IR 05000271/1998013
| ML20198L754 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 12/23/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20198L744 | List: |
| References | |
| 50-271-98-13, NUDOCS 9901050110 | |
| Download: ML20198L754 (43) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Docket No.
50-271 Licensee No.
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Repon No.
98-13
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Licensee:
Vermont Yankee Nuclear Power Corporation Facility:
Vermont Yankee Nuclear Power Station Location:
Vernon, Vermont
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Dates:
October 11 - November 21,1998
Inspectors:
Brian J. McDermott, Senior Resident inspector Edward C. Knutson, Resident inspector Jason C. Jang, Radiation Specialist George W. Morris, Reactor Engineer Douglas A. Dempsey, Reactor Engineer l
Thomas G. Scarbrough, NRR Approved by:
Curtis J. Cowgill, Ill, Chief, Projects Branch 5 Division of Reactor Projects
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9901050110 981223 PDR ADOCK 05000271
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EXECUTIVE SUMMARY NRC Inspection Report 50-271/98-13 i
This inspection included aspects of operations, engineering, maintenance, and plant l
support at the Vermont Yankee (VY) Nuclear Power Station. The report covers a six week period of routine resident inspector activities and inspections by specialists in effluent controls, engineering, and motor-operated valves.
Ooerations Human performance issues and weakness in attention to detail were identified
through NRC inspections and self-revealing events. However, in one instance the questioning attitude of an equipment operator prevented VY from missing a Technical Specification required surveillance. These examples of human performance were discussed with VY management and are being evaluated as part of broader efforts to improve performance. No safety related equipment was effected and no violations of NRC requirements were identified. (Section 01.1)
VY identified a degrading steam leak on a main steam line (MSL) drain pot gasket.
- Plant management was involved in assessing the situation and establishing an appropriate course of action. Operators performed well during the reactor shutdown and startup associated with the forced maintenance outage and there were no significant operational challenges. (Section 01.2)
Maintenance The maintenance activities observed during this period were performed well.
- Workers in the drywell demonstrated appropriate radiological control techniques and were proficient at using test equipment and performing component reassembly under difficult conditions. Close supervisory involvement and Quality Assurance support were observed during the service water pump replacement. (Section M1.1)
The surveillance activities observed during the period were correctly performed.
- Test activities were well controlled and coordinated by the control room operators.
(Section M1.2)
Inadequate post-maintenance testing of the main turbine control system during the
1998 refueling outage led to an operational problem after startup and a forced maintenance outage. Troubleshooting activities were methodical and well planned.
VY management is integrating the lessons learned from this event with the corrective actions being developed in response to previously identified maintenance oversight and control weaknesses. (Section M1.3)
The "C" Residual Heat Removal Service Water (RHRSW) pump failed a quarterly
surveillance test and surveillance testing of the redundant pumps revealed acceptable, although marginal, performance. Corrective maintenance on the "C" RHRSW pump was successful based on the post maintenance test results. VY l
demonstrated a good safety perspective by initiating plans to perform near term l
preventive maintenance on the remaining pumps. (Section M2.2)
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Enaineerina A number of open items related to the NRC Architect / Engineer Team inspection
50-271/97-201 were reviewed during this report period. Ten open items were resolved through either final NRC disposition of an unresolved issue or VY's corrective actions for items previously cited as violations. (Section E8.1)
Plant Support The licensee maintained effective radioactive liquid and gaseous effluent control
programs. The Offsite Dose Calculation Manual contained sufficient specification and instruction to acceptably implement and maintain the radioactive liquid and gaseous effluent control programs. (Section R1.1)
The licensee established, implemented, and maintained an effective radiation
monitoring system calibration program, including flow rate measurement systems.
As a result of self-assessment initiatives, the licensee implemented efforts to improve radiation monitoring system reliability. The licensee also established and implemented an effective hydrogen / oxygen monitor catibration program.
(Section R2.1)
The licensee established, implemented, and maintained an effective standby gas
treatment system surveillance program with respect to charcoal adsorption surveillance tests, HEPA mechanical efficiency tests, and air flow rate tests.
(Section R2.2)
The licensee established, implemented, and maintained an effective quality
assurance program for the radioactive effluent control program with respect to audit scope and depth, audit team experience, and response to audit findings. The licensee also implemented an effective quality control program to validate measurement results for radioactive effluent samples. (Section R7.1)
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TABLE OF CONTENTS i
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L EX EC UTIV E S U M M ARY.............................................. il i
TA BLE O F C O NTENTS.............................................. av i
Summary of Plant Status
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l. Operations
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- Conduct of Operations.................................... 1 01.1 Observation of Routine Plant Operations................... 1
.01.2 Shutdown Due to Balance-of-Plant Equipment Problems........ 2
Operator Knowledge and Performance......................... 3 04.1 Auxiliary Operator Rounds
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ll. M aint ena nce................................................... 4 M1 Conduct of Maintenance................................... 4
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M1.1 Maintenance Observations............................ 4 j
M1.2 Surveillance Observations............................. 4
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M1.3 Main Turbine Hydraulic Control System Maintenance.......... 5
M2 Maintenance and Material Condition of Facilities and Equipment....... 6 i'
M2.1 Steam Leak on Main Turbine Instrument Tap................ 6 M2.2 Residual Heat Removal Service Water (RHRSW) Pump Failure.... 7 M8 Miscellaneous Maintenance issues............................ 8 M8.1 Review of Open ite ms............................... 8 M8.2 in-office Review of LERs Related to Maintenance and Surveillance.9 111. E ngi ne e ring.................................................. 1 1-E8
. Miscellaneous Engineering issues............................ 11 E8.1 Review of Open items
..............................11 E8.2 in-office Review of LERs Related to Engineering
............23 E8.3 (Closed) LER 9 8-01 8 00............................. 2 4
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E8.4 (Closed) URI 98-08-06: Impact of Low Blowout Panel Set Point on i
L Secondary Containment............................. 2 5 IV. Plant Support
................................................26 R1 Radiological Protection and Chemistry (RP&C) Controls............ 26 R1.1 Radioactive Liquid and Gaseous Effluent Control Programs..... 26 L
R2 Status of RP&C Facilities and Equipment
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R2.1 Calibration of Radiation and Hydrogen / Oxygen Monitoring Systems
...............................................27 R2.2 Air Cleaning Systems............................... 28 R7 Quality Assurance (QA) in RP&C Activities..................... 29 l:
R7.1 Radioactive Effluent Control and Effluent Sample Validation.... 29
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R8 Miscellaneous RP&C issues................................ 29 R8.1 Review of Open items..............................29
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F8 Miscellaneous Fire Protections issues......................... 30 F8.1 Review of Open items
..............................30 V. Management Meetings
..........................................31 X1 Exit Meeting Summary................................... 31 ITEMS OPENED, CLOSED, AND DISCUSSED.............................. A-1 LIST O F ACRO NYMS U S ED.......................................... B-3 ENGINEERING INSPECTION OPEN ITEM CROSS REFERENCE.................. C-1 v
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e Report Details Summarv of Plant Status At the beginning of the inspection period, Vermont Yankee (VY) was operating at 95 percent power due to problems observed with the main turbine control system at close to 100% power. A short duration outage was planned to start October 19, for repair of the control system, however, VY commenced the shutdown three days early due to a main steam line condensate drain pot leak. Although a smallleak from drain pot's gasket was j
previously identified and scheduled for repair during the outage, the gasket degraded
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significantly during the week leading up to the outage. After an attempt to isolate the leak on October 15 was unsuccessful, VY began the shut down ahead of schedule. The main turbine was taken off-line at 4:34 a.m. on October 16, and cold shutdown conditions were i
established by late that afternoon. Planned outage activities were completed on October 20, and the reactor was taken critical at 3:43 p.m. The main generator was synchronized to the grid the following day at 11:30 p.m., and full power operation was achieved on October 23. With the exception of planned power reductions for surveillance testing and rod pattern adjustments, the plant operated at 100 percent power for the remainder of the inspection period.
1. Operations
Conduct of Operations'
01.1 Observation of Routine Plant Operations (71707)
a.
