IR 05000400/1998011
| ML18016A844 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 03/01/1999 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18016A842 | List: |
| References | |
| 50-400-98-11, NUDOCS 9903100164 | |
| Download: ML18016A844 (30) | |
Text
U. S. NUCLEAR REGULATORYCOMMISSION
REGION II
Docket No:
License No:
50-400 NPF-63 Report No:
50-400/98-11 Licensee:
Carolina Power & Light (CP&L)
Facility:
Shearon Harris Nuclear Power Plant, Unit 1 Location:
5413 Shearon Harris Road New Hill, NC 27562 Dates:
December 20, 1998 - January 30, 1999 Inspectors:
Approved by:
J. Brady, Senior Resident Inspector R. Hagar, Resident Inspector B. Bonser, Chief, Projects Branch 4 Division of Reactor Projects 9'P03iOOi64 99030i PDR ADOCK 05000400
PDR Enclosure 2
EXECUTIVESUMMARY Shearon Harris Nuclear Power Plant, Unit 1 NRC Inspection Report 50-400/98-1
This integrated inspection included aspects of licensee operations, engineering, maintenance, and plant support. The report covers a six-week period of resident inspection.
~Oerattone
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During the period, the conduct of operations was in accordance with applicable procedures (Section 01.1).
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Following a reactor trip on January 14, operators followed emergency operating procedures and equipment performed as designed.
The trip was, caused by an improperly performed preventive maintenance activity on a nonsafety-related electrical bus relay (Section 01.2).
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Plant Nuclear Safety Committee members thoroughly and critically reviewed a detailed report of a significant adverse condition (Section 07.1).
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Two examples of a violation were identified for failure to implement a post-trip review in accordance with the review procedure.
Proper plant response following the trip was not completely verified and documented in the post trip review. Another violation was identified for failure to promptly identify and correct that condition in the corrective action program (Section 08.1).
Maintenance
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The maintenance tasks observed were adequately conducted in accordance with procedures
. and approved work plans (Section M1.1).
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The surveillance performances observed were conducted in accordance with applicable procedures (Section M2.1).
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With respect to the equipment problems identified duririg the unit's response to the October 23 reactor trip, the licensee had implemented and was continuing to implement the maintenance rule program in accordance with approved procedures (Section M7.1).
k
~En ineerin
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Aviolation was identified for failure to develop an adequate design of the reactor coolant system with a vacuum skid attached, and for failure to adequately perform design verification functions (Section E1.1).
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A violation was identified for failure to properly classify as significant a condition adverse to quality, and for failure to identify the cause of a significant condition adverse to quality involving the loss of valid reactor vessel level indication during reduced inventory operations (Section E8.1).
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Radiological Control, Security, and Fire Protection activities observed were adequately conducted in accordance. with procedures (Sections R1,S1, andF1).
Re ort Details Summa of Plant Status Unit 1 began this inspection period at 100 percent power, and continued at that level until the unit tripped on January 14 due to a personnel error. The unit restarted on January 15, and returned to 100 percent power on January 17. The unit remained at 100 percent power through the end of the inspection period.
I. 0 erations
Conduct of Operations 01.1 General Comments ai Ins ection Sco e 71707 The inspectors conducted frequent reviews of ongoing plant operations including control room tours, shift turnovers, and observation of operations surveillance activities.
b.
Gbservations and Findin s In general, the conduct of operations was professional and safety-conscious.
Routine activities were adequately performed.
Operations shift crews were appropriately sensitive to plant equipment conditions and maintained a questioning attitude in relation to unexpected equipment responses.
Equipment condition, housekeeping, clearances, and log keeping were found to be accomplished in accordance with plant procedures.
Control room staffing was maintained in accordance with technical specification requirements.
C.
Conclusions During the period, the conduct of operations was in accordance with applicable procedures.
01.
- a.
