IR 05000400/1989017
| ML18005B076 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 09/14/1989 |
| From: | Bradford W, Dance H, Shannon M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18005B075 | List: |
| References | |
| 50-400-89-17, IEB-86-003, IEB-86-3, NUDOCS 8910030555 | |
| Download: ML18005B076 (22) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTASTREET, N.W.
ATLANTA,GEORGIA 30323 Report No.:
50-400/89-17 Licensee:
Carolina Power and Light Company P. 0.
Box 1551
'aleigh, N.
C. 27602 Docket No.: '0-400 Facility Name:
Harris
License No.:
NPF-63 Inspection Conducted:
July 21 - August 25, 1989 inspectors: '<"
.Y K H. Bradford M. C.
Shannon Accompanying Personnel:
R. Becker, Licensing Project Manager Of ice of Nuclear Reactor Regulation Approved by:
H.
C.
Dance, Section Chief.
Division of Reactor Projects ate igned ate igned D te S'gned SUMMARY Scope:
This routine, safety inspection was conducted in the areas of operational safety verification, surveillance observations, maintenance observations, 50.59 Review, and On-site Followup of Events.
Results:
Within the areas inspected no violations were identified.
Two unresolved items were identified.
The first unresolved item involved what appeared to be an inadequate dedication process of commercial grade breakers used in safety related applications; paragraph 5.a.
The second unresolved item involved a potential inadequate testing of containment electrical penetration protection devices (breakers)
because of the test method and/or test equipment; paragraph 5.a.
t The licensee issued a potential
CFR Part 21 report on ITE molded case circuit breakers because of their repeated failure to hold the load during motor starting evolutions; paragraph 5.a.
8910030555 890914 PDR ADQCK '05000400
PNU
The 50.59 review process at Harris has two characteristics which should be considered strengths.
First, a
cadre of trained; tested and experienced reviewers by technical discipline is maintained to perform 50.59 evaluations.'econd, when an evaluation is performed, two independent reviewers perform the evaluation to confirm the conclusion, paragraph 6.
This report also discusses problems encountered during spent fuel transfer, paragraph 5.c.;
and repairs to the emergency diesel generator sequencer control room indication and subsequent inadvertent service water booster pump actuation, paragraph REPORT DETAILS Persons Contacted Licensee Employees
~R. Biggerstaff, Principal Engineer, Onsite Nuclear Safety D. Braund, Supervisor, Security J. Collins, Manager, Operations G. Forehand, Director, OA/QC C. Gibson, Director, Programs and Procedures
'C. Hinnant, Plant General Manager L. Lentz, Operations Support Supervisor
+L., McKenzie, Principal Engineer, QA/QC T. Norton, Manager, Maintenance
+R. Oates, Principal Engineer, Harris Licensing C. Olexik, Supervisor, Shift Operations R. Richey, Manager, Haris Nuclear Project Department J. Sipp, Manager, Environmental and Radiation Monitoring H. Smith, Supervisor, Radwaste Operation
+D. Tibbits, Director, Regulatory Compliance B.
Van Metre, Manager, Technical Support E. Willett, Manager, Outages and Modifications Other licensee employees contacted during this inspection included engineers, operators, mechanics, security force members, technicians, and administrative personnel.
2.
- Attended monthly-exit interview on August 25, 1989
+Attended 50.59 review exit interview on August 4, 1989 Acronyms and initialisms used in the report are listed in Paragraph 9.
Operational Safety Verification (71707)
The inspectors observed control room operations, reviewed applicable logs, and conducted discussions with control room operators during the 'report period.
The operability of selected emergency systems was verified, tagout records were reviewed, and proper return to service of affected components was verified.
The inspector conducted routine plant tours during this inspection period to verify the licensee's requirements and commitments with selected LCOs and results of selected surveillance tests.
The verifications were accomplished by direct observation of monitoring instrumentation, valve positions, switch positions, accessible hydraulic snubbers, and review of completed logs, records, and chemistry results.
The licensee's compliance with LCO action statements were reviewed as events occurre '
The inspectors routinely attended meetings with certain licensee management and observed various shift turnovers between shift foremen and licensed operators.