Inspection Scope (71707)
Routine tours of the control room were made to assess the conduct of activities, verify safety system alignments, and determine compliance with Technical Specification (TS) requirements. Event Reports (ERs) used to document plant deficiencies were reviewed, and discussed with shift supervision, to evaluate both the equipment condition discussed and the licensee's initial response to the issue.
b.
Observations and Findinas Four events dunng this report period indicate an attention to detail concern during routine plant activities. The individualissues are discussed below and the inspector noted that VY had initiated appropriate corrective actions.
On October 23, the NRC identified that control rod drive pump "B" flow was
less than the minimum value specified on a caution tag on the main control board. Once notified, operators promptly corrected the condition. The operating limit specified on the caution tag provided margin above the necessary flow value, and therefore there was no immediate concern.
' Topical headings such as 01, M8, etc., are used in accordance with the NRC standardized reactor inspection report outline. Individual reports are not expected to address all outline topic _ _ _ _ _
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2 On October 27, a reactor building fan damper was not properly restored
following maintenance and its air supply remained isolated. A subsequent attempt to start the associated reactor building exhaust fan failed because the damper would not operate and the reactor building became pressurized for a short time. The licensee's preliminary cause of this event was human error on the part of the maintenance personnel.
On November 2, the NRC identified that the meter indication for the "C"
source range monitor (SRM) had decreased in comparison to previous control room tours. After determining a degraded condition existed, VY declared the
"C" SRM inoperable and placed it in bypass. However, the TS do not require the SRMs to be operable when the reactor mode switch is in the "RUN" position.
On November 17, during the tag-out of the "B" station air compressor for
maintenance, an AO opened the electrical circuit breaker for the wrong compressor. The air compressors are not safety related equipment and this event had no operational effect due to additional air compressors in service.
The error was promptly corrected after consultation with the control room.
The inspector also noted that an example of excellent AO performance occurred during this period.
On October 28, an AO observed that a TS-related instrument reading he had
previously been required to log during rounds was no longer on a new revision of the AO rounds log. The AO contacted the control room and the operators subsequently determined the reading was inadvertently omitted during administrative preparation of the document. Because this was the first use of the revised log sheet, no required readings were missed and the AO's questioning attitude prevented a TS violation.
c.
Conclusions Human performance issues and weakness in attention to detail were identified through NRC inspections and self-revealing events. However, in one instance the questioning attitude of an equipment operator prevented VY from missing a Technical Specification required surveillance. These examples of human performance were discussed with VY management and are being evaluated as part of broader efforts to improve performance. No safety related equipment was effected and no violations of NRC requirements were identified.
01.2 Shutdown Due to Balance-of-Plant Eauioment Problems a.
Insoection Scoce (71707)
A maintenance outage was planned to start October 19, to support the investigation and repair of a problem with the main turbine control system.
However, the shutdown was commenced three days early due to a s;asket leak on a
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steam line condensate drain pot located in the main steam tunnel. The inspector observed the licensee's response to the increar6g steam leak, and portions of the plant shutdown, cool down, and startup.
b.
Observations and Findinas The licensee appropriately evaluated the personnel safety hazards associated with the attempts to isolate the drain pot. The inspector noted that VY approached this activity as an option and that the plant manger was clear in communicating that the safety of the workers was more important than isolating the drain pot. VY sought to isolate the leaking drain pot as an option that would have provided more time for preparation and transition into the maintenance outage. When the licensee determined that the drain pot leak had increased beyond the workers level of
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comfort, VY management took appropriate action and directed operators to initiate the shutdown. The inspector noted that the temperature and radiation level in the main steam tunnel were both within the licensee's administrative limits, and significantly below TS setpoints for the MSIV isolation instrumentation, when the decision was made to shutdown. The shutdown and cooldown were well controlled.
On October 20, the inspector observed the reactor startup from the commencement of rod withdraw through criticality. Although there were a number of unrelated
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activities in the control room during the startup, the inspector observed that a senior reactor operator supervised all control rod manipulations and that operators remained focused on the reactor startup. The approach to criticality was controlled and deliberate, and no significant problems were encountered, c.
Conclusions VY identified a degrading steam leak on a main steam line (MSL) drain pot gasket.
Plant management was involved in assessing the situation and establishing an appropriate course of action. Operators performed well during the reactor shutdown and startup associated with the forced maintenance outage and there were no significant operational challenges.
O4 Operator Knowledge and Performance 04.1 Auxiliarv Operator Rounds (71707)
On November 11 and November 20, the inspector accompanied an AO during a routine set of rounds in the reactor building. The operators were knowledgeable of their responsibilities and the expected status of plant equipment. Requests from the control room to manipulate plant equipment were adequately performed. No problems were identified during these tours of the reactor buildin.
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11. Maintenance l
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M1 Conduct of Maintenance l
M 1.1 Maintenance Observations a.
Inspection Scooe (62707)
i The inspector observed portions of plant maintenance activities to verify that the correct parts and tools were utilized, the applicable industry code and technical specification requirements were satisfied, adequate measures were in place to ensure personnel safety and prevent damage to plant structures, systems, and components, and to ensure that equipment operability was verified upon completion of post maintenance testing, b.
Observations and Findinas The inspector observed all or portions of the following maintenance activities:
Source Range Monitor "B" detector troubleshooting, October 19
The instrument developed abnormalindication during the forced outage. The
problem was found to be a ground on the cable shield in the drywell.
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"D" Service Water pump and motor replacement, October 28
Observed torquing of the pump casing to discharge piping flange, motor installation, and shaft packing installation. An injection-type shaft packing material was installed as a temporary modification; shaft packing leakage has been a long-term
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housekeeping issue with the service water pumps, c.
Conclusions
l The maintenance activities observed during this period were perfornied well.
Workers in the drywell demonstrated appropriate radiological control techniques and were proficient at using test equipment and performing component reassembly under difficult conditions. Close supervisory involvement and Quality Assurance support were observed during the service water pump replacement.
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M1.2 Surveillance Observations
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Insoection Scoce (61726)
The inspector observed portions of surveillance tests to verify proper calibration of
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test instrumentation, use of approved procedures, performance of work by qualified
personnel, conformance to Limiting Conditions for Operations (LCOs), and correct post-test system restoration.
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b.
Observations and Findinas The inspectors observed portions of the following surveillance testing activities:
Reactor Core Isolation Cooling system pump surveillance, performed on
October 15. No problems were noted.
Turbine Hydraulic Control system post-maintenance testing, performed on
October 23.
The inspector observed bypass valve testing at 98 percent power. System response was very smooth, with only a slight perturbation in steam flow evident when the bypass valve opened, and no sustained oscillations. Turbine hydraulic control system maintenance is further discussed in section M1.3 of this report.
Service Water pump "D" post-maintenance testing, observed October 29
No problems were noted; the inspector observed that the new shaft packing was very effective at controlling leakage during the post-maintenance test.
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Conclusions The surveillance activities observed during the period were correctly performed.
Test activities were well controlled and coordinated by the control room operators.
M1.3 Main Turbine Hydraulic Control System Maintenance a.
Inspection Scoce (62707)
The inspector reviewed VY's troubleshooting and maintenance activities associated with the main turbine hydraulic control system instability that was observed at greater than 95 percent power operation.
b.
Observations and Findinas
' Troubleshooting of the main turbine hydraulic control system had been in progress l
since shortly after the 1998 refueling outage, as discussed in NRC Inspection Reports 50-271/98-12and 9811. Although several equipment problems had been-identified and corrected, the instability persisted. During this inspection period, the
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possibilities for on-line troubleshcoting were exhausted, and plans were developed for a plant shutdown to support continued troubleshooting.
During the October maintenance outage, numerous problems were identified with the main turbine hydraulic control system. Problems included loose hydraulic oil line connections that had been leaking, improper control valve adjustments, and
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mechanicallinkage problems (looseness / binding). VY determined that this combination of individual problems was the likely cause of the system instability.
During the subsequent startup and post-maintenance testing, the system performed
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very well. No instability in the control system was noted by VY during any testing through 100 percent power operation and the inspector observed that the monitored turbiae parameters showed significant improvement.
