IR 05000397/1999004
| ML17292B681 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 05/28/1999 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML17292B680 | List: |
| References | |
| 50-397-99-04, 50-397-99-4, NUDOCS 9906070047 | |
| Download: ML17292B681 (28) | |
Text
ENCLOSURE U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Docket No.:
License No.:
Report No.:
Licensee:
Facility:
Location:
Dates:
Inspectors:
Approved By 50-397 NPF-21 50-397/99-04 Washington Public Power Supply System Washington Nuclear Project-2 Richland, Washington March 21 through May 1, 1999 D. E. Corporandy, Acting Senior Resident Inspector J. E. Spets, Resident Inspector G. A. Pick, Senior Project Engineer Linda J. Smith, Chief, Project Branch E, Division of Reactor Proje'cts ATTACHMENT:
Supplemental Information 9906070047
'F90528 PDR ADGCK 05000397
EXECUTIVE SUMMARY Washington Nuclear Project-2 NRC Inspection Report No. 50-397/99-04 This information covers a 6-week period of resident inspection.
~oerations Key managers as well as quality assurance personnel were present in the control room to monitor the shutdown, which was conducted in a safe and deliberate manner.
Communications were good.
Supervisory oversight and direction of the operating crew and operator performance during the shutdown were good (Section 01.1).
The design basis of the residual heat removal system did not support the full range of applicability for Technical Specification, Limiting Condition of Operation 3.4.9, "Residual Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown" and the associated Technical Specification Bases.
The design basis was also inconsistently implemented in procedures and in instructions for the residual heat removal system in the shutdown cooling mode of operation.
Because the licensee is continuing to research the design basis for the system and because additional information is required on (1) related accident analysis assumptions, (2) generic implications, (3) prior system evaluations, and (4) notification, the issue is being identified as an unresolved item (Section 02.1).
The inspectors'uestions about the adequacy of control room instrumentation color banding were promptly addressed by the operators.
In addition, the operators demonstrated a good questioning attitude and a desire to resolve the issue (Section 07.1).
Corrective actions resulting from a 1996 problem evaluation request were never completed.
The problem evalu'ation request had been generated to address the failure to resolve control room design deficiencies associated with color banding of control'oom instrumentation, as required by License Condition 16. The problem evaluation request was closed and the work order to resolve the color banding issue was canceled during a backlog item review, without evaluating the work order cancellation for conflict with the license condition. This problem is a violation of 10 CFR Part 50, Appendix B, Criterion XVI;however, this Severity Level IV violation is being treated as a noncited violation, and is in the licensee's corrective action program as Problem Evaluation Request 299-0745 (Section 07:1).
Maintenance Maintenance work observed by the inspectors was conducted in a manner that ensured reliable, safe operation of the station (Section M1.1).
Surveillance testing was generally conducted in accordance with the licensee's programs and Technical Specifications.
The licensee had appropriately stringent expectations for measuring as-found setpoints (Section M1.2).
-2-Plant housekeeping and material condition were generally good; however, the inspectors found an unsecured portable eye wash station too close to the high pressure core spray batteries in violation of procedural requirements.
This is one example of a Severity Level IVviolation of Technical Specification 5.4.1.a, which is being treated as a noncited violation and is in the licensee's corrective action program as Problem Evaluation Request 299-0889 (Section M2.1).
The inspectors identified a procedure weakness that allowed potential interferences between scaffolding and instrument sensing lines to be evaluated by the craft erecting the scaffolding. This was inconsistent with other guidance in the procedure which required engineering evaluation and a 10 CFR 50.59 review for potential interferences between scaffolding and important-to-safety components.
At the close of the inspection, engineeiing was planning to revise the scaffolding procedure to ensure that potential interferences with instrument sensing lines will receive a similar degree of evaluation as other safety-related components (Section M3.1).
~En ineerin
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Technical Specification 3.3.6.1, "Primary Containment Isolation Instrumentation,"
Function 5, "Residual Heat Removal Shutdown Cooling System Isolation," and the associated bases section were incorrect. The Technical Specifications were not updated when the controls for the outboard isolation valve were removed from the alternate remote shutdown panel.
In addition, the bases section incorrectly stated that there are four pressure switches associated with the reactor high pressure isolation instrumentation, when only two exist. This issue is identifed as an unresolved item because additional information is required in order to confirm the facilitywas originally licensed with only two pressure switches and to review the 10 CFR 50.59 evaluation for the change (Section E1.1).
Plant Su ort Radiological controls associated with the unloading of fresh fuel were gene'rally good and health physics oversight helped personnel maintain exposure ALARA. However, the licensee failed to post or mark a contaminated area as required by procedure.
This is one example of a Severity Level IV violation of Technical Specification 5.4.1.a and is being treated as a noncited violation. This deficiency is in the licensee's corrective action program as Problem Evaluation Request 299-0718 (Section R1.1).
Calibration of radiation monitoring equipment at WNP-2 was being performed within the required calibration frequencies (Section R1.2).
Re ort Details Summa of Plant Status The station operated at essentially 100 percent power and maintained the capacity to operate at 100 percent power from the beginning of the inspection period until April 16, 1999.