Ins ection Sco e 93702 The inspectors observed operator and equipment response associated with a re'actor trip that occurred on January 14. The inspectors also observed the licensee's initial investigation into the cause of the reactor tri b.
Observations and Findin s The inspectors observed that in response to the reactor trip, operators followed the emergency operating procedures to diagnose the event and stabilize the plant. The initial investigation by the operations staff into the cause of the trip revealed that maintenance technicians were performing preventive maintenance testing on a relay for a nonsafety-related electrical bus. The technicians inadvertently left a test device in the relay when it was placed back in the plant causing an invalid undervoltage signal to strip the loads from the electrical bus. The loads included the "A"reactor coolant pump and a bus powering the "C" reactor coolant pump. The loss of the reactor coolant pumps caused the reactor trip. The inspectors found that safety systems performed as designed.
The inspectors will review the trip in more detail after the post-trip review report is prepared and the licensee event report is submitted.
c.
Conclusions
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Following a reactor trip on January 14, operators followed emergency operating procedures and equipment performed as designed.
The trip was caused by an improperly performed preventive maintenance activity on a nonsafety-related electrical bus relay.
Operational Status of Facilities and Equipment 02.1 En ineered Safe Feature S stem Walkdowns 71707 The inspectors walked down safety-related accessible portions of the main feedwater system.
Equipment configuiation, material condition, and housekeeping were acceptable.
One minor discrepancy was brought to the licensee's attention, and was addressed in accordance with site procedures.
The inspectors found that equipment in the feedwater system was aligned and operating in accordance with procedures and as described in the Final Safety Analysis Report.
Quality Assurance in Operations 07.1 General Comments a.
Ins ection Sco e 40500 71707 During the inspection period, the inspectors reviewed multiple licensee quality assurance activities, including:
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Condition Reports
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Plant Nuclear Safety Committee (PNSC) meeting on January 2 Observations and Findin s During the January 27 meeting, the inspectors observed that the PNSC quorum requirement was met. The inspectors also observed that as the PNSC considered the Significant Adverse Condition Evaluation report for Condition Report (CR) 98-03025-3, the PNSC asked probing questions, determined that the report was not complete, directed the team that prepared the report to expand their investigation, and tabled further consideration of the report until the next meeting. The inspectors considered that the PNSC discussion of this report indicated that PNSC members had thoroughly and critically reviewed the report.
C.
Conclusions PNSC members thoroughly and critically reviewed a detailed report of a significant adverse condition.
Miscellaneous Operations Issues (92700, 92901)
08.1 Closed LER 50-400/98-007-00:
Turbine control anomaly causes a manual reactor trip.
On October 23, 1998, while the licensee was conducting a controlled shutdown of the unit, all four of the main turbine governor valves unexpectedly went full open, and the licensee subsequently tripped the reactor and main turbine manually. The reactor trip itself was discussed in section 01.2 of NRC Inspection Report (IR) 50-400/98-09, and the inspectors'nitial review of this LER was described in section 08.1 of NRC IR 50-400/98-10.
The inspectors reviewed the corrective actions described in the LER, and confirmed that those actions have been successfully completed.
The inspectors considered that those actions were adequate to correct the cause of the event, and should prevent recurrence of the event.
The inspectors noted that procedure OMM-004, "Post-trip/Safeguards Actuation Review,"
Revision 10, implemented the requirements of Technical Specification (TS) 6.8.1, Regulatory Guide 1.33, and ANSI N18.7-1972/ANS-3.2, with respect to determining the circumstances associated with a reactor trip, analyzing the cause of the trip, and determining that operations can proceed safely before the reactor is returned to power after a trip.
The inspectors observed that the Post Trip/Safeguards Actuation Report (PTSAR) for the October 23, 1998, trip was presented for PNSC approval on November 20 (in PNSC meeting 98-54), and that the PNSC approved it with minor comments.
The corresponding meeting minutes indicated that=the PNSC concluded that "the plant responded as expected...".