'hese meetings and discussions provided a daily status of plant operations, maintenance, and testing activities in progress, as well as discussions of significant problems.
Selected activities of site security and the licensee's Radiological Protection Program were reviewed by the inspectors to verify conformance with plant procedures and NRC regulatory requirements.
No violations or deviations were identified.
3.
Surveillance Observation (61726)
Portions of the following surveillance inspections and tests required by the,TS were observed or reviewed by the inspectors:
MST-E0006 480 VAC Molded Case Circuit Breaker Test EPT-033 Emergency Safeguards Sequencer System Test OST-1823 IA-SA Emergency Diesel Gen'erator 18 month Operability Test FPT-3501 Fire Door Monitor Trip Actuating Device Operational Test The inspectors verified'hat the surveillances were performed in accordance with adequate procedures; instrumentation was calibrated; limiting conditions were met; test results met acceptance criteria and were reviewed by personnel other than the individual directing the test; deficiencies identified during the testing were properly reviewed and resolved by appropriate management personnel; and personnel conducting the test were qualified.
No violations or deviations were identified.
4.
Monthly Maintenance Observation (62703)
.The inspector observed/reviewed the following maintenance activities of safety-related and non safety-related systems and componen'ts to ascertain that they were conducted in accordance with approved procedures, regulatory guides, industry codes and standards, and were in conformance with TS:
Molded Case Circuit Breaker Testing S-4 Fan Meggering and Bridge Check Spent Fuel Transfer Activities Diesel Generator
"B" Sequencer Indica'tion Circuit Repair
"A" 8 "B" Diesel Generator Sequencer Improper I'ndication Modification No violations or deviations were identifie.
On-Site Follow-up of Events (93702)
a
~
Molded Case ITE Circuit Breakers The Licensee has experienced difficulties with molded case circuit breakers manufactured by Seimens Automated, Inc.
The 100 AMP ITE breakers were installed in the plant and have failed to maintain the load when energized.
Breaker testing discrepancies were discussed originally in Inspection Report 89-13.
During the gA followup of middle suppliers for Bulletin 88-10, fraudulent breakers were identified.
Replacement breakers were or had already been purchased from ITE.
The new breakers which had gone through various design changes were supposed to be identical replace-ments.
The design changes took place when Siemens Automated, Inc.
purchased ITE from Gould, Inc.
The only way to identify the new breakers is by the voltage rating of the breaker (i.e.,
a 600 volt rating indicates a probable Gould manufacture and a
480 volt rating indicates a probable Siemens manufacture).
The 480 volt breakers had an instantaneous trip rating that was double the 600 volt breaker rating (i.e.,
480 Vac instantaneous trip band is 1200 - 2000 amps and the 600 Vac instantaneous trip band is-600 - 1000 amps).
Difficulties were first encountered when the new breakers were placed on a test rig and failed to meet the in-house testing requirements.
An ITE technical representative was called in to resolve the testing discrepancies.
The testing method was changed to meet the manufacturer's requirements and the breakers passed the test.
Mhen installed and operated, the breakers would not hold the load and tripped on instantaneous current, just as predicted by the original test.
Eleven breakers were tested and failed to hold the load.
The manufacturer was again contacted and arrangements were made to take four breakers to the ITE manufacturing facility to be tested.
The NRC Resident requested access to the testing but was told access would be denied to the manufacturing facility if NRC credentials were shown.
Testing was performed on the four breakers with two CPSL Engineers present.
All four breakers failed to meet the manufacturer's require-ments.
Three breakers, of the new design (480V), failed by tripping low, below 1200 AMPS.
One breaker, of the older design (600Y),
failed by tripping high, above 1000 AMPS.
The following is a
tabulation of the test results:
600 volt class breaker old model number Poles Trip (amps)
Siemens Test 1773 1773 1703 Acceptance Criteria (amps)
600-1000 600-1000 600-1000 480V class breaker HE3B100 (a)
HE38100 (S4-B 82)
Pol es Trip (amps)
Siemens Test 1231 1489 886 Acceptance'riteria 1200-2000 1200-2000 1200-2000 (b)
HE3B100 (Bkr b3)
Poles Trip (amps)
Siemens Test 1063 1063 1063 Acceptance Criteria 1200-2000 1200-2000 1200-2000 (c)
HE3B100 (S4-B b'1)
Trip (amps)
Acceptance Pol es Si emens Test Criteri a
992 1200-2000 709 1200-2000 709 1200-2000 A total of eleven 100 amp, 480 Vac breakers had been purchased, installed, and failed to hold the load.