During that period of time through April 11, the station routinely reduced power to 85 percent during the evening hours and 65 percent over the weekends at the request of the Bonneville Power Administration for economic dispatch.
The licensee had planned to shut down for its fuel savings dispatch outage on April 10; however, at the request of Bonneville Power Administration, the licensee delayed the shutdown and remained at 85 percent power until April 16. On April 17, the unit was shut down for its fuel savings dispatch outage.
Unit 2 remained in Mode 4 through the end of the inspection period.
I. OPERATIONS
Conduct of Operations 01.1 Plant Shutdown a.
Ins ection Sco e 71707 On April 16 and 17, 1999, the inspectors observed the control room staff perform a planned reactor shutdown in preparation for a fuel savings dispatch outage.
b.
Observations and Findin s At 2 p.m. on April 16, operators started a normal plant shutdown per Procedure 3.2.1,
"Normal Shutdown to Cold Shutdown," Revision 37. Key managers and quality assurance personnel were present in the control room to monitor the shutdown..
Approximately 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> elapsed from the commencement of shutdown until the reactor was scrammed at 5:14 a.m on April 17. The shift managers and control room supervisors for both crews took the time to minimize crew distractions from any work other than that pertinent to the shutdown.
The control room supervisors reminded extra personnel (for example, quality assurance personnel, station managers, and the NRC),
when necessary, of the importance of not obstructing the operators'iew of their boards.
At critical times, such as the reactor trip, unnecessary communications into the control room were suspended.
Control room communications were generally good during the shutdown.
The inspectors observed good three-way communications between operators for most exchanges of important information. The turnovers observed by the inspectors at the change of shift appeared to be thorough and accurate.
Briefings were conducted before starting critical tasks.
At the briefings, task responsibilities were clearly assigned, and contingency plans were discussed where problems were known to have occurred in the past.
For example, during some recent shutdowns, immediately following the reactor scram, control rod position indication of several control rods was slow to respond.
The inspectors noted that the licensee had apparently identified the cause of the slow-to-respond control rod position indicators as faulty position indicating probes and
0'
-2-was planning to fixthem during the spring/summer 1999 outages.
This problem was discussed during the briefing immediately prior to the reactor scram.
Operators were reminded of the anticipated transient without scram contingency procedure and were given the time to have the procedure out and reviewed prior to initiating the reactor scram.
In fact, during the shutdown, four control rods did not indicate fully in immediately after the scram; however, operator response was fast and in accordance with procedures.
Allcontrol rods indicated fully in within 2 minutes of the scram.
When problems were encountered, the appropriate time was taken to ensure that conditions were in accordance with the shutdown plan. For example during shutdown, while removing second stage reheat, the 10-inch diameter main steam supply valves did not shut.
Operators investigated this situation and found that the valve position
'ndication for Valves MS-TCV-116A and MS-TCV-116C did not indicate fullyclosed, a prerequisite for closing the 10-inch diameter main steam supply valves.
Once Valves MS-TCV-116A and MS-TCV-116C were verified closed, steps were taken to close the 10-inch diameter main steam supply valves and continue with the shutdown procedure.
Proceeding in a cautious, deliberate manner was emphasized.
Conclusions Key managers as well as quality assurance personnel were present in the control room to monitor the shutdown, which was conducted in a safe and deliberate manner.
Communications were good. Supervisory oversight and direction of the operating crew and operator performance during the shutdown were good.
02.1 Operational Status of Facilities and Equipment Shutdown Coolin
- Hot Standb Technical S ecifications and 0 eratin Procedures Ins ection Sco e 71707 The inspectors walked down control room panels and questioned operators, a shift supervisor, and a system engineer on the use and bases of several labels.
In addition, the inspectors reviewed Procedure 2.4.2, "Residual Heat Removal System,"
Revisions 35 and 38; Procedure 4.12.1.1, "Control Room Evacuation and Remote Cooldown," Revisions 32 and 33; Procedure 3.2.1, "Normal Shutdown to Cold
.Shutdown," Revisions 36 and 37; the Technical Specifications; and the Technical Specifications Bases related to posted labels. The inspectors also reviewed several problem evaluation request generated from the inspector's findings.
Observations and Findin s The inspectors identified the following problems regarding the residual heat removal (RHR) system in the shutdown cooling mode of operation:
(1) Technical Specification Limiting Condition of Operation 3.4.9, "Residual Heat Removal Shutdown Cooling System - Hot Shutdown," and its associated basis could not be met; (2) operating procedures did not reflect RHR system limitations; and (3) two labels on Control Room Panel H13-601 were in erro The inspectors noted that a label on Control Room Panel H13-P601 read "RHR shutdown cooling shall not be entered with reactor pressure above 48 psig except for post accident shutdown cooling, due to excessive pressure across RHR-V-9" (Valve RHR-V-9 is the inboard isolation valve for the RHR shutdown cooling suction line). A similar caution existed for Valve RHR-V-8 (the outboard isolation valve for RHR shutdown cooling suction line). The inspectors questioned the operators and shift supervisor as to why the initiation of shutdown cooling was not allowed greater than 48 psig when the cut-in permissive for the system was set at approximately 125 psig.
