 89-10 program.
These were identical 16-inch, 150# pressure class, Allis-Chalmers
butterfly valves. The safety function of each was to close and closure
was controlled by torque switch settings with the valves torquing closed
17
into stopnuts. The current capabilities were marginal (0.4-to 16.9%
margin above the required torque). The licensee planned a control
scheme change from torque switch to position control for closing, which
would increase this margin to above 20%. The inspectors verified that
work requests 97-AEIU1, -AEIW1, and -AEIX1 provided for the change to
position control during RFO-18 scheduled in March 1998.
The inspectors reviewed the calculations for two of the three butterfly
valves (V6-16A and B) and found that the torque requirements were
calculated using an industry equation. Packing, seating, and bearing
torque loads were based on static and dynamic tests performed on V6-16B.
The licensee assumed that the hydrodynamic torque was negligible
however, -the dynamic test which the licensee performed to establish the
capabilities of the valves was at too low a flow to validate this. In a
letter to the NRC dated February 20, 1998, the licensee responded to
this issue and committed to perform calculations, tests, and/or
inspections to evaluate the hydrodynamic torque requirements for
butterfly valves V6-16A/B/C. This would establish the validity of the
licensee's assumption that hydrodynamic torque was negligible. The
letter indicated these actions would be complete by June 25, 1998.
MOV Operability Evaluations
In a few cases the licensee's revisions to its calculations resulted in
negative design margins which were documented in operability
evaluations. The following valves were affected:
RHR-744A and B
CC-749A and B
FCV-626
FW-V2-6B
CVC-381
The inspectors reviewed each assessment and agreed with the licensee's
conclusions. The inspectors also reviewed outage plans and verified
that all of these MOVs were scheduled to be modified during the upcoming
RO 18 to increase their actuator capabilities.
6. Concerns Identified by Inspection 50-261/97-12
Inspection 50-261/97-12 documented a number of generally minor concerns
regarding the licensee's procedures, assumptions, etc. The inspectors
re-examined these concerns during the current inspection. With one
important exception, the concerns were satisfactorily addressed by
already completed actions or by planned actions described above. The
exception involved the licensee's use of handwheel turns to establish
closing limit switch control and torque switch bypass settings. The
inspectors noted that a more precise method of verifying these settings
(such as diagnostic trace analysis) would be appropriate, based on
.industry experience. In a letter to the NRC dated February 20, 1998,
verification of close limit switch and torque switch bypass settings for
18
valves that were position-controlled for accident scenarios, if the
valves were capable of being diagnostically tested. The letter
indicated these actions would be complete by June 25, 1998.
c. Conclusions
With the commitments made in the licensee's letter-dated February 20.
1998, the inspectors determined that .the licensee met the intent of
'GL 89-10 in verifying the design-basis capability of the safety-related
MOVs at Robinson. The licensee's letter identified the following
commitments:
Testing will be conducted during Refueling Outage 18 to support
the stem coefficient of friction assumptions contained in ESR
9700331.
The hydrodynamic torque requirements for motor-operated
butterfly valves V6-16A/B/C will be evaluated.
Site procedures will be revised to require diagnostic
verification of close limit switch and torque switch bypass
settings for valves that are position-controlled for accident
scenarios, if the valves are capable of being diagnostically
tested.
Based on the NRC inspections of the licensee's implementation of
GL 89-10 and on the licensee's'commitments in its letter dated
February 20, 1998, the NRC is closing its review of the GL 89-10 program
at Robinson. Resolution of the three outstanding licensee commitments
listed immediately above is identified as inspector followup item
50-261/98-01-01, GL 89-10 Commitments.
E2
Engineering Support of Facilities and Equipment
E2.1 Preventive and Predictive Maintenance Activities
a. Inspection Scope
The inspector reviewed and discussed licensee's preventive and
predictive maintenance activities.