Initially, two of the breakers were installed on a
safety-related component, diesel generator room cooling fan.
The breakers failed in this application
and were then installed on the reactor support cooling fan.
They failed in this application as well.
With the multiple failures and initial use on a safety-related component, the licensee initiated a
potential
CFR Part 21 notification.
The licensee is continuing to investigate the root cause and will use the root cause to determine if an actual Part 21 Report is required.
During an NRC Maintenance Team Inspection, the inspectors were initially informed that the ITE Nolded Case Circuit Breakers were purchased commercial grade.
After further review, it was'oted that the switchgear cabinets were purchased safety grade with the breakers included.
The breakers wer'e purchased commercial grade and upgraded by the switchgear supplier.
The Licensee subsequently purchased commercia'1 grade breakers and used a dedication process to upgrade the breakers to safety grade.
The dedication process appeared to be inadequate, in that various critical characteristics were not tested or verified as part of the dedication process.
IEEE Standard 323-1974 indicates that:
the qualification methods described shall be used for qualifying equip-ment.
The capability of all class IE equipment, including inter-faces, of a
nuclear power generating station for performing its required functions shall be demonstrated.
It is preferred that the demonstration be done by type tests on actual equipment.
The type test shall be designed to demonstrate that the equipment performance meets or exceeds the requirements of the equipment specifications for the plant, The type test shall consist of a planned sequence of test conditions that meet or exceed the expected or specified service conditions, including performance margin, and shall take account of both normal and abnormal operation.
During the licensee's dedication process, the procurement documents are reviewed to verify conformance and the component is placed in service.
In the case of the commercial grade breaker, following installation in the applicable circuit and operation of the supplied equipment, the dedication process is completed.
The dedication process also relies upon UL testing for equipment reliability.
Conversations with NRR indicate that UL testing cannot be used as part of the dedication process and that UL testing has no statistical sample reliability.
Critical characteristics such as thermal over-load and instantaneous trip setpoints are not verified as par t of the dedication process.
The majority of the breakers that were purchased commercial grade were tested by the electrical department prior to installation.
What appears to be a program deficiency is identified as an
- unresolved item:
URI 89-17-01, Potentially Inadequate Commercial to Safety Grade Dedication Process.
"Unresolved items are matters about which more information is required to determine whether they are acceptable or may involve violations or deviation As also noted in Inspection Report 89-13, the licensee's testing equipment appeared to be inadequate.for measuring the instantaneous current trip setpoints for the molded case circuit breakers.
The test equipment uses a
memory circuit and an analog meter to record the instantaneous trip current.
The inspectors noted that this was
'robably inadequate because of the circuit delays.
Additionally, the ITE technical representative noted that the test equipment was inadequate in that an oscilloscope and shunt were required, due to the less than 1 cycle trip time.
The licensee performed trip testing on each pole of the four breakers prior to testing by ITE at the ITE Wilmington manufacturing facility.
The following is a tabulation of the results from both tests:
600 volt clas's breaker old model number Trip (amps)
.
Trip (amps)
Poles CP&L Test (CM-E0010)
Siemens Test 580
670
560 480V class breaker HE3B100 (a)
HE38100 (S4-B 82)
Trip (amps)
Poles CP&L Test (CM-b'0010)
275 375 375 (b)
HE3B100 (Bkr P3)
Trip (amps)
Poles CP&L Test (CM-80010)
370 410 320 1773 1773 1703 Trip (amps)
Siemens Test 1241
'489 886 Trip (amps)
Siemens Test 1063 1063 1063
(c)
HE3B100 (S4-B 81)
Trip (amps)
Trip (amps)
Poles CPSL Test (CN-E0010)
Siemens Test
425 992 400 709 400 709 The test results indicated that CPSL's testing has not been adequate when compared to the manufacturer's testing.