The crew stated that they did not know the exact reason for the limit, but that it appeared to be because of limitations on the valves.
The inspectors reviewed Procedure 2.4.2 and found that the system was limited to 48 psig because it corresponded to a saturation temperature of 295 F, which was the actual limiting parameter.
Procedure 2.4.2 specifically stated "295 F 48 psig this limit set by maximum temp RHR piping between HX [heat exchanger] and RV [reactor vessel] can stand." The system engineer informed the inspectors that portions of the RHR system were not analyzed at temperatures greater than 295'F and that there were thermal transient restrictions on the system; therefore, the limitwas established.
However, the inspectors found that tw'o procedures and the control room labels did not restrict operation of the RHR system in the unanalyzed area.
Specifically, Procedure 4.12.11 allowed entry into shutdown cooling at less than 135 psig (saturation temperature at 135 psig is 358 F), and Procedure 2.4.2 allowed entry into shutdown cooling at greater than 48 psig postaccident.
The licensee issued revisions to the procedures to address the issue.
In addition, the licensee initiated PER 299-0691 to address a general labeling problem within the plant.
The inspectors also found that Technical Specification 3.4.9, "Residual Heat Removal Shutdown Cooling System - Hot Shutdown," requires two RHR shutdown cooling subsystems to be operable when in Mode 3 with reactor steam dome pressure less than the RHR cut-in permissive pressure.
However, because of the design limitations associated with the shutdown cooling mode of the RHR system, it could not be operable when required by the Technical Specification (i.e., Mode 3 with steam dome pressure less than RHR cut-in permissive pressure).
The licensee initiated PER 299-0574 to address the problem.
In addition, because the licensee required the shutdown cooling mode of the RHR system to support an upcoming reactor shutdown, Procedure 3.2.1 was revised as follows: "NOTE: RHR Shutdown Cooling [SDC] is required to be operable at the RHR cut-in permissive pressure, however RHR SDC can not be placed in service until RPV [reactor pressure vessel] pressure is LE[s]48 PSIG... When the RHR SDC permissive is satisfied (approximately 125 PSIG), enter RHR Shutdown Cooling as inoperable..."
The inspectors found that:
(1) Technical Specification Limiting Condition of Operation 3.4.9 and its associated basis; (2) Procedures 2.4.2, 3.2.1, and 4.12.1.1; and (3) labels for the RHR system did not reflect the design basis of the RHR system.
In addition, because additional information is required on other related issues such as:
(1) accident analysis, (2) generic implications, (3) prior system evaluations, and (4) notification, the issue is identified as an unresolved item (50-397/99004-01).
This unresolved item is also discussed in Section E Conclusion The design basis of the RHR system did not support the full range of applicability for Technical Specification, Limiting Condition of Operation 3.4.9, "Residual Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown" and the associated Technical Specification Bases.
The design basis was also inconsistently implemented in procedures and in instructions for the RHR system in the shutdown cooling mode of operation.
Because the licensee is continuing to research the design basis for the system and because additional information is required on (1) related accident analysis assumptions, (2) generic implications, (3) prior system evaluations, and (4) notification, the issue is being identified as an unresolved item.
Quality Assurance in Operations 07.1 Inade uate Corrective Actions Associated with Color Bandin of Instrumentation Ins ection Sco e 71707 The inspectors walked down control room panels, observed the performance of high pressure core spray (HPCS) service water surveillance testing, and questioned operators and shift supervisors on instrumentation color bands.
In addition, the inspectors reviewed Operating Instruction (OI)-45, "Main Control Room Meter Banding,"
PERs relating to color banding, and related commitments to the NRC.
Observations and Findin s The inspectors found that actions to correct inadequacies associated with color banding of instrumentation in the control room were inadequate.
Specifically, the corrective actions associated with PER 296-0490 were never completed as required.
The inspectors observed operators perform a portion of HPCS service water surveillance testing and noted that Service Water Pump HPCS-P2 discharge pressure was above the green band established on the instrument.
The operators stated that the pump pressure was probably above the green band because of the surveillance lineup.
They provided curves from the surveillance procedure to demonstrate that the pump was within its allowed operating curves.
Additionally, the crew demonstrated a good questioning attitude and a desire to fullyresolve the inspectors'uestions and provided OI-45, which dictated banding requirements.
The inspectors and operators noted that the instrument was not banded in accordance with the procedure, which required that banding represent the acceptable and unacceptable operating ranges.
In addition, the inspectors noted that other instrumentation within the control room was not banded as required by Ol-45. For example, reactor pressure Instruments MS-LR/PR623A & 623B did not have blue indicator bands to mark the entry point into emergency operating procedures.
The licensee initiated PER 299-0745 to address the problem and identified that a previous PER (296-0490) had been initiated to address the issue of inadequate control
-5-room instrumentation color banding and that itwas in response to an NRC commitment.