b. Observations and Findings
During a review of the scope of the preventive maintenance (PM)
activities as they apply to major rotating equipment, the inspector
noted that numerous safety related motors, including the motors for the
containment spray pumps, safety injection pumps, EDG lube oil pumps,
and boric acid transfer pumps did not have a PM nor had they been
overhauled since initial installation. Upon questioning, the inspector
was informed that PM and overhaul of motors was not necessary as there
were other means of detecting an impending failure, such as the ASME
Section XI testing as well as predictive maintenance activities
19
performed on these equipment. Further, the inspector was also informed
that some of the motors associated with safety related systems were
normally in the standby mode, and therefore were not subjected to
excessive run times. The inspector reviewed procedure MMM-005,
Preventive Maintenance Program and noted that it did not give any
specific guidance with regard to motor PM and overhaul, including that
based on vendor recommendation. The inspector also noted that the
licensee did not have a documented basis for not.performing PM/overhaul
on certain motors. The inspector discussed this with the licensee and
was informed that this aspect of the PM program would be further
reviewed by the system engineer, including verification of vendor
recommendations. Further, based on plant specific requirements, the PM
would either be performed, or if not, appropriately justified.
The inspector also reviewed activities related to predictive
maintenance. The inspector reviewed a Self Assessment Report, RESS
96-34, that was conducted in late 1996 relative to the predictive
maintenance program, as well as Condition Report (CR) 96-02908 that was
generated to track the corrective actions from the self assessment.
This self assessment had identified several weaknesses: most significant
was that the predictive maintenance program was too segregated, leading
to inconsistent expectations for review and disposition by the system
engineers. Additionally, the self assessment identified weakness in the
overall integration of the predictive maintenance program with other
programs, including PM. Since the self assessment, the licensee had
initiated numerous efforts to address the weaknesses. This includes
management of the program by the rapid response team supervisor within
the site engineering organization, increasing program awareness, and
issuing a monthly status report. Key attributes of the program include
lube oil analysis, 'Vibration analysis, and thermography.. The licensee
currently has three engineers who are primarily responsible for the
implementation of the program. The inspector did note that the licensee
did not have an overall program document that describes and prescribes
the various attributes of the program. Currently, the licensee plans to
develop an overall program document in the middle part of 1998.
c. Conclusion
Overall,.the inspector noted that licensee efforts related to the
predictive maintenance program continue to be on a positive trend. The
absence of a predictive maintenance program document as well as the lack
of PM, including overhaul of certain safety related motors without
adequate justification was discussed with the licensee. The licensee
plans to address these issues in the near future to improve on existing
activities.
20
E3
Engineering Procedures and Documentation
E3.1 Steam Generator Power Operated Relief Valve (S/G PORV) Transducer (I/P)
Replacement
a. Inspection Scope (37551)
The inspector reviewed implementation of ESR 970047. This ESR was
originated to address a problem related to S/G PORV setpoint drift which
the plant experienced, and required frequent calibration. The ESR was
developed to replace the existing I/P with those that would yield better
dependability and accuracy.
b. Observation and Findings
Currently, the three S/G PORVs utilize Fisher 546 electro-pneumatic
signal I/Ps. These I/Ps were noted to be drifting due to sensitivity to
temperature variations as well as vibration. This drift required
frequent calibration to assure correct S/G PORV lift setpoint. The S/G
PORVs were tracked as maintenance rule a(1) category in accordance with
10 CFR 50.65. Consequently, the licensee developed and ESR to replace
the Fisher 546 I/Ps with Ronan X55-600 I/Ps. The Ronan I/P utilizes
advanced solid state technology and were designed to be less succeptable
to drift caused by variatiohs in temperature and vibration. During the
installation of the Ronan model, the licensee discovered'that the
failure mode of the newly installed Ronan model was different than what
was expected as well as required. The Fisher model I/P failure.mode
upon a loss of power was to fail the S/G PORV in the closed position.
The safety function of the S/G PORV is the closed position. The Ronan
I/P was found to default to a low pneumatic output upon loss of loop
current, i.e., loss of power. The low pneumatic output equates to the
S/G PORV failing in the open, i.e., non-safe position, upon loss of
power. The failure mode of the newly installed Ronan I/P was discovered
"co-incidently" by an I&C technician during installation when the loop
power isolated for an unrelated reason. Upon identification of this
problem, the I&C technicians stopped the installation and contacted the
responsible engineer.
The .inspector reviewed the circumstances related to the issue. It was
noted that the ESR had not specified the verification of the failure
mode as part of receipt (from vendor) or post modification testing. The
post modification testing associated with the ESR only included
calibration of the Ronan I/P. The inspector was informed by the
licensee that the manufacturer's documentation had not indicated the
failure mode and that the licensee was erroneously informed by the
manufacturer that the I/P fails with high output, i.e. safe position.