An additional problem was noted, in that the original breaker that was tested, tripped high.
In a variety of applications, this type breaker is used on containment penetrations and is therefore tested per TS 3/4.8.4, Electrical Equipment Protective Devices.
This TS requires testing by injecting a current equal to plus or minus 20% of the "pickup value", which has been interpreted as the manufacturer's instantaneous trip curve values.
The test results indicate that the test methodology and test equipment may have been inadequate to verify conformance to TS 3/4.8.4.
Discussions with the licensee indicated that appropriate test methodology will be established and'dditional test equipment will be used for breaker testing in the future.
The concern with inadequate breaker testing is identified as an unresolved item:
URI 89-17-02, Potentially Inadequate Testing of Containment Penetration Conductor Overcurrent Protective Devices.
b.
Emergency Sequencer Indication Problems The Licensee identified a
problem with the diesel generator load sequencer in that the main control board sequencer indicators were improperly wired.
The improper wiring did not cause an operability concern with the sequencers, but could have caused an indication problem for the reactor operators.
Troubleshooting activities were initiated to correct the various wiring 'discrepancies.
During the corrective maintenance the tech-nician inadvertantly jumpered contacts which caused the
"B" ESM booster pump to start.
The pump was stopped by the Control Room and work on the sequencer was halted until the event was investigated and understood.
The technician stated that he had found the terminals called for by the'ystem engineer, reached for the jumper, and mistakenly attached the jumper to terminals located directly above the correct terminals.
The technician was counselled on attention to work activities and the
c ~
consequences resulting in unnecessary challenges of safety equipment.
This item was identified by the licensee and report in LER-89-013; appropriate notifications were made.
Spent Fuel Cask During the second fuel transfer evolution the licensee experienced difficulties in removal of the cask head.
During the head lift'one of four lifting cables broke.
The head had shifted and this resulted in the head catching on the head studs.
By manipulating the crane, the head was freed and the decision was made by management to continue the liftwith three cables.
The head became stuck again and a second cable broke.
At this point the ca'sk had to be lifted to the upper part of the pool and come alongs were used to free the head.
The head was then placed back down on the cask.
Plant management was aware of difficulties encountered by the lifting crew and health physics personnel were present during all of the various evolutions.
After receipt of four, cables from Brunswick, the head was lifted wi.th no further difficulties and the fuel was transferred; Following the transfer it was noted that the'head seal retaining clips had been broken and required replacement.
The head studs were inspected and 11 of 33 were found to be damaged and required repair.
The two guide pins were, also found to be damaged and were replaced.
The licensee has developed a six point action plant to deal with the problems experienced during the lifting evolution.
The plan deals with responsibilities for various evolutions in the fuel transfer cycle, and will be fully implemented prior to the next spent fuel transfer.
6.
CFR 50.59 Safety Review Audit During the week of July 31, 1989, the Headquarters licensing Project Manager performed an audit of the Shearon Harris Nuclear Power Plant, Unit 1 (Harris),
safety reviews for conformance with the provisions of
CFR 50.59.
The following discussion summarizes the results of that audit.
a ~
Approach The audit was commenced from a list of recent 50.59 evaluations performed by the licensee; From the 19 evaluations on the list supplied by the licensee, four were selected that appeared of interest from a safety content perspective.
A fifth evaluation came from a group of procedure change 50.59 evaluation The controlling plant procedure, AP-011, Safety Reviews, was reviewed against the most current industry guidance for performing 50.59 reviews, NSAC-125 (Guidelines for 10 CFR 50.59 Safety Evaluations),
dated June 1989, prepared by Nuclear Management and Resources Council.
Although NSAC-125 is not formally endorsed by the NRC, the current version generally contains NRC review comments from previous drafts.
Therefore, NSAC-125 is considered the best consensus guidance currently available.
The plant change requests (PCR)
and the procedure change were then evaluated for conformance to plant procedure, AP-Oll, Safety Reviews.