In addition, the PER states that a work order was initiated in 1996 to resolve the issue, but that it was canceled during a backlog item review without evaluating the work order cancellation for conflict with the license conditions.
C.
The inspectors identified a violation of 10 CFR Part 50, Appendix B, Criterion XVI,
"Corrective Action," that states, in part, measures shall be established to assure that conditions adverse to quality, such as deficiencies and deviations, are promptly corrected.
However, the inspectors found that a corrective action associated with PER 296-0490 was never completed.
This Severity Level IVviolation is being treated as a noncited violation, consistent with Appendix C of the NRC Enforcement Policy (50-397/94004-02).
This violation is in the licensee's corrective action program as PER 299-0745.
Conclusion The inspectors'uestions about the adequacy of control room instrumentation color banding were promptly addressed by the operators.
In addition, the operators demonstrated a good questioning attitude and a desire to resolve the issue.
Corrective actions, resulting from a 1996 PER were never completed.
The PER had been generated to address the failure to resolve control room design deficiencies associated with color banding of control room instrumentation, as required by License Condition 16. The PER was closed and the work order to resolve the color banding issue was canceled during a backlog item review, without evaluating the work order cancellation for conflict with the license condition. This problem is a violation of 10 CFR Part 50, Appendix B, Criterion XVI;however, this Severity Level IVviolation is being treated as a noncited violation and is in the licensee's corrective action program as PER 299-0745.
Miscellaneous Operations Issues (92901)
08.1 Administrative Closure of Violations Based U on Chan es in the Enforcement Polic The inspectors reviewed outstanding violations in the Operations area.
The Severity Level IVviolations listed below were issued in Notices of Violation prior to March 11, 1999.
On this date, the NRC implemented a new policy for treatment of Severity Level IVviolations (Appendix C of the Enforcement Policy). Because these violations would have been treated as noncited violations in accordance with Appendix C, they are being closed out in this report. The inspectors verified that the licensee had generated a corrective action program reference (PER) for each of the violations listed.
In addition, these violations already have docketed responses or were generated with no response require Violation Number 50-397/98013-01 50-397/98009-05 50-397/98009-04 50-397/98009-02 50-397/98009-01 50-397/98006-03 50-397/98006-02 50-397/98006-01 Description improper lineup of the low pressure core spray system failure to followprocedure for testing excess flow check valves failure to followprocedure for testing excess flow check valves failure to properly evaluate a partially elevated fuel assembly failure to verify prerequisites for moving heavy loads failure to establish adequate controls for residual heat removal suppression pool return 10 CFR 50.59 and 50:9 violation for maintaining residual heat removal minimum flow valve closed failure to identify and track and inoperable train of control room heating, ventilation, and air conditioning CA.Program Reference 298-0740 298-0695 298-0695 298-0629 298-0424 298-0382 298-0402 298-0632 298-0305 50-397/98005-04 two examples of inadequate procedures 298-0375 298-0223 298-0220 50-397/98005-03 failure to properly implement procedure change 298-0375 50-397/98005-02 failure to meet surveillance requirements for a surveillance procedure 298-0264 50-397/98005-01 failure to provide 10 CFR 50.72 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> notification 298-0231 50-397/98003-05 50-397/98003-04 50-397/98003-03 50-397/98003-02 inadequate corrective action for control of transient combustibles inadequate procedure to address requirements of Technical Specification 5.5.2 improper application of Technical Specification 3.0.2 improper application of Surveillance Requirement 4.0.2 298-0204 298-0144 297-0965 298-0003 298-0003 50-397/97018-07 failure to control transient combustibles 297-0999 50-397/97018-01 failure to properly secure transient or portable equipment 298-0033 297-1025 297-0960 Review of the effectiveness of the corrective actions for selected violations will be performed in the future as a routine part of the review,of the corrective action progra II. MAINTENANCE M1 Conduct of Maintenance M1.1 General Comments - Maintenance a.
Ins ection Sco e 62707 The inspectors observed or reviewed portions of the following work activities:
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Work Order Task No. NMF 103 Cable Pull from J3/14.7 to K2/12.3 for WMA-FN-52B
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Work Order No. PVX 302 Installation of a Strain Gage on RCIC-MO-22 b.
Observations and Findin s With some minor exceptions, maintenance work observed was performed according to the work instructions.
The inspectors observed supervisors and system engineers monitoring job progress, and quality control personnel were present when required. The inspectors found that maintenance was being conducted in a manner sufficient to ensure reliable, safe operation of the station and plant equipment.
c.
Conclusions Maintenance work observed by the inspectors was conducted in a manner that ensured reliable, safe operation of the station; M1.2 General Comments - Surveillance a.. Ins ection Sco e 61726 The inspectors observed or reviewed all or portions of the following test activities:
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OSP-ELEC-M701, Revision 6, "Diesel Generator 1 Monthly Operability Test"
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ICP-RHR-Q901, Revision 1, "RHR SDC Mode High Flow Isolation - CFT/CC" b.
Observations and Findin s The diesel generator surveillance testing was conducted satisfactorily in accordance with the licensee's procedures, programs, and Technical Specifications.