The licensee contacted the manufacturer following the identification of
the failure mode. At this time, the licensee.was informed that the
Ronan I/P failed with a low pneumatic output, i.e., different failure
mode than initially communicated.
21
The inspectors reviewed applicable requirements, including ANSI 45.2.11,
Quality Assurance Requirements for the Design of Nuclear Power Plants
and implementing licensee procedure EGR-NGGC-0005, Engineering Service
Requests. ANSI 45.2.11, Section 3 Design Input Requirements and
procedure EGR-NGGC-005, Attachment 2, Design Inputs, state that "The
design input shall include ... Failure effects requirements of
structures, systems, and components". Additionally, EGR-NGGC-005,
Section 9.4.7.j. Testing Requirements, states that "Testing shall verify
that: the modified system/component functions/performs as intended...".
Contrary to the above, the licensee had not adequately considered the
failure effect of the new Ronan I/P and appropriately prescribed a post
modification. test to verify that the modified system would perform as
intended. The inspector determined that failure to follow EDG-NGGC-005
in considering the failure effects was a violation. This licensee
identified, corrected, and non-repetitive violation is being treated as
an NCV, consistent with Section VII.B.1 of the NRC enforcement policy.
This issue is documented-as NCV 50-261/98-01-02:
Failure to Consider
Failure Modes For S/G PORV I/P ESR.
Upon identification, the licensee made a decision to back out of the ESR
and reinstall the "old" Fisher Model 546 I/P. Additionally, the
licensee decided to further review the availability of other designs
that best suited plant needs. Current licensee plans are to perform the
ESR in May 1998, following the completion of the outage. In the mean
time, the licensee will calibrate the I/Ps as needed, following drift.
Additionally, the S/G PORV will continue to.be monitored as a(1) system
in accordance with 10 CFR 50.65.
c. Conclusion
The I&C technician exhibited good questioning attitude when he noticed
that the Ronan I/P had behaved differently than what he had remembered
from a previous conversation with the responsible engineer. A non-cited
violation was identified for not adequately considering the failure
modes associated with the installation of new Ronan I/P. Upon
identification, the licensee immediately stopped the installation
activity and reconsidered the merits of the ESR as written.
E7
Quality Assurance in Engineering Activities
E7.1 Special UFSAR Review (37551)
A recent discovery of a licensee operating their facility in a manner
contrary to the UFSAR description highlighted the need for a special
focused review that compares plant practices, procedures and/or
parameters to the UFSAR descriptions. While performing the inspections
discussed in this report, the inspector reviewed the applicable portions
of the UFSAR related to the areas inspected. The inspector verified
that for the select portions of the UFSAR reviewed, the UFSAR wording
was consistent with the observed plant practices, procedures and/or.
parameters.
22
E8
Miscellaneous Engineering Issues (92903) (37551)
E8.1 (Closed) Unresolved Item 50-261/97-201-05, AFW UFSAR Discrepancies:
Sections E.1.2.6. E.1.3.6. and E.1.4.2.6 of NRC Inspection Report (IR)
50-261/97-201 listed 11 comments/discrepancies in the updated final
safety analysis report (UFSAR). These were as follows:
1)
Changes to design and operation of the safety injection
(SI) and residual heat removal (RHR) pumps and the
effects of the design and operational changes on SI,
containment spray, and RHR pump NPSH were not discussed
in the UFSAR.
2)
UFSAR Table 6.2.4-1 (Table 6.4.2-1 listed in the IR was
the incorrect table number) had not been revised to
indicate that the discs of containment isolation valves
SI-860A, SI-860B. SI-861A, and SI-861B had been drilled
for pressure relief.
3)
UFSAR Table 6.3.2-5 stated that the maximum SI pump flow
rate is 550 gpm, which was less than the actual pump
maximum flow rate.
4)
UFSAR Section 10.4.8.2 did not list Anticipated Transient
Without Scram (ATWS) Mitigation System Actuation
Circuitry (AMSAC) as a start signal for AFW.
5)
UFSAR Table 10.4.8-1 incorrectly stated that the SDAFW
pump was 387 horsepower. The correct horsepower was 733.
6)
UFSAR Section 6.3.2.2.8 implied that a minimum of 300.000
gallons was available for delivery from the RWST. The
correct value was 277,999 gallons. This figure has been
revised again by the modification which increased RWST
level to address the insufficient NPSH problem.