Based on the October maintenance outage findings, VY management recognized i
that work on the turbine control system during the refueling outage was not l
adequately controlled and that the post-maintenance setup had apparently not been l
completed. In light of this, and other earlier instances of refueling outage problems, l
VY is examining their management and control of outage maintenance. The inspector noted that the main turbine hydraulic control system is not a safety system and that the problems identified through the course of this troubleshooting activity would not have impacted interfaces with the reactor protection system. No violations were identified.
Inadequate post-maintenance testing of the main turbine control system during the 1998 refueling outage led to an operational problem and a forced maintenance I
outage. This issue is considered an additional example of problems with the oversight and control of maintenance during the 1998 refueling outage identified in the NRC Systematic Assessment of Licensee Performance (SALP) Report, dated August 28,1998.
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Conclusions
Inadequate post-maintenance testing of the main turbine control system during the 1998 refueling outage led to an operational problem after startup and a forced maintenance outage. Troubleshooting activities were methodical and well planned.
VY management is integrating the lessons learned from this event with the corrective actions being developed in response to previously identified maintenance oversight and control weaknesses.
M2 Maintenance and Material Condition of Facilities and Equipment M2.1 Steam Leak on Main Turbine instrument Tao (62707)
On October 23, while performing routine radiation surveys, a Radiation Protection (RP) technician noted steam wisping from the area of the high pressure (HP) turbine steam inlet. After insulation removal, maintenance personnelidentified the source to be a crack in a small bore pipe connected to the HP turbine inlet manifold. The spare instrument / test line is capped and appears to have been slightly bent to one side during a previous work activity. Because no work had been done in this area during the October 1998 mini-outage, VY concluded the condition existed since the 1998 refueling outage, at a minimum. Based on the size of the crack, VY determined immediate actions were not necessary. Remote video monitoring of the l
steam leak was established while plans were initiated to perform a temporary repair.
l There was no apparent change in the magnitude of the steam leak throughout the l
remainder of the inspection period. The inspectors discussed this issue with representatives of the NRC Region I, Civil, Mechanical and Materials Engineering Branch, and concluded that VY was taking an acceptable course of action.
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M2.2 Residual Heat Removal Service Water (RHRSW) Pumo Failure
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Inspection Scoos (61726,62707)
l The inspector reviewed pump performance data from quarterly RHRSW pump
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surveillances since 1994, the licensee's justification for establishing new pump l
reference values in May 1998, and the licensee's initial corrective actions in
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response to a test failure of the "C" RHRSW pump.
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Observations and Findinas
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- On November 3,1998, the "C" RHRSW did not meet the minimum pump
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differential pressure requirement of CP 4124, " Residual Heat Removal and RHR Service Water System Surveillance." The pump produced a differential pressure of 138 psid which was below the minimum required value of 144 psid. As corrective i
action the pump's bowls and rotating elements were replaced. Also, a 3/4" layer of microbiological!y induced tubercles were cleaned off of the inside surface of the pump's suction barrel. During post maintenance testing, the pump produced a differential pressure of 165 psid which represented a 15% margin above the minimum acceptance criteria.
VY revised the RHRSW pump surveillance acceptance criteria in May 1998 to account for instrument uncertainties. With an increase in test flow from 2700 gpm
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to 2810 gpm, the pumps became marginalin their ability to develop the minimum required differential pressure. This marginal performance was also evident because l
the minimum allowable differential pressure for ASME Section XI requirements (+/-
L 10% of the reference value) would be below the pump's the minimum design basis requirement. As a result, VY established the surveillance test acceptance criteria at the design basis minimum differential pressure.
At the close the this inspection period, VY had developed a corrective action plan to l
address each of the three remaining RHRSW pumps prior to their next scheduled
surveillance.
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Conclusions j_
- The "C" Residual Heat Removal Service Water (RHRSW) pump failed a quarterly surveillance test and surveillance testing of the redundant pumps revealed acceptable, although marginal, performance. Corrective maintenance on the "C" RHRSW pump was successful based on the post maintenance test results. VY demonstrated a good safety perspective by initiating plans to perform near term
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preventive maintenance on the remaining pumps.
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M8-Miscellaneous Maintenance issues M8.1 Review of Open items (92902)
The following open items were reviewed for closure based on a sampling of the l
licensee's corrective actions.
l (Closed) VIO 95-22-01: Inservice Test (IST) Program Scope This violation involved failure to test periodically certain power-operated and relief valves in the HPCI, RCIC, and RHR systems in accordance with 10 CFR 50.55a
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(Inservice Testing). In its November 16,1995 response to the Notice of Violation, l
VY listed several corrective actions including: (1) addition of the valves to the IST program, (2) performance of additional tests of valves ofready in the program, and clarification of certain test requirements and methods. The licensee also performed l
a detailed program scope review and documented the results in computerized component basis data sheets. Scope and test deficiencies identified by the licensee during its review were documented in LERs95-017,95-018, and 96-001. NRC i
l review of the LERs was documented in inspection Reports 50-271/95-25 and 96-l 03. The inspector reviewed procedure PP-7013, " Inservice Testing Program Plan,"
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Revision 19, dated October 1998, and verified that the corrective actions were completed. VY's comprehensive corrective actions adequately addressed this violation.
i (Closed) VIO 95-22-02: Valve Inservice Test Deficiencies VY failed to specify limiting values of full stroke time for certain power-operated valves, and did not adequately verify the remote position indicators of sealed solenoid-operated valves (SOVs). The licensee established limiting stroke time acceptance criteria for some valves based on accident analysis assumptions. Where an analytical limit was not ' assumed, VY established the upper reference value limit as the limiting value of full stroke time. The latter practice is more conservative than the ASME Code (1989 Edition) requires and it, acceptable. The inspector also reviewed surveillance procedures and verified that SOV position indicators were being verified by positive means as required by Part 10 of ASME/ ANSI OMa-1988.
(Closed) VIO 95-22-03: Inadequate Testing of Minimum Flow Check Valves During quarterly surveillance testing of the HPCI and RCIC pumm VY incorrectly credited flow through the minimum flow line check valves as futalow exercise tests. However, because the required accident flow rate was not verified, credit could only be taken for a partial flow exercise. The licensee's subsequent attempts to measure the flow rates through the minimum flow lines with ultrasonic l-instruments'were unsuccessful. VY currently treats the quarterly tests as partial L
stroke exercises and disassembles and inspects the check valves each refueling outage. This approach is consistent with the Positions 1 and 2 of Generic Letter 89-04," Guidance on Developing Acceptable Inservice Test Programs," and the c
l ASME Code. The inspector verified through review of completed work orders and l
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other documents that the inspections were performed during the 1996 refueling outage, that future inspections were scheduled properly, and that the IST program documents were updated appropriately.
(Closed) VIO 95-23-01: Inadequate inservice Testing of Stop Check Valves VY had been verifying the ability of stop check valves V13-817 and V23-842 to check flow in the reverse direction by closing the valves with the manual handwheels. This method did not test the check valve function adequately and was inconsistent with the guidance contained in Appendix A (Question Group 25) of NUREG 1482, " Guidelines for Inservice Testing at Nuclear Power Plants." VY initially revised its program documents to disassemble and inspect the valves each refueling outage. Subsequently, the valves we e replaced with nozzle-type check valves.
VY attributed the violation partly to inadequate review of industry operating expenence and performed a review of its operating experience program. In IR 50-271/96-200,the NRC documented that VY was implementing an ecceptable industry operating experience program. The licensee also concluded that the IST program coordinator had been unable to perform quality reviews of industry information due to being overburdened with collateral duties. Procedure PP-7013,
" Inservice Testing Program," establishes the IST program coordinator as the
" central point of focus" and program owner. Through discussions with the current program coordinator, the inspector confirmeo that he is assigned no collateral duties. The licensee's corrective actions were acceptable.
M8.2 In-office Review of LERs Related to Maintenance and Surveillance (90712)
An in-office review of the following LERs was performed to assess whether further NRC actions were required. The adequacy of the overall event description, immediate actions taken, cause determination, and corrective actions were considered during this review. The following issues were closed-out based on the in-office review.