However, the inspectors found that the PTSAR included a statement that the steam-dump valves had closed when reactor coolant system (RCS)
average temperature reached 544 F. It also stated that "the most important factors that impact the magnitude of the RCS cooldown are the initial RCS temperatur'e and the time it takes the steam dump valves to fullyshut." The inspectors questioned those statements, because they implied that the steam-dump system had failed to operate as
expected, while the report had not identified any problems with the steam-dump system.
(The steam-dump system operates to bring RCS average temperature to 557'F, and its control system is designed to shut the steam-dump valves completely when RCS temperature falls below 553'F.)
Furthermore, the inspectors considered the duration and magnitude of auxiliary feedwater (AFW) flowto be a major influence on the magnitude of the cooldown, and noted that the licensee's explanation of the cooldown did not even mention AFW flow. In addition, the inspectors noted that a strip-chart record of AFW flowwas not included in the PTSAR.
After the inspectors questioned the licensee about these issues on November 20, the licensee initiated a supplemental study of the post-trip cooldown rate. The inspectors observed that the results of the supplemental study of the post-trip cooldown rate were presented to the PNSC on November 25 (in PNSC meeting 98-55). The inspectors noted that those results indicated that the steam-dump system had operated properly during the October 23 post-trip transient, and had no adver'se impact on the magnitude of the RCS 'cooldown, thereby validating the inspectors'oncerns.
The inspectors noted that the PNSC concurred with the results of the supplemental study, and that the plant manager designated those results as "Addendum 1 to OMM-004 for 10/23/98 Trip."
As discussed above, the PTSAR approved by the PNSC on November 20 was not technically adequate.
The inspectors considered that this finding indicated that the November 20 version of the PTSAR had not effectively determined whether all important equipment had operated as expected during that transient, yvhether'the event had any detrimental effect on plant equipment, and whether conditions had been acceptable for restart of the reactor. The inspectors considered that failure to produce an adequate PTSAR was a violation of TS 6.8.1 as implemented by section 5.2.4 of OMM-004, and designated this as VIO 50-400/98-11-02, failure to effectively implement the post-trip review procedure, example 1. The inspectors considered that this violation indicated that plant management did not effectively assure that the PTSAR was technically adequate.
The inspectors'eview of the PTSAR revealed several details in the PTSAR that did not satisfy OMM-004 requirements.
Those details are summarized in the following table:
OMM-004 Requirement (1.2) The PTSAR shall include copies of the ten stripcharts identified in Attachment 13 (5.2.2) The Immediate Incident Investigation Team (IIIT)shall reconstruct a chronological description of the event that includes
"initial, minimum and maximum values of pertinent process parameters."
PTSAR Detail The PTSAR included copies of only two stripcharts.
One of those (PR-0475, steam generator pressure) is among the ten identified in Attachment 13, while the other (ER-0569, main generator output) is not.
The chronological description included in the PTSAR did not include initial, minimum and maximum values of pressurizer level, steam generator levels, charging flow, letdown flow, average RCS temperature, and auxiliary feedwater flow, ail of which were pertinent to the transien \\
(5.2.1) The Superintendent
- Shift Operations (SSO) shall designate an Immediate Incident Investigation Team (IIIT)to perform an investigation.
(5.4)'The PTSAR shall be reviewed by a Follow-up Review Committee convened by the Incident Manager, and the Follow-up Review Committee-shall prepare a Follow-up Review Report.
On the PTSAR that had been approved by the PNSC on November 20, no one had been identified as a member of the IIIT.
No Incident Manager was named, no Follow-up Review Committee was formed, and no Follow-up Review Report was prepared, until after the inspectors identified this issue.