One purpose of the RHR surveillance was to measure the "as-found" setpoint for the shutdown cooling high flow isolation. The inspectors observed an instrument and controls (I&C)technician in training perform this test as part of his certification.
In addition to the second verifier (a qualified I&C technician), two supervisors were present to evaluate performance.
The l&C technician in training demonstrated good knowledge
-8-of the equipment being tested and communicated well with the other technician.
Step 7..1.16 of the procedure instructed the technician to "Slowly lower pressure until contact closure is obtained on DMM"(digital multi-meter).
It was important that the pressure be lowered slowly in order to avoid overshooting the as-found trip setpoint and to obtain an accurate as-found reading of pressure.
While lowering pressure, an overshoot of the setpoint was observed.
This required a second attempt which was acceptable in identifying the setpoint.
The licensee evaluators did not find this portion of the performance within their.exp'ectations.
They considered that the first attempt had preconditioned the instrument, thus removing the opportunity to measure the true as-found value.
The licensee evaluators explained that l8C technicians receive training on how to slowly lower pressure and were expected not to overshoot the setpoint.
The inspectors recalled previous instances where these expectations were not enforced by the licensee.
C.
Conclusions M2 M2.1 Surveillance testing was'generally conducted in accordance with the licensee's programs and Technical Specifications.
The licensee had appropriately stringent expectations for measuring as-found setpoints.
Maintenance and Material Condition of Facility and Equipment Review of Material Condition Durin Plant Tours Ins ection Sco e 62707 The inspectors performed routine plant tours to evaluate housekeeping and material condition at the station.
Observations and Findin s.
In most cases plant housekeeping and material condition were good.
However, the inspectors identified an unsecured portable eye wash station adjacent to the HPCS batteries.
The inspectors reported this condition to an equipment operator in the area who immediately returned the eye wash station to its proper storage location. The licensee later noted that the identified condition did not comply with step 7.2a of
'Procedure 10.2.53.
Step 7.2a required that the height-to-width ratio of the eye wash station not exceed an overturning criteria of 1.7 for an unrestrained object at the 441-foot level of the diesel generator building.
The inspectors considered that failure to comply with the requirements of Procedure 10.2.53 for storage of the eye wash station in the vicinityof safety-related equipment was one example of a noncited violation (NRC Enforcement Policy, Appendix C) of Technical Specification 5.4.1 (50-397/99004-03).
This example of a violation is in the licensee's corrective action program as PER 299-088 Conclusions Plant housekeeping and material condition were generally good; however, the inspectors found an unsecured portable eye wash station too close to the HPCS batteries in violation of procedural requirements:
This is one example of a Severity Level IVviolation of Technical Specification 5.4.1.a, which is being treated as a noncited violation and is in the licensee's corrective action program as PER 299-0889.
Maintenance Procedures and Documentation Review of Scaffoldin Procedure Ins ection Sco e 62707 During a plant tour, the inspectors observed scaffolding less than 2 inches from a safety-related instrument sensing line. The inspectors reviewed the applicable procedure, Procedure 10.2.53, Revision 16, "Seismic Requirements for Scaffolding, Ladders, Man-lifts, Tool Gang Boxes, Hoists, and Metal Storage Cabinets," to determine if the condition was acceptable and if the procedure was adequate.
Observations and Findin s The inspectors observed a section of scaffold bracing less than 2 inches from Reactor Recirculation Pump 13 pressure sensing lines. Procedure 10.2.53 provides guidelines for scaffolding clearances and recommends that all scaffolding items be greater than or equal to 2 inches away from instrument racks and QC-1 equipment.
Step 7.1.6 of the procedure states "Contact Engineering for any of the following conditions... A scaffold is desired which may not fall within these guidelines... a. Engineering to evaluate these specific scaffolds and conditions and provide acceptance/rejection, b. Engineering to provide construction criteria and a [10 CFR] 50.59 review for each request."
The guidelines also recommend a minimum distance of 2 inches between scaffolds and instrument sensing lines; however, the guidelines allow less than 2 inches with special provisions. The guidelines give little information on what constitutes "special provisions" (i.e., relying on the craft using the procedure to determine what constitutes a special provision). The inspectors considered this to be a procedural weakness, because the procedure allowed for acceptance to be determined by the craft erecting the scaffold rather than receiving an engineering evaluation and 10 CFR 50.59 review. At the close of the inspection period, engineering was planning to revise Procedure 10.2.53 to ensure that potential interferences with instrument sensing lines will receive a similar degree of evaluation as other safety important components.
r Conclusions The inspectors identified a procedure weakness that allowed potential interferences between scaffolding and instrument sensing lines to be evaluated by the craft erecting the scaffolding. This was inconsistent with other guidance in the procedure which required engineering evaluation and a 10 CFR 50.59 review for potential interferences between scaffolding and important to safety components.
At the close of the inspection,
-10-engineering was planning to revise the scaffolding procedure to ensure that potential interferences with instrument sensing lines will receive a similar degree of evaluation as other safety-related components.