(Closed) LER 97-013-00: Unknown Cause(s) Result in Extended Actuation Times for Three-Out-of-Four Safety Relief Valves Three of the four safety relief valves (SRVs) were removed from the plant during the 1996 refueling outage were found to have actuation times greater than listed in the FSAR during testing performed in June 1997. One valve exhibited delays in the time from pilot actuation to main disk motion and in the time for the main disk to fully stroke. Two valves exhibited delays in the time from pilot actuation to main disk motion. VY determined the as-found SRV actuation times would not have caused any design criteria to be exceeded in the event of a postulated transient or accident. An investigation by the contracted test facility and the manufacturer (Target Rock) included valve disassembly, inspection, and test data evaluation. The licensee was unable to determine the cause of the increased actuation times and a VY review of industry experience concluded that this was an isolated occurrenc.
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No weakness in the work practices used during the handling or te? ting of the valves were identified by the licensee and therefore, no corrective actions were planned.
The inspector found that the LER adequately described the event and the licensee's basis for concluding the findings did not pose an increased risk to public health and safety. The inspector noted that the investigation described covered the most common causes expected for this type of test results. The inspector noted that the four SRVs removed during the 1998 refueling outage were tested in November and exhibited response times within the expected tolerance.
(Closed) LER 98-009-00: Main Steam Isolation Valve Leakage Exceeds Technical Specification Limit Which Could Have Impacted the Ability of a System to Mitigate Consequences of an Accident During the 1998 refueling outage, test results identified that the local leakage rates for the inboard and outboard isolation valves on the "B" and "C" steam lines were in greater than Technical Specification (TS) limits for the individual valves.
However, VY determined that the overall containment leakage was within TS limits.
All MSIVs that exceeded the TS leakage rates were rebuilt prior to startup from the refueling outage.
VY had not completed their root cause investigation within the 30 day time frame for reporting this event. The LER states that a supplement will be submitted once the root cause analysis is complete and the expected submission date was listed as i
July 31,1998. As of November 20,1998, VY had not submitted the supplement.
This issue was discussed with plant management. No significant safety issues were identified during the inspector's review of the original LER and the technical issue will be further evaluated during review of the supplemental report. LER 98-
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009-00is administratively closed.
(Closed) LER 98-011-00.01: High Pressure injection end Reactor Core Isolation Cooling Systems low Steam Supply Pressure Isolation Function Bypassed During Start-up Contrary To Technical Specification Requirements in April 1998, the licensee iaentified that the operation of a bypass switch for low steam pressure isolation signals to HPCI and RCIC during plant startup was in violation of TS requirements. A supplemental report was necessary because VY had not determined the cause of the event within the 30 day time frame for reporting this event.
VY's investigation found that the bypass switches for the low pressure isolation signals were installed, and written into procedures, in early 1972 after a field design change was approved by General Electric. The change was implemented to allow
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heatup of the HPCl and RCIC steam supply lines when reactor pressure is below l
150 psig. VY's investigation found that the original FSAR specifically identified l
these isolation signals were not primary containment ist.!ation system functions.
l However, use of the bypass switches was inconsistent with TS Table 7.3.1 and the
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design / procedure change processes at that time did not identify the conflict. VY l
appropriately reported this condition prohibited by the TS.
VY concluded that the low pressure isolation circuit design and original FSAR designations indicate the original design was not intended for the immediate
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isolation of primary containment in response to a line break. VY personnel are evaluating whether the low pressure isolation was intended to isolate potential leak paths through the turbine seals when reactor pressure decreases below the HPCI l
range for operation in a post accident scenario. Although the long term solution for this issue is still being developed, the appropriate interim actions have been taken.
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Since the high steam flow and high steam line space temperatures were available to l
provide the PCIS function, the inspector did not consider use of this bypass a i
significant safety issue. Appropriate short term corrective actions were promptly taken, including a revision to eliminate the procedural guidance that allowed use of the bypass switch. This non-repetitive, licensee-identified, and corrected violation
is being treated as a non-cited violation, consistent with Section Vll.B.1 of the NRC Enforcement Policy. (NCV 98-13-01: Bypass of HPCI and RCIC Low Pressure Isolation Signals)
lll. Engineering E8 Miscellaneous Engineering issues i
E8.1 Review of Open items (92902)
The following open items were reviewed for closure based on additional information from the licensee and sampling of the licensee's corrective actions.
Some of the issues discussed in this section were originally identified during an NRC Architect / Engineer (A/E) Inspection, 50-271/97-201 Inspection Report 50-271/97-10 documented a followup engineering inspection tnat evaluated the A/E Team findings and identified potential escalated enforcement issues (Eels). Following an enforcement conference on March 2,1998,the Eels were cited as violations by letter dated April 14,1998. All of the Eels were converted to violations in the April 14,1998,lettar, with three exceptions. The three Eels that were determined to not have been violations are closed in this inspection. A matrix showing the disposition of the original A/E inspection items is provided as an attachment to this report entitled, Engineering inspection Open item Cross Reference.
J (Closed) VIO 97-531-01013: Maximum Torus Temperature (also reference eel 97-10-01a)
The licensee identified that Technical Specification (TS) Amendment No. 88, which permitted operation with the suppression pool temperature above 90 F, was not consistent with that specified in the Final Safety Analysis Report (FSAR). The licensee documented this deficiency in Event Report (ER) 96-0644, dated November 2,1995. The licensee performed an operability determination, Basis for
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Maintaining Operation (BMO) 95-05 and, in response to this BMO, established an administrative limit of 90*F for the suppression pool temperature. Tha NRC architect / engineer (A/E) inspection team documented in inspection report
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l 50-271/97-201(A/E report) that the licensee had failed to maintain maximum
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suppiession pool temperature consistently within the value used in the FSAR Chapter 14.6.3 accident analysis. This discrepancy was cited in inspection report l
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50-271/97-10as an apparent violation of 10 CFR 50, Appendix B, Criterion lil,
" Design Control."
The licensee identified their containment initiatives, including a reduction in the normal operating suppression pool temperature limit, in their letter BVY 97-152. VY l
submitted a revised suppression pool temperature calculation in support of their TS chnge request No. 204 via their letter BVY 98-102, dated July 10,1998. The inspectors confirmed that the calculation, VYC-1628, Rev. O, dated April 27,1998,
" Torus Temperature Response to Large Break LOCA and MSLB Accident Scenarios,"
was based on an initial torus temperature of 90*F.
In their response letter, BVY 98-73, dated May 14,1998, VY stated they had issued procedure, AP-6043, " Document Change Request," dated March 31,1998, to add controls to changes in plant de ign that do not result in physical changes to the plant. The inspectors confirmed that this procedure contained appropriate controls for design input, design output and review and approval. VY also indicated that additional improvements were planned to identify where technical specification setpoints were used in engineering calculations.
Based on: (1) The licensee having submitted the TS change request that was supported by a validated suppression pool temperature calculation, and (2) the
!icensee having established appropriate design control, the inspectors closed this item.
(Uodated) VIO 97-531-05014: Residual Heat Removal Net Positive Suction Head (also reference eel 97-10-01 b)
The A/E team identif.ad that the licensee had used non-conservative curve-fit data instead of actual vendor data to demonstrato NPSH margin for RHR pumps. In response to the A/E finding, the licensee issued ER 97-0664, dated May 29,1997, and memorandum VYS-60/97, dated June 6,1997, which justified operability based on conservatism documented in other supporting calculations. These corrective actions had been confirmed during inspection 50-271/97-10 but the original issue was cited as apparent violations of 10 CFR 50, Appendix B, Criterion Ill, " Design Control."
in their response letter, BVY-98-73, VY indicated that calculation VYC-808, " Core
Spray and Residual Heat Removal Pump Net Positive Suction Head Margin following a Loss of Coolant Accident with Fibrous Debris on the intake Strainers," had been revised on January 13,1998 to correct the error. VY also indicated that engineering personnel were trained on the need for verifying the validity of l
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calculation inputs. Also, VY indicated that the review of critical system calculations was an element in their design basis document (DBD) validation process.