The inspectors examined the vaulted records of several earlier reactor trips to determine whether the licensee had previously completed follow-up reviews. The inspectors examined the OMM-004 records associated with the reactor trips that occurred on September 3, 1996; April25, 1996; January 31, 1997; and July 20, 1997, and found that none of those records included evidence that a follow-up review had been completed, The inspectors also noted that:
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Section 3 of OMM-004 requires that the PTSAR be reviewed by several designated personnel, including the Senior Reactor Operator, the Superintendent-Shift Operations, the Engineering Manager, and the Operations Manager.
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On November 20, the "Report Reviews" section of attachment 1 to OMM-004 indicated that the PTSAR had not been reviewed by those personnel or their designees on November 18.
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Neither the reviewers who signed the PTSAR on November 18 nor the PNSC members who reviewed it on November 20 identified any of the inadequacies described above.
The inspectors concluded that none of the required reviews had been adequate to assure that the PTSAR had been completed in accordance with OMM-004, and that the intent of OMM-004 had been satisfied.
The inspectors considered the failure to complete the PTSAR in accordance with OMM-004 requirements, conduct adequate reviews of the PTSAR, and complete follow-up reviews to be a violation of TS 6.8.1 as implemented by OMM-004. The inspectors designated these failures as VIO 50-400/98-11-02, failure to effectively implement the post-trip review procedure, example 2. The inspectors considered that this violation indicated that management did not effectively implement the post-trip review process in accordance with OMM-004 requirements.
The inspectors considered that the results of the supplemental study indicated that an inadequate PTSAR had been presented to the PNSC on November 20. The inspectors noted (as described above) that required reviews of the PTSAR had not identified
inadequacies in the PTSAR, and were therefore also inadequate.
The inspectors considered that production of an inadequate PTSAR and completion of inadequate reviews constituted an adverse condition as defined in procedure CAP-NGGC-0001,
"Corrective Action Management," Revision 1. The inspectors considered that the licensee implicitlyacknowledged that adverse condition, by initiating the supplemental study, by concurring with its results, and by revising the PTSAR to incorporate those results; and therefore should have initiated a CR as required by CAP-NGGC-001.
However, the inspectors noted that the licensee did not initiate a CR. The inspectors considered the failure to initiate a CR to be a violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," as implemented by procedure CAP-NGGC-0001, and have designated this as VIO 50-400/98-11-01, failure to promptly identify and correct a condition adverse to quality, example 2.
II. Maintenance M1 Conduct of Maintenance M1.1 General Comments a.
Ins ection Sco e 62707 The inspectors observed all or portions of the following work activities to determine whether the activities were being performed in accordance with approved procedures and regulatory requirements, personnel were appropriately trained and qualified, and appropriate radiological controls were being followed:.
WR/JO
~Descri tion 98-AISA2 &
repair a plugged pressure-tap indicator on the emergency service 98-AISB2 water screen wash pump 98-AHUR1 rebuild the accumulator air pump on the feedwater preheater bypass isolation valve 98-AISZ1 investigate/repair the recirculation valve on train "A"of the residual heat removal system, which failed to open on low flow 98-AHNR1 repair the packing leak on the crossover isolation valve between low-head safety-injection train "B" and the hot-leg injection path AEQN 003 perform routine maintenance on the diesel generator starting air compressor b.
Observations and Findin s The inspectors found the work performed under these activities to be professional and thorough.
Allwork observed was performed with the work package present and in active use. Technicians were experienced and knowledgeable of their assigned tasks.
The
inspectors frequently observed supervisors and system engineers monitoring job progress, and noted that quality control personnel were present whenever required by procedure.
Peer-checking and self checking techniques were being used.
When applicable, appropriate radiation control measures were in place.
c.
Conclusions The maintenance tasks observed were conducted adequately and in accordance with procedures and approved work plans.
M2 Maintenance and Material Condition of Facilities and Equipment M2.1 Surveillance Observation The inspectors observed all or portions of the following surveillance tests:
MST l0125
"Main Steamline Pressure, Loop 1 (P-0474), Operational Test,"
Revision 5 OST-1048
"Fuel Handling Building Emergency Exhaust System Operability 18 Month Interval At AllTimes," Revision 9 EST-400
"Engineered Safety Feature Air Filtration Testing," Revision 8 The inspectors observed that the tests were conducted in accordance with the subject procedures, and were conducted by knowledgeable personnel.