The PCR evaluations selected for audit were:
PCR-000362 PCR-003993 PCR-004434 PCR-004615 PLP-106 Changes to Gross Failed Fuel Detection System Spent Fuel Cask Purge System 480 Volt Circuit Breaker Replacement Containment Recirculation Sump Level Instrumentation Technical Specification Equipment List Program Discussion and Results (1)
AP-011, Safety Reviews, Plant Operating Manual, Vol. 1, Part 1, approved by Plant General Manager, June 28, 1989.
AP-011. is the plant procedure which governs the performance of
CFR 50.59 reviews for Harris.
There are six sections to this procedure which cover Purpose, References, Responsibilities, Definitions/Abbreviations, Procedure and Attachments.
The procedure was clear and, in general, followed NSAC-125 guide-lines.
Reviews are performed by a qualified reviewer following a
Nuclear Safety Review Checklist.
Although the title might imply a simple check-off list, the procedure clearly states that the analysis shall provide a sufficiently detailed basis to justify the conclusions reached.
A second reviewer independently confirms the conclusion of the first reviewer.
The second reviewer may be in another discipline, if i.t is felt necessary.
The second review establishes'hether an interdisciplinary unreviewed safety question exists or not.
A register of quali-fied reviewers is maintained.
Reviewers are trained and qualified on an annual basis.
gualification is based on an annual course on how to perform a 50.59 review, an examination tailored to the specific discipline, and experience in the specific discipline.
Procedure AP-011 generally conformed to the guidance of NSAC-125, and no discrepancies or weaknesses were note PCR 000362, Changes to the Gross Failed Fuel Detection System (GFFD)
PCR 000362 is a change to reroute tubing for the,. GFFD so that samples for the primary system and GFFD may be taken without having to manually bypass the GFFD when a primary system sample was required.
The GFFD and the primary system sample system (PSS)
were connected in series.
The PCR package was reviewed with special attention to the Nuclear Safety Checklist which accompanies each package.
Two checklists were included in the package, one for each qualified reviewer.
A review for equip-ment qualification was also requested and included in the PCR package.
This evaluation was in conformance with AP-011.
No discrepancies or weaknesses were identified'CR 003993, Spent Fuel Cask Purge System PCR 003993 deals with a
method to purge and dispose of contaminated helium from the spent 'fuel cask during cask operations.
The purging operation was to be accomplished by connecting a
source of demineralized water to the cask drain connection and connecting a temporary hose between the cask vent connection and a permanently installed hose fitting and valve at the heating, ventilation and air conditioning (HVAC) exhaust vent/floor drain system piping.
As demineralized water is pumped into the cask, helium is exhausted into the HVAC exhaust vent piping.
The PCR package was checked with respect to the Nuclear Safety Evaluation Checklist.
In the intervening time between this (PCR 003993)
and the former PCR (PCR 000362),
AP-011 had been modified which included the checklist which now has a slightly different format.
The safety review was per-formed by two qualified reviewers and in conformance with AP-011.
'his evaluation was in conformance with AP-011.
No discrepancies or weaknesses were identified.
PCR 004434, 480 Volt Circuit Breaker Replacement PCR 004434 deals with replacing 480 volt ITE Safety Related/
g-class breakers with a suitable substitute.
The modification is required to provide a
normal replacement and to replace breakers which are suspected of being refurbished with sub-standard parts.
In this case, the evaluation was made by two qualified reviewers in different disciplines.
Different disciplines were required because of the seismic aspects of the qualification.
The evaluations were complete and in conformance with AP-01 This evaluation was in conformance with AP-011.
No discrepancies or weaknesses were identified.
(5)
PCR 004615, Containment Recirculation Sump Level Instrumentation PCR 004615 deals with eliminating a ground loop resulting from the normal cabinet ground and a ground at the level element in the containment sump.
The ground loop was causing level indi-cation errors.
The correction of this ground loop caused the loss of a nonsafety grade low level alarm which will no longer be provided to the operator.
The PCR was reviewed with respect to the Nuclear Safety Review Checklist.
The evaluations were complete, performed by two qualified reviewers, and in con-formance with AP-011.
This evaluation was in conformance with AP-011.
No discrepancies or weaknesses were identified.
(6)
PLP-106, Technical Specification Equipment List Program PLP-106 is a plant procedure which is reviewed under AP-101, the same as the PCRs.