M8 Miscellaneous Maintenance Issues (92902)
M8.1 Administrative Closure of Violations Based U on Chan es in the Enforcement Polic The inspectors performed an in-office review of outstanding violations in the maintenance area.
The Severity Level IVviolations listed below were issued in Notices of Violation prior to March 11, 1999.
On this date, the NRC implemented a new policy for treatment of Severity Level IVviolations (Appendix C of the Enforcement Policy).
Because these violations would have been treated as noncited violations in accordance with Appendix C, they are being closed out in this report. The inspectors verified that the licensee had generated a corrective action program reference (PER) for each of the violations listed.
In addition, these violations already have docketed responses or were generated with no response required.
Violation Number Description CA Program Reference 50-397/98020-02 failure to include sump cross-connect valves in
- the scope of the Maintenance Rule program 298-1243 298-0780 Review of the effectiveness of the corrective actions for selected violations willbe performed in the future as a routine part of the review of the corrective action program.
III. ENGINEERING E.1 Miscellaneous Engineering Issues
'1.1 Technical S ecification Bases Inaccuracies Associated with RHR Valves 8 and 9 a.
Ins ection Sco e 37550 The inspectors reviewed a recent change to controls associated with Valves RHR-V-8 and RHR-V-9. In addition, the inspectors reviewed associated Technical Specifications and supporting bases.
The inspectors also discussed the recent RHR control changes with the system engineer.
b.
Observations and Findin s While reviewing a recent change to the RHR system, the inspectors identified that Technical Specification Table 3.3;6.1-1 note (d) was no longer applicable..ln addition, the licensee determined that the associated basis section was no longer applicable and contained an error. Specifically, note (d) discussed transferring the outboard valve, RHR-V-8, to the alternate remote shutdown panel.
However, because of a recent design change, control of Valve RHR-V-8 is no longer transferred to the alternate
-11-shutdown room.
In addition, the basis section incorrectly states that there are four pressure switches associated with the reactor high pressure isolation instrumentation, when only two exist.
The licensee initiated PER 299-0584 to address the identified issues.
The PER indicates that there was a missed opportunity to identify the problems during the recent RHR modification and there was an inadequate review of the Improved Technical Specification Bases.
The inspectors identified that the Technical Specification and associated bases for Technical Specification 3.3.6.1 were incorrect. The bases section stated that four pressure switches are associated with the reactor high pressure isolation instrumentation, when only two exist. The licensee stated that the reference to four pressure switches for reactor high pressure isolation was due to an error when converting to the new improved Technical Specifications.
Title 10 CFR Part 50, Appendix B, Criterion III, "Design Control," states, in part, that measures shall be established to assure that the design basis is correctly translated into specifications and instructions.
This issue is an unresolved item, pending verification that the plant was originally licensed with two pressure switches and review of the 10 CFR 50.59 evaluation of the change (50-397/94004-04).
This issue is in the licensee's corrective action program as PER 299-0584.
Conclusion Technical Specification 3.3.6.1, "Primary Containment Isolation Instrumentation,"
-.
Function 5, "Residual Heat Removal Shutdown Cooling System Isolation," and the associated bases section were incorrect. The Technical Specifications were not updated when. the controls for the outboard isolation valve were removed from the alternate remote shutdown panel.
In addition, the bases section incorrectly stated. that there are four pressure switches associated with the reactor high pressure isolation instrumentation, when only two exist. This issue is identified as an unresolved item because additional information is required in order to confirm the facilitywa's originally licensed with only two pressure switches and to review the 10 CFR 50.59 evaluation for the change.
ES Miscellaneous Engineering Issues (92903)
E8.1 Administrative Closure of Violations Based U on Chan es in the Enforcement Polic The inspectors performed an in-office review of violations that are outstanding in the engineering area.
The Severity Level IVviolations listed below were issued in Notices of Violation prior to March 11, 1999.
On this date, the, NRC implemented a new policy for treatment of Severity Level IVviolations (Appendix C of the Enforcement Policy).
Because these violations would have been treated as noncited violations in accordance with Appendix C, they are being closed out in this report. The inspectors verified that the licensee had generated a corrective action program reference (PER) for each of the violations listed.
In addition, these violations already have docketed responses or were generated with no response require Violation Number 50-397/98020-01 EA 98-480; 01013 50-397/98019-02 50-397/98015-02 50-397/98015-01 Description design inadequacies in the fire protection water supply system inappropriate approval of a test procedure that involved a change in Technical Specifications procedures not followed or established for marking Categories 1 and 2 postaccident monitoring instruments and verifying short circuit calculation assumptions design basis for condensate storage tank capacity and reactor vessel water level not correctly translated into Technical Specifications CA Program Reference 298-1243 298-0780 298-1024 298-0967 298-0898 298-0963 298-0897 Review of the effectiveness of the corrective actions for selected violations willbe performed in the future as a routine part of the review of the corrective action program.
IV. PLANTSUPPORT R1 Radiological Protection and Chemistry Controls R1.1 General Comments a.
Ins ection Sco e 71750 The inspectors routinely toured the radiologically controlled areas and observed health physics personnel and radiation workers in the field.
b.