7)
UFSAR Section 6.3.2.2.17 implied that the SI system high
pressure branch lines were designed for a pressure of
1500 psig. The correct value was 1750 psig.
8)
UFSAR Section 8.3.2 stated that station battery A cell
type was NCX. The correct cell type was NCN.
9)
UFSAR Section 8.3.1.1.2 referenced Motor Control Center
(MCC) numbers 5A1,
5A2, and 6A, and implied the number of
phases for the transformer supply was 30.
No MCCs
existed with these number designations. The correct
number of phases for the transformer supply to MCCs 9 and
10 was 3.
23
10)
UFSAR Table 8.3.1 did not reference Calculation RNP-E
8.016, Revision 5. However, this calculation was
referenced in Section 8.3 of the UFSAR.
11)
UFSAR Figures 8.3.1-3 and 8.3.1-4 stated that the CCW
pump motor was 400 horsepower. The correct horsepower
rating for the CCW motor was 350 horsepower.
The inspectors concluded that none of the above discrepancies
resulted in an unreviewed safety question. The inconsistencies in
the UFSAR concerning design and operation of the SI and RHR pumps
and the effects of the design and operational changes on SI,
containment spray, and RHR-pump net positive suction head were
included in resolution of Apparent Violation Item EEI 50-261-98
03-04. After further review of the issue concerning the need to
update UFSAR Table 6.2.4-1 concerning drilling of the discs of
containment isolation valves SI-860A. SI-860B, SI-861A. and SI
861B for pressure relief, the inspectors concluded that this did
not represent an error in the UFSAR. Further review of UFSAR
Section 6.3.2.2.3 disclosed additional inconsistencies in the SI
pump flowrate. The inspectors also noted that the SI pump design
parameters listed in Table 6.3.2-5 differed from those shown in
UFSAR Figure 6.3.2-4.
10 CFR 50.71(e) requires the UFSAR to be revised to include the
effects of all changes made in the facility or procedures as
described in the UFSAR. Examples 3-5 listed above of failure to
revise the UFSAR to include the effects of changes made in the
facility or procedures as described in the UFSAR was identified to
the licensee as violation item 50-261/98-01-03. Failure to Update
the UFSAR.
The licensee identified the discrepancies listed as Examples 6-11,
above, during their UFSAR review program, and corrective actions
were being initiated to update the UFSAR. This licensee
identified, corrected, and non-repetitive violation is being
treated as an NCV. consistent with Section VII.B.1 of the NRC
enforcement policy. This issue is documented as NCV 50-261/
98-01-04: Licensee Identified UFSAR Discrepancies.
E8.2 (Closed) Violation 50-261/97-07-03: ESR design verification
requirements. The corrective actions presented in the licensee's
response, dated August 5, 1997 were reviewed and verified by the
inspector. The NRC accepted the response by letter, dated August 18,
1997. Design verification was performed on the ESRs that had not been
design verified. The inspector verified a sample of these and had no
questions or problems. This violation is closed.
E8.3 (Closed) URI 50-261/97-07-02: Spent Fuel Pool Level Issues. By letter
dated March 27, 1997, the licensee informed the NRC that a discrepancy
had been identified with regard to the basis for the Improved Technical
Specification,-ITS, requirement to maintain 21 feet of water above the
24
spent fuel in the spent fuel pit. The issue was discussed in NRC
inspection report 50-261/97-07, paragraph E2.1. Due to the physical
limitation of the spent fuel pool, the ITS proposed limit of 23 feet
above the spent fuel could not be achieved. The licensee performed an
analysis of the issue and proposed a revision to the UFSAR Section
15.7.4, Design Basis Fuel Handling Accident. By letter dated October 2.
1997, the licensee submitted the issue to the NRC as a Unreviewed Safety
Question and requested the NRC to review and approve the proposed change
to the UFSAR. The licensee's analysis concluded that a significant
hazard would not be created by the change. The NRC reviewed the
methodology used by the licensee and concluded that the requested change
was acceptable. By letter dated January 27, 1998, the NRC informed the
licensee of the acceptance of the change and issued Amendment 177 to the
Facility Operating License No. DPR-23 for the Robinson facility. Unit 2.