M7 Quality Assurance in Maintenance Activities M7.1 Maintenance Rule a.
Ins ection Sco e 62707 b.
The inspectors conducted a risk-informed, performance-based review and assessment of several equipment malfunctions identified during the unit's response to the October 23, 1998, reactor trip to determine whether procedure ADM-NGGC-0101, "Maintenance Rule," Revision 9, was being followed.
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Observations and Findin s The inspectors noted that the Significant Adverse Condition Report for CR 98-2630 listed equipment problems that had been identified during the unit's response to the October 23, 1998, reactor trip: From among the problems identified in that report, the inspectors selected the following for this review and assessment:
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Feedwater regulating valves A, B, & C closed to approximately 10 percent open, and did not fullyshut;
~, The moisture-separator-reheater controller did not reset;
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The "C" power-operated relief valve opened at a pressure below its setpoint; and
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The digital electro-hydraulic control computer generated an erroneous demand signal.
The inspectors found that, following the reactor trip the licensee accomplished numerous maintenance rule monitoring activities. These activities included function and scoping determinations, determinations of safety significance, unavailability tracking, and functional failure decisions.
The inspectors found that these activities had been completed in accordance with procedure ADM-NGGC-0101, with one possible exception: the inspectors noted that in the maintenance-rule database, the system engineer had not designated the multiple failures of the feedwater regulating valves as being repetitive failures, even though they appeared to satisfy the procedure's definition of repetitive failures. The inspectors also noted that:
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Designating those failures as repetitive would require that the performance monitoring group which includes the feedwater regulating valves be classified a(1),
and that a condition report describing the failures be classified as "Significant Adverse." The inspectors noted that the subject performance monitoring group was already classified a(1), and that the condition report describing the failures (CR 98-02630) was already classified as "Significant Adverse."
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The licensee had previously scheduled an expert panel meeting for February 11, 1999, to discuss maintenance rule activities associated with the performance monitoring group that contained functions associated with the feedwater regulating valves, and that the subject failures would be reviewed in that meeting.
The inspectors therefore concluded that the licensee had not yet completed the
'lassification of the subject failures, and that delaying that classification until after the expert panel meeting did not have a significant impact.
Conclusions With respect to the equipment problems identified during the unit's response to the October 23 reactor trip, the licensee had implemented and was continuing to implement the maintenance rule program in accordance with approved procedure III. En ineerin E1 Conduct of Engineering E1.1 En ineerin Service Re uests a.
Ins ection Sco e 37551 As part of the inspection related to the licensee's attempt to vacuum-fill the RCS (described in section E8.1 of this report), the inspectors reviewed Engineering Service Request (ESR) 94-00099, "RCS vacuum refill,"Revision 10, to determine whether procedure EGR-NGGC-0005, "Engineering Service Requests," Revision 9, had been followed.
b.:Observations and Findin s
The inspectors noted that the ESR Closeout Form (form EGR-NGGC-0006-11-0) for the subject ESR had been signed, indicating that activities associated with that ESR had been completed, on December 8, 1998. As discussed in section E8.1 of this report, in response to the inspectors'indings, the licensee initiated a Significant Adverse Condition Evaluation of CR 98-03025, and found that the design described in ESR 94-00099 was inadequate, in that the design caused a local vacuum to be sensed by the reactor vessel water level instrumentation, thereby effectively disabling that instrumentation.
The inspectors noted that 10 CFR 50, Appendix B, Criterion III requires, in part, that measures shall be established to assure that applicable regulatory requirements are correctly translated into specifications, drawings, procedures, and instructions, and that measures shall provide for verifying or checking the adequacy of design.