The inspectors confirmed that calculation VYC-808, Revision 4, dated May 14, 1998, included a revised curve-fit which included the pump vendors recommended value for flow at 4600 gpm. The inspectors also connrmed that the results of the calculation concluded that sufficient margin existed for the limiting conditions evaluated.
At the time of the inspection, the review of critical system calculations (an element in their DBD validation process) had not yet been established. The Configuration Management improvement Project (CMIP) manager expected this item to be completed by December 31,1998. This item remains open pending NRC review of the CMIP response.
(Closed) VIO 97-53';-08014: RHR Minimum Flow Protection (also reference eel 97-10-01c)
The A/E team identified that the licensee had failed to correct the discrepancy between the RHR pump minimum flow and the pump vender's documented minimum flow protection requirements. At VY, the minimum flow lines of both RHR pumps had orifices restricting the flow to 350 gpm. The pump manufacturer recommended a continuous minimum flow of 270G gpm with a one time only allowence rf 350 gpm for 30 minutes based on a pump in g :od condition, inspecta cort 50-271/97-10 documented this issue as a violation of 10 CFR 50, Appendix B, Criterion 111.
In their response letter, BVY-98-73, VY stated that they had performed testing to determine vibration levels at the originally specified minimum flow value. The test results showed that the pumps operated with acceptable vibration levels. The licensee had submitted the test results to the pump vendor, and the vendor had established new minimum flow requirements permitting pump operation with a minimum flow of 350 gpm for up to four hours.
The licensee also had stated that they had replaced the flow instruments to ensure the pumps operate within the established minimum flow range. During a previous inspection (50-271/98-12),the inspectors had confirmed that the operating and surveillance procedures had been revised to add appropriate cautions concerning operation at minimum flow conditions.
The inspectors reviewed the pump vendor's NPSH/ Minimum Flow Study, F-97-10782, dated May 26,1998, (included in calculation VYC-808 as attachment No. 5) and confirmed that this latest study included the results from the licensee's
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pump testing. The inspectors confirmed that the RHR flow instrument loops had been upgraded to safety class electrical (SCE) as part of engineering design change request (EDCR) No.97-420. The inspectors also confirmed that the design change and safety evaluation adequately addressed the change from an analog indication to a microprocessor driven digital indicator.
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l (Closed) VIO 97-531-06014: Failure to the Correct Calculation Assumption for the Room Cooler Thermal Performance (also reference eel 97-10-01 e)
Inspection report 50-271/97-201 documented that the licensee based the acceptance criteria for room cooler thermal performance on a fouling caused by l
silting. The A/E team identified that the licensee had failed to revise the fouling assumptions used in calculation VYC-1329 (Room Refrigeration Unit (RRU) 7 and 8 l
Performance Assessment) as a result of the heat exchanger inspection of the safety-related unit coolers RRU 7 and 8 in April 1995. That surveillance inspection confirmed that the increase in pressure drop across the cooler coils was not due to silting, but possibly due to micro-fouling. This failure to revise the calculation assumption was cited in inspection report 50-271/97-10as an apparent violation of 10 CFR 50, Appendix B, Criterion Ill.
In their resoonse letter, BVY 98-73, VY indicated they had developed thermal performance test method that eliminated the need for calculation VYC-1329. The inspectors' discussions with the responsible system engineers and the design engineers confirmed that the basis for the new performance test method was contained in calculation VYC-1340A,"RRU 7&8 Thermal Performance Acceptance Criteria," Revision 1, dated March 11,1998. In addition, the inspectors confirmed that the new performance test method was included in procedure OP-4181,
" Service Water / Alternate Cooling System Surveillance," Revision 29, dated April 18,
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The licensee acknowledged a weakness in their configuration management that failed to update the salculation assumption. This area will be followed as part of the NRC review of the CMIP as noted above for VIO 97-531-05014.
The inspectors considered the licensee's specific corrective actions of the revised performance test methodology contained in the surveillance procedure and the calculated bases for the acceptance criteria adeauate. This item is closed.
(Uodated) VIO 97-531-07014: Design Calculations (also reference eel 97-10-01 h)
The A/E team identified that the licensee had fai!ed to use the latest documents for calculation inputs, in accordance with engineering instruction WE-103, " Engineering Calculations and Analyses." The A/E team observed that, even though the licensee
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had embarked on a configuration management program, including design basis I
documentation and verification, two recent calculations (VYC-298, " Battery Sizing for VY 125 Volt Station Batteries A-1 and B-1," Revision 10. dated April 22,1997 and VYC-1349,"125 Volt DC Control Circuit Voltage Drop Study for Batteries A-1 and 8-1," Revision 1, dated April 30,1997) did not use the latest documents for
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in their response letter, BVY 98-73, the licensee indicated that calculation VYC-298 had been revis9d, and that no technicalissue was found during their review of calculation VYC-1349, Revision 1, as a result of the use of superseded references.
The inspectors' review of both calculations confirmed that the revised calculations incorporated the latest reference information in addition, the inspectors confirmed that the licensee had documented this issue (ER 97-0665), had entered it into their corrective action program, and that the Electrical and l&C design engineering department had reviewed other calculations for similar concerns and had documented the results of that review in department memorandum VYE 80/97, dated August 11,1997.
The licensee also indicated that the process improvement element and the DBD program of their configuration management improvement pr, ject (CMIP) would address similar issues in other critical system calculations. This item remains open pending the licensee's identificntion and development of the process improvements.
(Closed) VIO 97-531-01023: Maximum Torus Temperature (also reference eel 97-10-03a)
The A/E team identified that the licensee had failed to evaluate, in a timely manner, the operability of the emergency core cooling system (ECCS) pumps when taking suction from the suppression pool at elevated temperatures. Elevated temperatures of the suppression pool affects the available net positive suction head for the ECCS pumps. This concern was first identified by the licensee in 1994 as Engineering Deficiency Report 94-05 at Yankee Atomic, and later entered into the licensee's problem reporting system as ER 95-0644, dated November 11,1995. This lack of a prompt evaluaGon of a potential safety significant issue resulted in an apparent violation being documented in inspection report 50-271/97-10.
s in their response letter, BVY-98-73, the licensee indicated that the root cause of this violation was a lack of focus on a complex issue due to a non-integrated corrective action process. The licensee also indicated that an internal self-assessment performed in late 1997 recommended improvements in their Basis for Maintaining Operatinn (BMC) process and establish performance indicators for tracking and timeliness. Additional process improvements mentioned were the ER backlog reduction efforts and improvements for management of work and open corrective action issues.
The inspectors confirmed that the former Yankee Atomic engineers (now employed by VY directly) and the VY site staff were now vforking with an integrated corrective action process. The inspectors found that the licensee had not yet issued a revised BMO guideline but expected to issue the revised guidelines in the form of a plant procedure some timc in the future by Systems Engineering.
The inspectors attended the weekly engineering status review meeting with the vice-president of engineering and all the engineering managers. The inspectors
,i confirmed that engineering management had started to develop performance
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indicators for use in the weekly engineering management meeting but the indicators had not progressed to their satisfaction.
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The inspectors confirmed that the specific evaluation of torus temperature had been
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adequately addressed in the licensee's technical specification change request (BVY 98-69, dated May 8,1998) and their technical supporting calculation package (BVY 98-102, dated July 10,1998).
This item was closed based on engineering's progress in the improvements for their corrective action tracking progres:; and their submittal of supporting data for their
technical specification change request for torus temperature.
I (Closed) eel 97-10-Q3b: Failure to enter a Limiting Condition for Operation (LCO)
Statement During RHR System Testing Inspection report 50-271/97-10 confirmed that, when the RHR system entered the i
suppression pool cooling mode, the licensee took one subsystem out of its design basis line-up, leaving the RHR system susceptible to a single failure. Technical Specification (TS) 3.5.A requires the residual heat removal / low pressure coolant
injection mode (RHR/LPCI) systems to be declared inoperable if either sub-system is l
not available to perform its safety function. This was previously identified in NRC inspection report 50-271/93-80as a weakness in the administrative control process and as an unresolved item by the A/E team in inspection report 50-271/97-201.