The inspector verified the implementation of the change by reviewing the
UFSAR change, which is held in a pending file until six months after
Refueling Outage 18. scheduled for March 7, 1998. Night Order 97-016,
which was issued at the time this issue was discovered was canceled and
activities in the spent fuel pool are no longer restricted due to this
issue. This URI is closed.
E8.4 (Closed) Violation (VIO) 50-261/96-12-05: Unjustified design
assumptions and incorrect stem rejection load. Inspection 50-261/97-12,
determined that the corrective actions to resolve this violation had
been adequately completed, except for completion of assumption
justifications and correction of calculations. The assumption
justifications and calculations were reviewed by the inspectors during
the currentinspection (see E1.2 above).
The inspectors found that the
assumption justifications had been completed and that the calculations
had been corrected as stated in the licensee's response letters for this
violation. Several concerns were identified which are being
appropriately addressed by the licensee's Generic Letter 89-10 Site
Improvement Plan. The inspectors questioned whether the licensee had
sufficiently justified its design assumptions for stem coefficient of
friction. The licensee provided a commitment to address that issue,
which will be reviewed in the future as part of inspector followup item
50-261/98-01-OX, GL 89-10 Commitments.
This violation is closed.
E8.5
(Closed) Violation 50-261/96-12-06:
Inadequate evaluation of test
results.
Inspection 50-261/97-12, determined that the corrective
actions to resolve this violation had been adequately completed, except
for revisions to calculations and certain additional training of MOV
personnel. The calculation revisions were reviewed by the inspectors
during the current inspection (see E1.X above) and found to have been
satisfactorily completed. The training was not completed yet but was
scheduled to be performed in the upcoming outage (March 1998) and was
being tracked as an open corrective action in the licensee's database.
This violation is closed.
25
E8.6 (Closed) Inspector Followup Item (IFI) 50-261/96-12-07: Actions to
preclude pressure locking. This item was identified to verify the
adequacy of the licensee's long term actions to preclude pressure
locking of certain valves. This issue is now being addressed through a
safety evaluation report being prepared by the NRC Office of Nuclear
Reactor Regulation. This item is closed.
IV.
Plant Support
R1
Radiological Protection and Chemistry Controls
R1.1 General Comments (71750)
The inspector periodically toured the Radiological Control Area (RCA)
during the inspection period. Radiological control practices were
observed and discussed with radiological control personnel including RCA
entry and exit, survey postings, locked high radiation areas, and
radiological area material conditions. The inspector concluded that
radiation control practices were proper.
R1.2 Radiological Protection (1997 Performance Data)
a. Scope
The inspector reviewed and discussed overall licensee performance for
1997 in the Radiological Protection area.
b. Observations and Findings
For 1997, the cumulative dose at Robinson was 12.991 Person-Rem. This
was the lowest dose ever received at Robinson as well as at any of the
CP&L plants. Of the 775 monitored individuals, 471 individuals had non
measurable exposure. Of the people who had measurable exposure, the
five highest dose recipients had 285, 281, 246, 234, and 220 mRem. Four
of the five individuals were affiliated with the Radiation Protection
organization. The exposure level of the five individuals receiving the
highest dose were well within the 10 CFR 20 exposure limits. The
contaminated floor space was maintained at approximately 1.101 square
feet verses a plant goal of less than 1,500 square feet. There were a
total of 14 personal contamination events verses a goal of 50. The
number of locked high radiation areas remained unchanged at 5.
c. Conclusions
The inspector concluded that overall licensee performance in 1997 in the
Radiological Protection area was excellent.
26
Si
Conduct of Security and Safeguards Activities
51.1 General Comments (71750)
During the period, the inspector toured the protected area and noted
that the perimeter fence was intact and not compromised by erosion or
disrepair. Isolation zones were maintained on both sides of the barrier
and were free of objects which could shield or conceal an individual.
The inspector periodically observed personnel, packages, and vehicles
entering the protected area and verified that necessary searches,
visitor escorting, and special purpose detectors were used as applicable
prior to entry. Lighting of the perimeter and of the protected area was
acceptable and met illumination requirements.