Observations and Findin s The inspectors walked down the plant and observed overall postings, radiological controls, and work practices.
Specifically, the inspectors observed radiological control associated with the unloading and movement of fresh fuel. In addition, the inspectors reviewed Procedure 11.2.7.1, "Area Posting," Revision 12.
The inspectors observed personnel open and remove fresh fuel from shipping packages and move it to a storage areas within the plant. The inspect'ors noted that radiological controls were generally good and that health physics oversight helped personnel maintain exposure ALARA.
The inspectors noted that, on the 606-foot elevation of the reactor building, a contaminated area boundary was not marked or posted.
Specifically, a platform that overhangs the spent fuel pool had no rope or other markings to define it or the spent fuel pool as a contaminated area.
In addition, the area status board showed it as being
-13-posted.
Procedure 11.2.7.1, Section 4.6, "Contaminated Area," stated, "Barrier the perimeter of the Contaminated Area at approximately waist height...Yellow and magenta "Contaminated Area" tape may be used on the floor... At the accessible perimeter to the Contaminated Area, post signs bearing the radiation symbol and the words CAUTIONwith an insert of CONTAMINATEDAREA." The licensee initiated PER 299-0718 to address the issue.
The inspectors identified a violation of Technical Specification 5.4.1.a in that Procedure 11.2.7.1 was not complied with, since a contaminated area was not posted or marked as required.
This Severity Level IV violation is being treated as the second example of a noncited violation, consistent with Appendix C of the NRC Enforcement Policy (50-397/99004-03).
This violation is in the licensee's corrective action program as PER 299-0718.
C.
Conclusion R1.2 Radiological controls associated with the unloading of fresh fuel were generally good and health physics oversight helped personnel maintain exposure ALARA. However, the licensee failed to post or mark a contaminated area as required by procedure.
This is one example of a Severity Level IVviolation of Technical Specification 5.4.1.a and is being treated as a noncited violation. This deficiency is in the licensee's corrective action program as PER 299-071 8.
Calibration of Radiation Monitorin E ui ment a0 Ins ection Sco e 71750 b.
The inspectors performed an audit of the radiation monitoring equipment at WNP-2 to verify that each of the radiation monitors had been successfully calibrated within their required calibration frequency.
'bservations and Findin s Calibrations of all of the radiation monitors at WNP-2 had been successfully performed within their required calibration frequency.
'
Conclusion Calibration of radiation monitoring equipment at WNP-2 was being performed within the required calibration frequencies.
RS Miscellaneous Radiation Protection and Chemistry Issues (92904)
R8.1 Administrative Closure of Violations Based U on Chan es in the Enforcement Polic
The inspectors performed an in-office review of outstanding violations for Radiation Protection.
The Severity Level IVviolations listed below were issued in Notices of Violation prior to March 11, 1999.
On this date, the NRC implemented a new policy for treatment of Severity Level IV violations (Appendix C of the Enforcement Policy).
-14-Because these violations would have been treated as noncited violations in accordance with Appendix C, they are being closed out in this report. The inspectors verified that the licensee had generated a corrective action program reference (PER) for each of the violations listed.
In addition, these violations already have docketed responses or were generated with no response required.
Violation Number Description CA Program Reference 50-397/98006-04 failure to implement requirements of Procedure PPM 11.2.7.3, "High, High High, and Very High Radiation Area Controls" 50-397/98010-01 failure to followALARArequirements 298-0810 298-0332 50-397/98004-03 lack of experience on radioactive effluent audit 298-0258 team 298-0251 Review of the effectiveness of the corrective actions for selected violations will be performed in the future as a routine part of the review of the corrective action program.
PS Miscellaneous Emergency Planning Issues (92904)
P8.1 Administrative Closure of Violations Based U on Chan es in the Enforcement Polic The inspectors performed an in-office review of outstanding violations for Emergency Preparedness.
The Severity Level IVviolations listed below were issued in Notices of Violation prior.to March 11, 1999. On this date, the NRC implemented a new policy for treatment of Severity Level IVviolations (Appendix C of the Enforcement Policy).
Because these violations would have been treated as noncited violations in accordance with Appendix C, they are being closed out in this report. The inspectors verified that the licensee had generated a corrective action program reference (PER) for each of the violations listed.
In addition, these violations already have docketed responses or were generated with no response required.
Violation Number 50-397/98014-01 Description reduction of training requirements/lack of training program rbquirements CA Program Reference'
298-0921 50-397/98009-07 failure to get NRC approval for an emergency 298-0377 plan change Review of the effectiveness of the corrective actions for selected violations willbe performed in the future as a routine part of the review of the corrective action progra V. MANAGEMENTMEETINGS X1 Exit Meeting Summary The inspectors presented the inspection results to members of licensee management on May 6, 1999. The licensee acknowledged the findings presented.