The licensee had planned to address this concern as part of its improved Technical Specification (ITS) program. When the iTS program was delayed, the licensee failed to take appropriate action to address the operation of the RHR system'outside its design basis.
Inspection report 50-271/97-10 documented that the licensee had failed to enter a limiting condition for operation (LCO) for this condition on April 17,1997, prior to the A/E team inspectic,a, but when the "A" LPCI subsystem was operated on September 26,1997,it had been entered on the inoperable list. Inspection report j
L 50-271/97-10also documented that the licensee had administratively addressed the LPCl/ torus cooling issue. The inspectors did not identify any instance where one
division was inoperable while the other division was in the suppression pool cooling mode, and therefore, the limiting condition for operation of TS 3.5.A was met, i
The licensee contracted with an independent consulting company to identify any similar generic concerns by performing a review of the existing surveillance procedures. The contractor's review, documented in a report, Technical Specifications Operability impact Assessment, dated July 17,1998, resulted in the identification of 103 procedures (in a population of 177 technical specification related surveillance procedures) which will need revision to address entry into LCOs.
i-Six of the 103 identified deficiencies could have resulted in conditions adverse to quality, and two of those six could have resulted in a loss of safety function. Those six deficiencies had been entered into the VY corrective action system as event reports (ERs). The inspectors confirmed that one of the two procedures, OP 4114, i
Standby Liquid Control System, had already been corrected using Department Instruction (DI 98-300) written against the existing procedure.
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The inspectors were informed that the remaining procedure, OP 4122, Auto l
Blowdown System Surveillance, was only used during a refueling outage and was l
scheduled to be revised prior to its use during the next outage. The inspectors
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confirmed that the discrepancy had been adequately identified and evaluated in
. ER 98-1434. A requirement to revise the procedure prior to its next use had been entered into the licensee's AP 0028 tracking system.
Based on the extensive review commissioned by the licensee and their corrective actions completed, and the fact that no instance of both divisions being inoperable
were identified by the inspector, this item is closed.
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(Closed) eel 97-10-03c: Service Water Pump Operability NRC Inspection Report 50-271/97-201 identified that the licensee had failed to revise non-conservative technical specifications, section 3.5.D.3, for SW pump operability in a timely manner after the licensee discovered the condition on February 24,1997, as documented in ER 97-0198. During a previous inspection,
-(IR 50-271/97-10)the inspector confirmed that the licensee had revised procedure OP-2181, Service Water / Alternate Cooling Water Operating Procedure, to add the appropriate action statement via Deoartment Instruction (DI) 97-28.
In letter BVY 98 52, dated April 23,1998, VY submitted a TS change request for the requirement of station service water and alternate cooling systems. As noted in letter BVY 98 73, dated May 24,1998,VY recognized a weakness in their licensing
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function and increased staffing in that area, including the addition of a dedicated manager for the group. Based upon the administrative controls established by VY, the submittal of the TS change request, and the increased attention in the licensing area, the inspector considered the licensee's corrective actions adequate. The
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inspectors did not find any examples where the SW system was inoperable and VY had relied on the alternate cooling water system (as would have been allowed by the non-conservative TS). Therefore no violation of NRC requirements was found and this item was closed.
(Closed) VIO 97-531-09014: EDG Susceptibility to Tornado Effects and Use of PRA to Justify EDG Operability (also reference eel 97-10-03d)
The A/E team had raised a concern on the use of PRA for operability determinations and had documented that concern in inspection report 50-271/97-201. During the follow-up inspection (50-271/97-10),the inspectors reviewed the closed ER 97-0584 package which determined that the emergency diesel generators (EDGs) were operable snd that the identified conditions were not deemed reportable. However,
'the inspectors had determined that the Basis for Maintaining Operation (BMO)
process had not been initiated.
The inspectors' review of a iicensee internal memorandum dated May 20,1997 indicated that VY was unsure of the application of PRA criteria for this issue based on their statement that, "...use of a probabilistic approach for design basis tornado considerations is currently open to question..." The May 20,1997, memorandum also stated that "...it is possible we [VY] will need to separately submit an amendment for NRC review and acceptance...," referring to the VY's not-yet submitted Individual Plant External Event Evaluation (IPEEE). Further, the May 20,
- 1997, memorandum stated there was an " apparent lack of specificity in the FSAR"
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with respect to the specific mbsile protection to be afforded the EDG support l
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Based upon these documented VY staff conclusions, the inspectors had determined l-that VY had not adequately evaluated the potential challenge to the exposed diesel l
generator support systems from the effects of tornados, including tornado missile strikes. The inspectors also determined that the application of PRA methods to support the EDG operability determination was in conflict with Generic Letter 91-18, section 6.9.
l In their response letter BVY-98-73, the licensee indicated that BMO 97-44, i
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" Tornado Missile Protection - EDG Service Water," was approved and issued on November 3,1997, and resulted in a number of corrective actions including the development of an external events design basis document. The inspectors confirmed that the issue of an external events DBD, which would address tornados as well as flooding and hurricanes, was being tracked in the licensee's AP-0028 commitment tracking process, with a due date of February 1,1999. Based on the licensee's immediate and committed (AP-0028) corrective actions to address tornado generated missile protection design bases, this item was closed.
(Closed) eel 97-10-03e: Standby Battery Charger in the A/E inspection Report 50-271/97-201,the NRC described that VY identified a l
non-conservative technical specification (TS 3.10) regarding a potential single failure vulnerability of the DC system when operating with the standby battery charger for an indefinite period of time, inspection report 50-271\\97-10 documented that the licensee had originally identified this deficiency as an open item in the 125 volt DC System Design Basis Document on February 21,1997. The inspection report also documented that the licensee had issued a department instruction on May 5,1997, revising operating procedure OP-2146, Operation of 125 Volt Battery Chargers to enter the LCO when operating with the spare battery charger as a compensatory measure but had failed to identify the non-conservative technical specification and take prompt corrective action.
The inspector confirmed that the licensee had initiated a technical specification change request to add additional controls on the use of the spare battery charger.
In addition, the inspector could not identify any instance where the licensee had used the spare charger in a cross-division arrangement. Based on the licensee's corrective action for the self identified issue, and further NRC evaluation of the corrective action timeliness, the NRC concluded that no violation occurred.
Therefore, this item is closed, (Closed) VIO 97-10-06: Inadequate Test Cor. trol t.
Inspection report 50-271/97-201 documented an example of inadequate control of test instruments. Specifically, VY failed to establish measures to assure that i
instruments required for the RHR beat exchanger performance testing, including the i
Emergency Response Facility Information System, were appropriate for the l
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application. The licensee initiated Event Notification 32285, dated May 6, 1997,in accordance with 10 CFR 50.72, and issued multiple internal event reports in response to this concern. The licensee determined that improvements were required in flow instrumentation and possibly temperature instrument accuracy to
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enhance the thermal performance testing of the RHR heat exchangers. NRC Inspection 50-271/97-10found that the instruments were calibrated at the wrong flow condition and the small changes in temperature across the heat exchanger led to large uncertainties.
In their response letter, BVY 98-33, the licensee indicated that new baseline tests would be performed using supplemental, high accuracy instrumentation and using the expertise of an independent industry expert.
The inspectors confirmed that the licensee had employed an independent consulting firm to assist them in the preparation, implementation and evaluation of new baseline testing for the RHR and SFPC heat exchanger thermal performance testing.
The test setup utilized calibrated ultrasonic flow transmitters to increase the accuracy of the test results. The consultant's software model of the heat exchanges was used for the performance data evaluation. The inspectors confirmed that the licensee had designated the system engineer responsible for the implementation of future heat exchanger performance tests similar to the new baseline tests. The inspectors noted that the licensee had initiated AP-0028 commitments to repeat the heat exchanger tests during the following two refueling outages following the recommendations of generic letter GL 89-13, " Service Water System Problems Affecting Safety Related Equipment." Based on the results of the licensee's testing to date and their internal commitment to repeat the tests during the following two refueling outages, the inspectors closed this item.