The inspectors also noted that the licensee was committed to 10 CFR 50 Appendix A, General Design Criterion 13 in FSAR Section 3.1.9 which required, in part, that instrumentation be provided to adequately monitor reactor vessel water level. The RCS vacuum refill modification did not adequately translate this requirement into the design.
The inspectors reviewed the design described in the subject ESR, and noted that it included connections between a vacuum manifold and all available RCS water level instruments, such that a vacuum established in the manifold would be communicated to all level instruments.
As noted above, the licenseefound that when the vacuum manifold was placed in service, those connections effectively disabled all level instruments.
The inspectors noted that with the RCS in a reduced-inventory configuration, inadequate water level could lead to loss of decay heat removal capability from the reactor core. The inspectors therefore considered that RCS water level was a variable that could indirectly
affect the integrity of the reactor core. Because the subject design did not provide instrumentation to accurately monitor RCS water level, the inspectors concluded that the subject design had not correctly translated the regulatory requirement described in GDC 13 into the specifications, drawings, procedures, and/or instructions affected by the ESR.
The inspectors reviewed the design-verification functions associated with the subject design, to determine whether they had been adequately performed.
The inspectors noted that section 9.4.7 of procedure EGR-NGGC-0005 stated that:
"Testing shall verify that: the modified system/component functions/performs as intended; the design change has been correctly implemented; the revised design is correct,"
In ESR 94-00099, Revision 10, the inspectors noted that "Testing Requirements" were described on pages 13.1 through 13.3. On those pages, the inspectors found descriptions of a live pressure test of all tubing and associated tubing joints, a pressure test of the standpipe connections, and a vacuum test of a section of the standpipe.
However, the inspectors found no description of testing that verified either that the modified system (the RCS with the vacuum skid attached) functioned/performed as intended, that the design change had been correctly implemented, or that the design was correct. When questioned by the inspectors, the responsible engineer for the ESR confirmed that only the testing described in pages 13.1 through 13.3 of the ESR had been completed.
The inspectors therefore concluded that the testing which was completed did not adequately verify the design.,
The inspectors also noted that although the subject ESR had been reviewed by several independent reviewers, none of those reviewers had noted either that the design was inconsistent with regulatory requirements, or that testing to verify the design had not been specified.
The inspectors considered that the failure of those reviewers to note those deficiencies was evidence that the reviews themselves had not been adequate to verify or check the adequacy of the design.
The inspectors found that no design verification activities other than the testing and reviews described above had been performed.
For. ESR 94-00099, the inspectors considered the failure of the design to correctly translate reactor vessel water level monitoring requirements into specifications, drawings, procedures, and instructions; and.the failure to adequately perform design verification functions to be a violation of 10 CFR 50, Appendix B, Criterion III, "Design Control", as implemented by EGR-NGGC-0005, and have designated this as VIO 50-400/98-11-03, inadequate design of the reactor coolant system with the vacuum skid attached:
Conclusions A violation was identified for failure to develop an adequate design of the reactor coolant system with a vacuum-fill skid attached, and for failure to adequately perform design-verification function ~
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ES Miscellaneous Engineering Issues (92700, 92903)
E8.1 Closed URI 50-400/98-10-01:
loss of valid reactor vessel level indication during reduced-inventory operations.
As described in section 01.3 of NRC IR 50-400/98-10, the inspectors noted that the licensee had initiated CR 98-03025 to document the unsuccessful attempt to vacuum-fill the RCS during refueling outage 8, and that the licensee had completed an "Adverse Condition Evaluation" of that CR on December 3.
The inspectors noted that 10 CFR 50, Appendix B, Criterion XVI, requires, in part, that conditions adverse to quality are promptly identified and corrected.
With respect to that criterion, the inspectors'eview of the Adverse Condition Evaluation for CR 98-03025 revealed that:
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CR 98-03025 had been classified as an "Adverse Condition." However, the inspectors considered that while the unit was operating in a reduced-inventory configuration (as it was during the subject event), indicated reactor vessel water is a key plant parameter, because it relates directly to the ability to remove decay heat from the reactor core.