The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. A vendor evaluation of fuel storage criticalitywas identified as proprietary. Although the evaluation was examined by the inspectors, it was not included or described in this repor Licensee ATTACHMENT SUPPLEMENTAL INFORMATION PARTIALLIST OF PERSONS CONTACTED D. W. Hillyer, Radiation Protection Manager P. J. Inserra, Licensing Manager J. A. McDonald, Production Manager W. S. Oxenford, Operations Manager J. V. Parrish, Chief Executive Officer D. C. Perry, Radiation Operations Supervisor D. J. Poirier, Maintenance Manager G. O. Smith, Vice President - Generation/Nuclear Plant General Manager R. L. Webring, Vice President - Operations Support IP 37551:
IP 61726:
IP 62707:
IP 71707:
IP 71750:
IP 92901:
IP 92902'P 92903:
IP 92904:
INSPECTION PROCEDURES USED Onsite Engineering Surveillance Observations Maintenance Observations Plant Operations Plant Support Plant Operations Followup Maintenance Followup Onsite Engineering Followup Plant Support Followup ITEMS OPENED AND CLOSED
~Oened 50-397/99004-01 URI design basis of the RHR system (Section 02.1)
50-397/99004-02 50-397/99004-03 50-397/99004-04 NCV failure to complete corrective actions associated with color banding of instrumentation (Section 07.1)
NCV two examples of a violation of Technical Specification 5.4.1:
(1) unsecured eyewash station (Section M2.1); and (2) failure to post contaminated area (Section R1.1)
URI violation of 10 CFR 50.59; Technical Specification Table 3.3.6.1-1 note (d) no longer applicable
Closed 50-397/99004-02 50-397/99004-03 50-397/98013-01 50-397/98009-05 50-397/98009-04 50-397/98009-02 50-397/98009-01 50-397/98006-03 50-397/98006-02 50-397/98006-01 50-397/98005-04 50-397/98005-03 50-397/98005-02 50-397/98005-01 50-397/98003-05 50-397/98003-04 NCV NCV VIO VIO VIO VIO VIO VIO VIO VIO VIO VIO VIO VIO VIO VIO VIO-2-failure to complete corrective actions associated with color banding of instrumentation (Section 07.1)
two examples of a violation of Technical Specification 5.4.1:
(1) unsecured eyewash station (Section M2.1); and (2) failure to post contaminated area (Section R1.1)
improper lineup of the low pressure core spray system (Section 08.1)
failure to followprocedure for testing excess flowcheck valves (Section 08.1)
failure to followprocedure for testing excess flowcheck valves (Section 08.1)
'ailure to properly evaluate a partially elevated fuel assembly failure to verify prerequisites for moving heavy loads (Section 08.1)
failure to establish adequate controls for residual heat removal suppression pool return (Section 08.1)
10 CFR 50.59 and 50.9 violation for maintaining residual heat removal minimum flowvalve closed (Section 08.1)
failure to identify and track and inoperable train of control room heating, ventilation, and air conditioning (Section 08.1)
two examples of inadequate procedures (Section 08.1)
failure to properly implement procedure change (Section 08.1)
failure to meet surveillance requirements for a surveillance procedure (Section 08.1)
failure to provide 10 CFR 50.72 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> notification (Section 08.1)
inadequate corrective action for control of transient combustibles (Section 08.1)
inadequate procedure to address requirements of Technical Specification 5.5.2 (Section 08.1)
improper application of Technical Specification 3.0.2 (Section 08.1)
-3-50-397/98003-02 VIO improper application of Surveillance Requirement 4.0.2 (Section 08.1)
50-397/97018-07 VIO failure to control transient combustibles (Section 08.1)
50-397/97018-01 VIO failure to properly secure transient or portable equipment (Section 08.1)
50-397/98020-02
. VIO failure to include sump cross-connect valves in the scope of the Maintenance Rule program (Section M8.1)
50-397/98020-01 EEI design inadequacies in the fire protection water supply system VIO -
(Section 8.1)
50-397/98019-02 VIO inappropriate approval of a test procedure that involved a change in Technical Specifications (Section E8.1)
50-397/98015-02 VIO procedures not followed or established for marking Category
and two postaccident monitoring instruments and verifying short circuit calculation assumptions (Section E8.1)
50-397/9801 5-01 VIO design basis for condensate storage tank capacity and reactor vessel water level not correctly translated into Technical Specifications (Section E8.1)
50-397/98010-01 50-397/98006-04 50-397/98004-03 50-397/98014-01 50-397/98009-07 VIO failure to followALARArequirements (Section R8.1)
VIO failure to implement requirements of Procedure PPM 11.2.7.3,
"High, High High, and Very High Radiation Area Controls" (Section R8.1)
VIO lack of experience on radioactive effluent audit team (Section R8.1)
VIO reduction of training requirements/lack of training program requirements (Section P8.1)
VIO failure to get NRC approval for an emergency plan change (Section P8.1)
-4-LIST OF ACRONYMS USED ALARA CFR HPCS I&C NCV NRC OI PDR PER RHR URI VIO WNP-2 as low as reasonably achievable Code of Federal Regulations high pressure core spray instrument and controls noncited violation U.S. Nuclear Regulatory Commission Operating Instruction public document room problem evaluation request residual heat removal unresolved item violation Washington Nuclear Project-2
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