(Closed) VIO 98-80-02: Inadequate implementation of the Corrective Action Program inspection report 50-271/98-80, documented examples of the licensee's inadequate implementation of several corrective actions associated with the degraded control room ventilation (CRV) system. Examples included the licensee's failure to recognize and assess the concerns in the CRV system, following the issuance of GL 88-14, " Instrument Air Support System Problems Affecting Safety-Related Equipment." Later, upon recognition in 1996 that the CRV system was degraded, the licensee failed, again, to perform a comprehensive root cause evaluation of this condition, as required by their event report process, and also failed to perform an operability determination, as required oy their corrective action procedures, in their response letter, BVY 98-123, dated August 17,1998, the licensee agreed with this violation. The licensee stated that the first problem occurred due to a lack of appropriate detail existing in their design drawings used during the review of GL 88-14. The second issue was caused by a narrow scope of their review of GL 88-14. The third concern was caused by a lack of recognition of the discrepancy between the as-built condition and the FSAR system description of the CRV syste.
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To address the immediate concerns, the licensee re-evaluated the operability of the as-configured CRV system (BMO 98-23 and Safety Evaluation No.96-052, Revision 1), on July 9,1998 ar'd determined that the system was operable. The licensee concluded that a failure of CRV due to damper failure caused by a loss of instrument air was unlikely because of the redundant design features of the j
instrument air system, and procedures in place to operate the dampers manually
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when required.
The inspectors noted that the licensee had also improved the root cause determination process, since April 1997, by assigning a dedicated and experienced team to perform the root cause investigations. The inspectors' interview with one team member indicated that the individual was experienced and knowledgeable of
systems and process.
Additionally, the licensee had established a long term action plan to improve the f
corrective action implementation process, as tracked by their commitment tracking system items INS 988002-03,-04,-05 and -06. This action plan included: (1) to evaluate existing configuration /FSAR discrepancies, (2) to evaluate the methodology l
and work instructions used for previous review of GL 88-14, and (3) to provide training to promote more thorough evaluations to engineering support personnel.
This action plan was scheduled to be completed by the end of 1999. The i
inspectors considered this plan acceptable.
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The inspectors reviewed the licensee's operability evaluation of the CRV system and I
safety evaluation and found no concerns. Since all the long term corrective actions, including a final design review of this system for consistency with the FSAR, were l
appropriately being tracked by the licensee in their corrective action process (AP-0028) and were scheduled to be completed by the end of 1999, the inspectors closed this item, i
l (Closed) URI 97-10-14: Delayed Access Power Supply l
i The VY FSAR Section 8.1, FSAR Appendix F, and VY Technical Specifications section 3/4.10 described the Vernon tie line as a delayed access supply required by General Design Criteria 17. In addition, the NRC safety evaluation for the station
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blackout (SBO) implementation at VY described the Vernon tie line as the SBO alternate alternating current (AC) supply. VY had justified the dual function i
(SBO/GDC 17) of the Vernon tie based on its reliability, independence, capacity ano capability to be available within ten minutes. This position was documented in letters to the NRC on March 26,1997 (BVY 97-40) and April 24,1997 (BVY 97-
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54).
As a result of a licensee review of industry concerns on the availability of the delayed offsite power supply and discussions with the NRC on the licensing basis
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for VY, the licensee recognized that they should not be taking credit for the Vernon tie line as both a delayed access supply and the SBO alternate AC supply. VY had prepared two Bases for Maintaining Operation (BMO 97-03, Revision 1 and BMO 97-18, both dated April 28,1997), justifying cuntinued operation with the Vernon tie acting as the delayed access power supply and withc+. having separate dedicated alternating AC supply to conform to the requirements of 10 CFR 50.63,
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" Station Blackout." The VY justification was based on the reliability of the offsite power systems and the administrative limits which supplemented the existing technical specifications. These BMOs were submitted to the NRC (BVY 97-57, dated April 29,1997) and the NRC indicated that a VY schedule to achieve full compliance with 10 CFR 50.63 prior to restart from the next refueling outage would be acceptable.
The licensee identified their proposed resolution for this concern to the NRC (BVY 97-67, dated May 29,1997) prior to the end of the A/E team inspection. In inspection 50-278/97-10,the inspectors confirmed that the licensee initiated a design to install a generator disconnect switch to reduce the time required to backfeed the plant through the main transformer. In their letter BVY 98-71, dated May 12,1998, the licensee indicated the modification was complete.
During this inspection, the inspectors confirmed that a no-load generator disconnect switch was installed and the Vernon tie was once again considered the alternate AC supply to support SBO. The inspectors also confirmed that TS Amendment No.155 had been incorporated into TS 3/4.10, to add restrictions on startup and continued operation with a diminished offsite power supply. The inspectors also confirmed that the TS change also removed credit for the Vernon tie as a delayed access offsite power source.
The inspectors concluded that, even though the use of the Vernon tie as both a delayed access power supply and the SBO alternate ac supply was a weakness in the original VY design, the Vernon tie did not violate the NRC requirements of 10 CFR 50.63 for an alternate supply as defined in 10 CFR 50.2, " Definitions."
Therefore, this item was closed.
(Withdrawn) VIO 98-80-01: Failure to Perform Adequate Safety Evaluation (Withdrawn) VIO 98-80-03: Failure to issue Required Reports to the NRC Inspection Report 50-271/98-80 identified two items related to degraded conditions in the Control Room Ventilation System as violations of NRC regulations. In their response letter, BVY-98-123, dated August 17,1998, the licensee provided additional information. After further evaluation of the issues, the NRC withdrew the violations. This was documented in a letter dated October 26,1998.
(Closed) VIO 98-06-01: 10 CFR 50.73(A)(2), Failure to issue an LER During the 1997 emergency preparedness annual exercise, VY identified that the vocess of containment flooding in response to a large break LOCA could cause containment pressure to exceed the actuating pressure of the hardened vent system rupture disc. VY's initial response to the issue was to close the downstream isolation valve, and to initiate a BMO to justify continued operation in this configuration and to develop corrective action. VY did not consider the issue to be reportable to the NRC, based on the conclusion that the hardened vent system would only actuate in response to beyond-design-basis plant condition __
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BMO 97-15, " Torus Vent Rupture Disc / EOP Design Basis," was reviewed in inspection report 50-271/98-06. The inspector determined that the issue constituted a reportable condition under 10 CFR 50.73, and the subject violation was issued for failure to submit the required LER. An inspector follow-up item, IFl 98-06-02, was also generated to track the technical resolution of the rupture disc issue.
VY determined that the cause of the violation was cognitive error on the part of plant personnel performing reportability determinations. In this case, the relationship between the plant design bases, ihe emergency operating procedures, and the severe accident management strategy guidance was misinterpreted with respect to the reportability requirements. As long term corrective action, VY changed their reporting procedure to emphasize conservatism in making reportability determinations, and initiated benchmarking and process improvement efforts. The inspector determined that these actions appropriately addressed the problem, although it is too early to assess their effectiveness.
VY's response to the violation was to issue LER 98-010-00. The inspector also conducted an in-office review of this LER as part of the violation response review.
The inspector concluded that the LER appropriately discussed the event and that the root cause determination was reasonable. However, the inspector noted that the level of qualification of the rupture disc downstream isolation valve, TVS-86, was overstated. Specifically, in the event description, the LER states that, "A safety class motor operated isolation velve in the vent line was shut," shortly after the issue was identified. However, BMO 97-15 indicated that, while TVS-86 was installed and tested as a safety class valve, it was not being maintained as such; it was not included in the motor operated valve test program, and it is not powered from a safety class electrical source. The qualification requirements for TVS-86 will be examined as part of the overall technical resolution of this issue during closure of IFl 98-06-02. Accordingly, VIO 98-06-01 and LER 98-010-00 are closed.