Furthermore, the inspectors had observed that during the
, subject event, indicated reactor vessel water level had experienced rapid variations with a magnitude of approximately 40 inches in a relatively short time, and considered those variations to be outside normal plant variances.
Because the subject event had involved a significant change outside normal plant variances in a key plant parameter, the inspectors therefore considered that the subject event had satisfied one of the criteria in the "Criteria for Significant Adverse Conditions" (Attachment 1 of procedure CAP-NGGC-001) for the CR to be classified as a "Significant Adverse Condition."
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The only corrective action described in the Adverse Condition report for CR 98-03025 was to revise GP-001, Revision 14, to ensure that the reactor vessel head was vented while draining down the vessel level. The inspectors considered that the failed attempt had shown that GP-001 and/or. the ESR 94-00099 design had been inadequate, and that the Adverse Condition evaluation should therefore have considered whether and why GP-001 had been inadequate, and whether and why the design had been both adequate and properly implemented.
However, the inspectors observed that the report did not describe those topics; it described only the failure of GP-001 to include instructions for venting the vessel head.
The inspectors therefore considered that the Adverse Condition evaluation was not broad enough in scope to determine the cause(s) of the event.
After the inspectors identified the inadequacies described above, the licensee re-opened CR 98-03025 and initiated a Significant Adverse Condition Evaluation of the event. The inspectors observed that the team conducting that investigation presented its report to the PNSC on January 27. That team found that the ESR 94-0099 design was not valid, and that Revision 14 of GP-001 did not correctly implement that design.
(GP-001 had incorrectly implemented an inadequate design.)
Therefore, the inspectors considered that the licensee's own investigation of the event had proven that the Adverse Condition report for CR 98-03025 had failed to identify the cause, and that the corrective actions described in that report were inadequate to preclude repetitio ~
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The inspectors considered the failure to properly classify CR 98-03025 and the failure to determine the cause of an identified significant adverse condition to.be a violation of 10 CFR 50, Appendix B, Criterion XVI,as implemented by procedure CAP-NGGC-001, and have designated this as VIO 50-400/98-11-01, failure to promptly identify and correct conditions adverse to quality, example 1.
The inspectors noted that VIO 50-400/98-11-01 is similar. to VIO 50-400/98-03-01, in that both violations involved the failure to properly classify condition reports.
The inspectors considered that the corrective actions taken in response to VIO 50-400/98-03-01 reasonably could have prevented VIO 50-400/98-11-01.
In accordance with Section IV.B of NUREG-1600, "Statement of Policy and Procedure for NRC Enforcement Actions", the inspectors therefore considered VIO 50-400/98-11-01 to be repetitive.
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After the inspectors described the findings associated with VIO 50-400/98-11-01 to the licensee, the inspectors observed that the licensee initiated and completed several major changes to improve the site Corrective Action Program (CAP). Those changes included:
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The. licensee established a daily meeting of selected members of the station management team to conduct reviews of newly-initiated CRs, and to review classification and assignment of previously-initiated CRs;
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The licensee elevated the role of unit evaluators, to ensure that their CAP function is their highest priority;
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The licensee established a daily meeting of the CAP unit evaluators, to classify and assign newly-initiated CRs; and
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The licensee established daily feedback from CAP unit evaluators to their managers regarding newly-initiated CRs and other key CAP activities.
IV. Plant Su ort R1 Radiological Protection and Chemistry (RP&C) Controls R1.1 General Comments 71750 The inspectors found radiological controls to be acceptable during routine tours of the plant. The general approach to the control of contamination and dose for the site was good. Teamwork between the various departments continued to be a major contributor to the good control of dose.