R8
Miscellaneous Radiation Protection -and Controls (92904)
R8.1 (Closed) URI 50-261/97-01-06: Demonstrate Accurate Dose Monitoring and
Dose Assignment Practices and Procedures. An assessment of the
dosimetry program was completed during the report period for NRC
inspection report 50-261/97-01. Issues regarding demonstration of
accurate and reasonable dose tracking and dose assignment practices
were identified. Condition Report 97-00172 was issued by the licensee
to resolve these issues. CP&L corporate dosimetry procedure DOS-NGGC
0002, Dosimetry Issuance was revised on October 23, 1997. In addition,
the location of the thermo-luminescent-dosimeter (TLD) badge rack was
moved from the RCA entrance to a lower background area outside the
Security Building. On January 1, 1998 the licensee instituted the
policy of requiring personnel with access to the RCA to wear their TLD
home with them or store it outside the security fence. The site TLD
background badges are stored at the Visitor's Center to minimize any
contribution from the operating unit when background is subtracted.
These changes were reviewed and verified by the inspector. The
inspector concluded that the licensee had adequately addressed the
issues identified by this URI.
This URI is closed.
V. Management Meetings
X1
Exit Meeting Summary
The inspectors presented the inspection results to members of licensee
management at the conclusion of the inspection on No proprietary
information was identified.
27
PARTIAL LIST OF PERSONS CONTACTED
Licensee
J. Boska, Manager, Operations
H. Chernoff. Supervisor, Licensing/Regulatory Programs
T. Cleary, Manager. Maintenance
J. Clements. Manager, Site Support Services
J. Keenan, Vice President, Robinson Nuclear Plant
R. Duncan, Manager, Robinson Engineering Support Services
R. Moore, Manager, Outage Management
J. Moyer, Manager, Robinson Plant
D. Stoddard, Manager, Operating Experience Assessment
R. Warden, Manager, Nuclear Assessment Section
T. Wilkerson, Manager, Regulatory Affairs
D. Young. Director,. Site Operations
NRC
B. Desai, Senior Resident Inspector
M. Shymlock. Branch Chief, Region II
28
INSPECTION PROCEDURES USED
IP 37551:
Onsite Engineering
IP 40500:
Effectiveness of Licensee Controls in Identifying, Resolving,
and Preventing Problems
IP 61726:
Surveillance Observations
- IP
62700:
Maintenance Implementation
IP 62707:
Maintenance Observation
IP 71707:
Plant Operations
IP 71750:
Plant Support Activities
IP 92901:
Followup - Operations
IP 92902:
Followup - Maintenance
IP 92903:
Followup - Engineering
IP 92904:
Followup - Plant Support
T12515/109:
Inspection Requirements for Generic Letter 89-10, Safety
Related Motor-Operated Valve Testing and Surveillance
T12515/130:
Improved Standard Technical Specification Implementation Audits
ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
Oype
Item Number
Status
Description and Reference
IFI
50-261/98-01-01
Open
GL 89-10 Commitments (Section E1.2)
50-261/98-01-02
Open
Failure to Consider Failure Modes For S/G
50-261/98-01-03
Open
Failure to Update the UFSAR (Section E8.1)
50-261/98-01-04
Open
Licensee Identified UFSAR Discrepancies
(Section E8.1)
Closed
Type Item Number
Status
Description and Reference
50-261/97-06-01
Closed
Inadequate safeguards procedures that
allowed ESF train being out-of-service
without invoking a Technical Specification
action statement.(Section 08.1)
50-261/97-09-02
Closed
Failure to properly calibrate OPDT
channels (Section M8.1)
50-261/98-01-02
Closed
Failure to Consider Failure Modes For S/G
50-261/97-201-05 Closed
AFW UFSAR Discrepancies (Section E8.1)
29
50-261/98-01-04
Closed
Licensee Identified UFSAR
Discrepancies (Section E8.1)
50-261/97-07-03
Closed
ESR design verification requirements
(Section E8.2).
50-261/97-07-02
Closed
Spent Fuel Pool Level Issues
(Section E8.3).
50-261/96-12-05
Closed
Unjustified Design Assumptions and
Incorrect Stem Rejection Load
(Section E8.4)
50-261/96-12-06
Closed
Inadequate Evaluation of Test
Results (Section E8.5)
IFI
50-261/96-12-07
Closed
Actions to Preclude Pressure Locking
(Section E8.6)
50-261/97-01-06
Closed
Demonstrate Accurate Dose Monitoring
and Dose Assignment Practices and
Procedures (Section R8.1)