PLP-106 contains a number of,items that have been removed from the TS.
The change to PLP-106 was reviewed in the same fashion as PCRs for its conformance with AP-011.
The change was reviewed by two qualified reviewers and the evalua-tions were complete.
This evaluation was in conformance with AP-011.
No discrepancies or weaknesses were identified.
No violations or deviations were identified.
7.
Follow-up on Previous Inspection Findings (92702)
(Closed) Violation 88-20-01, Failure to verify correct valve positions at least once per
days.
The licensee's letter of September 2,
1988, describes the corrective action initiated.
The inspectors verified that the corrective action was completed, in that the subject skid-mounted valves have been identified and proper procedural changes have been made.
(Closed) Violation 88-40-01, Failure to perform adequate post-maintenance testing.
This violation involved a safety-related room cooler unit being inoperable after maintenance because the post maintenance testing did not identify that the electrical leads had not been properly terminated.
The licensee detailed the corrective action in a letter to the NRC dated March 22, 1989, The inspectors have reviewed the corrective action and find it to be acceptabl '
(Closed)
Violation 88-40-02, Failure to follow procedures.
This violation involved a tagging error which allowed approximately five feet of water in the spent fuel pool to drain into the new fuel storage pool.
The licensee's letter of March 29, 1989, describes the corrective action.
The inspectors verified that this corrective action is adequate.
(Closed)
IFI 88-40-03, Task force resolution of turbine driven auxiliary overspeed trips.
The inspectors have reviewed the resolution to this problem and have observed the corrective action in place.
There have been'o further problems since this action was implemented.
(Closed)
URI 89-08-01, Check valve inservice testing appears to be inadequate due to the acceptance criteria used by the licensee.
The inspector reviewed the testing and codes.
The acceptance criteria is in accordance with the code requirement.
(Closed)
IEB 86-03, Potential failure of multiple ECCS Pumps due to single failure of air-operated valve in minimum flow recirculation line.
The licensee's letter of November 13, 1986, states that the Harris plant is not susceptible to the disablement of multiple ECCS pumps due to a
minimum flow line single failure.
During normal plant operation the CVCS uses a
common recirculation line.
During a safety injection actuation the common recirculation line is automatically isolated and a redundant recirculation line opens for pump protection.
The RHR system does not have a
common recirculation line.
No violation or deviation were identified.
8.
Exi't interview The inspection scope and findings were summarized during management interviews throughout the reporting period and on August 4. and 25, 1989, with those persons indicated in paragraph 1.
The inspection findings, listed below and those addressed in the report summary were discussed in detail.
The licensee acknowledged the inspection findings and did not identify as proprietary any material reviewed by the inspector during the inspection.
Item Number Descri tion/Reference Para ra h
89-17-01 URI-Potentially Inadequate Commercial to Safety Grade Dedication Process, paragraph 5.a.
89-17-02 URI-Potentially Inadequate Testing of Con-tainment Penetration Conductor Overcurrent Protective Devices, paragraph.
Acronymns and ALARA ANSAC ATWS ECCS LCO.
LER MCB NFP MS MSIV NST NRC OP OST PCR PIC.
Cab PNTR QA QC RAB RCDT RCS/RC RHR RPS RTD RWP SF SG SIS STA TFW TS VAC WR/JO initialisms.
As Low As Reasonably Achi'evable ATWS Mitigation System Actuation Circuitry Anticipated Transient Without Scram Emergency Core'ooling System Limiting condition for Operation Licensee Event Report Hain Control Board Hain Feed Pump Main Steam Main Steam Isolation Valve Maintenance Surveillance'est Nuclear Regulatory Commission Operating Procedure Operations Surveillance Test Plant Change Request Primary Instrument Control Cabinet Post Maintenance Test Requirements Quality Assurance Quality Control Reactor Auxiliary Building Reactor Coolant Drain Tank Reactor Coolant System Radiation Heat Removal System Reactor Protection'ystem Resistance Temperature Detector Radiation Work Permit Spent Fuel, System Steam Generator Safety Injection Signal Shift Technical Advisor Temperature
- Feedwater Technical Specification Volt Alternating Current Work Request/Job Order