NRC Form 3 was posted at the appropriate places.'rimary and secondary chemistry were maintained within technical specification limit Conduct of Security and Safeguards Activities S1.1 General Comments 71750 The inspectors observed security and safeguards activities during the conduct of plant tours including general integrity of the protected area barrier, maintenance of the isolation zones, illumination levels, access control, and vital area controls.
The inspectors found that equipment was properly maintained, that security activities were conducted in accordance with the security plan, and that, when necessary, compensatory measures were posted and were properly conducted.
F1 Control of Fire Protection Activities F1.1 General Comments 71750 The inspectors observed fire protection equipment and activities during the conduct of tours and observation of maintenance activities and found them accomplished in accordance with required procedures.
V. Mana ement Meetin s X1 Exit,Meeting Summary The inspectors presented the inspection results to members of licensee management at the conclusion of the inspection on February 5, 1999. The licensee acknowledged the findings presented.
The inspectors asked the licensee whether any of the material examined during the inspection should be considered proprietary.
No proprietary information was identifie PARTIALLIST OF PERSONS CONTACTED Licensee D. Alexander, Manager, Regulatoty'ffairs J. Bates, Superintendent, Environmental and Chemistry D. Batton, Superintendent, On-Line Scheduling D. Braund, Superintendent, Security B. Clark, General Manager, Harris Plant A. Cockerill, Superintendent, l&C Electrical Systems J. Collins, Manager, Maintenance J. Cook, Manager, Outage and Scheduling J. Curley, Maintenance Rule Coordinator J. Eads, Supervisor, Licensing and Regulatory Programs R. Field, Manager, Nuclear Assessment M. Keef, Manager, Training G. Kline, Manager, Harris Engineering Support Services R. Moore, Manager, Operations K. Neuschaefer, Superintendent, Radiation Protection
'.
Scarola, Vice President, Harris Plant NRC S. Flanders, Harris Project Manager, NRR B. Bonser, Chief, Reactor Projects Branch 4 INSPECTION PROCEDURES USED IP 37551:
IP 40500'P 61726:
IP 62707:
IP 71707:
.IP 71750:
IP 92700:
IP 92901:
IP 92903:
IP 93702:
Onsite Engineering Effectiveness of Licensee Controls in Identifying, Resolving, and Preventing Problems Surveillance Observations Maintenance Observation Plant Operations Plant Support Activities Onsite Followup of Events Followup - Plant Operations Followup - Engineering Onsite Response to Events
M. Keef, Manager, Training G. Kline, Manager, Harris Engineering Support Services R. Moore, Manager, Operations K. Neuschaefer, Superintendent, Radiation Protection J. Scarola, Vice President, Harris Plant NRC S. Flanders, Harris Project Manager, NRR B. Bonser, Chief, Reactor Projects Branch 4 INSPECTION PROCEDURES USED
IP 37551:
IP 40500:
IP 61726:
IP 62707:
IP 71707:
IP 71750:
IP 92700:
IP 92901:
IP 92903:
IP 93702:
Onsite Engineering Effectiveness of Licensee Controls in Identifying,'esolving, and Preventing Problems Surveillance Observations Maintenance Observation
.
Plant Operations Plant Support Activities Onsite Followup of Events Followup - Plant Operations Followup - Engineering Onsite Response to Events
-C
~Oened
- 50-400/98-11-01 50-400/98-11-02 50-400/98-11-03 Closed 50-400/98-007-00 50-400/98-10-01 ITEMS OPENED, CLOSED, AND DISCUSSED VIO failure to promptly identify and correct a condition adverse to quality (2 examples) (Sections 08.1 and E8.1).
VIO failure to effectively implement the post-trip review procedure (2 examples) (Section 08.1).
VIO inadequate design of the reactor coolant system with the vacuum skid attached (Section E1.1).
LER turbine control anomaly causes a manual reactor trip (Section 08.1).
URI loss of valid reactor vessel level indication during reduced-inventory operations (Section E8.1).