(Closed) IFl 97-04-04: RHRSW Flow Requirements This IFl was opened to track the licensee's resolution of a potentially degraded /unanalyzed condition involving the R!lRSW system that was initially reported by VY on May 6,1997,in accordance with 10 CFR 50.72 (reference EN 32285). Subsequently, Licensee Event Reports 97-012-00 and 97-012-02, were issued on June 6,1997, and November 21,1997. The condition was identified by VY during preparations for NRC Team inspection 97-201. NRC Inspection Report 97-10, issued February 5,1998, reviewed the NRC Team's open item relative to this VY discovery. A violation against 10 CFR 50 Appendix B, Criterion XI, Test Control, was issued because instruments used in activities affecting quality were not properly selected or controlled to maintain accuracy within necessary limits to assure that the test requirements were satisfied. This violation was identified as VIO 97-10-06. Based on the previous NRC inspection of this issue, and the violation which was issued, no additional issues remain to be resolved. Since the licensee's corrective actions for the above violation will be reviewed, this IFl is administratively close..
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E8.2 In-office Review of LERs Related to Enaineerina (90712)
An in-office review of the following licensee event reports (LERs) was performed to assess whether further NRC actions were required. The adequacy of the overall event description, immediate actions taken, cause determination, and corrective actions were considered during this review. The following issues were closed-out based on the in-office review.
(Closed) LER 98-010-00: Failure to Foresee System Interactions ANws the Installation of a Wetwell Hardened Vent System Which Presents Conditions Which Challenge Containment Systems in the Event of a Postulated LOCA The LER is discussed, and dispositioned, in section E8.1 of this inspection report in conjunction with VIO 98-06-01. This LER is administratively closed.
(Closed) LER 98-012-00: Failure to Adequately Define Standards for Reducing Contact Forces Caused by Sharp Edges on MOV Internals Results in Six Primary Containment Isolation Valves Being Declared inoperable This event occurred during the 1998 refueling outage, when troubleshooting of one of the subject valves identified that edges of valve disc and seat were not chamfered as required. For this valve, chamfered edges are required to ensure proper valve operation during the high flow conditions that would exist if the associated line were to break.
VY's preliminary determination of the root cause of this event was failure to adequately define and communicate job performance standards. During earlier work on the subject valves, the requirements for chamfering was still under development.
The tentative value of 1/32-inch chamfer was apparently interpreted to mean, "no l
sharp edges," which was then used as a qualitative acceptance criteria.
As immediate corrective action, VY identified all other valves that were required to have chamfered internals (i.e., containment isolation valves that may be required to operate under blowdown conditions) and inspected the internals. Where necessary, additional work was done to establish the required chamfer. Long term corrective action include establishing additional training requirements and procedural enhancements, and are expected to be complete by the end of 1998.
This issue was reviewed in inspection report 50-271/98-08,section E1.1. Further NRC actions relating to the event are being tracked by two unresolved items which resulted from that inspection. Accordingly, LER 98-12-00is closed.
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E8.3 (Closed) LER 98-018-00: ASME Section XI Code VT-3 Examination not Completed Following Repair of Main Steam Isolation Valve due to the Omission of Relevant Information from a Plant Procedure a.
Insoection Scope (90712,92700)
After an in-office review of LER 98-018-00,the inspector determined that additional on-site followup was required, b.
Observations and Findinas i
During the 1998 refueling outage, VY identified two flaw indications in the body of main steam isolation valve (MSIV) V2-80B. The indications were ground out and the affected areas were restored by weld repair. The repairs were inspected by j
magnetic particia and radiographic examination. Following reassembly, the valve was stroke tested satisfactorily and was subject to normal operating pressure during the reactor pressure vessel operational system leakage test. Subsequently, VY identified that a visual (VT-3) examination of the repairs had not been performed as required by section XI of the ASME code. Technical specification 4.6.E requires that inservice inspections be performed in accordance with this code.
The LER indicated that the root cause of this event was omission of relevant information from the plant procedure that provides general guidelines and instructions for maintenance activities on safety related valves. Corrective action consisted of a revision to that procedure, OP-5201," Safety System Valves," to include the VT-3 pre-service inspection requirement. However, the root cause analysis that was generated as a result of the VY event report indicated that the
root cause was inadequate work control process. Specifically, the requirement to perform the pre-service VT-3 inspection (contained in procedure YA-VT-11) was included in the work order only by reference to the applicable procedure, rather than
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being incluried as a step within the detailed work instructions. In addition to the corrective action specified in the LER, the ER root cause analysis indicated that an ASME Section XI repair / replacement program would be developed to provide guidance and direction on required code repairs and inspections.
In addition, the inspector noted that the event date (May 28) was prior to plant startup from the refueling outage (June 1). At the time of discovery, the plant was in cold shutdown and had been since before work on the MSIV had commenced.
Therefore, the valve had not yet been required to perform any safety function, and, had the issue been resolved prior to startup, no violation of the code or TS would have occurred. However, this aspect of the event was not addressed in the LER.
After discussions with maintenance management, VY initiated an investigation of this issue. Pending completion of VY's investigation, this issue remains unresolved.
(URI 98-13-02: ASME Pre-service inspection of MSIV Not Resolved prior to Plant Startup)
The inspector concluded that LER 98-18-00 was weak, in that the root cause did not identify problems with the disposition of the ASME Code inspection or VY's
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apparent decision to restart the plant prior to resolving the issue, which caused the TS violation described by the LER. These findings were discussed with VY management. VY initiated ER 98-2084 to investigate these issues, and VY's plant l
manager stated that a supplemental LER will be issued. The inspector considered VY's response to this issue appropriate and a final determination regarding potential
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violations of TS requirements for ASME Code implementation, or adequacy of corrective actions required by 10 CFR 50 Appendix B, will be evaluated during review of the unresolved item identified above. Therefore, LER 98-018-00is closed.
c.
Conclusions Weaknesses in VY's disposition of a missed ASME Code inspection were not
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identified during the licensee's root cause investigation. These problems were identified during an NRC review and followup inspection associated with Licensee Event Report (LER) 98-018-00. VY has initiated an Event Report to investigate and assess the previous corrective actions. An unresolved item has been opened pending additionalinformation from the licensee and a supplement to the LER.
E8.4 (Closed) URI A8-08-06: Impact of Low Blowout Panel Set Point on Secondary Containment a.
Insoection Scoce (92903)
The inspectors reviewed VY's evaluation and corrective action associated with the blowout panel set point deficiency. The unresolved item was opened pending VY's assessment of the consequences of the lower-than-desicn set point and a review of potentially applicable 10 CFR 50.73 reporting requirements, b.
Observations and Findinas in 1997, VY identified that two Reactor Building (RB) blowout panels located in the main steam tunnel would relieve at approximately 0.10 psi, rather than the 0.25 psi described in FSAR section 5.3, " Secondary Containment System." This deficiency was identified as part of the corrective action for BMO 96-018, and the licensee concluded that there was no adverse impact on the high energy line break analysis or the tornado design basis. VY initially determined that the deficiency was not reportable under 10 CFR 50.72 or 50.73, as a condition outside the plant's desipn basis, because the blowout panel set point was listed in the FSAR design description of secondary containment, and not in the section labeled " Safety Design Basis." The inspectors observed this initial evaluation was narrowly focused and was not based on existing guidance provided in NUREG 1022, Revision 1, " Event Reporting Guidelines,10 CFR 50.72 and 50.73.
f The inspector's review found that FSAR section 14.4.4.3, " Containment Damage,"
l states, " Satisfaction of Safety Design Limit 1 for accidents requires that the primary and secondary containments retain their integrities for certain accident situations.
Containment integrity is maintained as long as internal pressures remain below the
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maximum allowable values. The maximum allowable internal pressure [for]
Secondary Containment [is].25 psig." Safety Design Limit 1, listed in FSAR section 14.3, states "The radiological effects of accidents shall not exceed the guideline values of 10 CFR 100." VY determined the basis for these FSAR statements was a j
general reference to the structural integrity of the reactor building and that there is no tie to a specific accident analysis.
The licensee identified that the capability of the original blowout panels was not consistent with their intended design. The degraded condition was adequately evaluated to assess potential consequences, and no adverse impact on the design basis functions for HELB and tornado mitigation was identified. Corrective actions restoring the intended design capability were completed during the 1998 refueling outage. The inspector concluded that because the deficiency had a conservative effect on the HELB and tomado mitigation capability, the degraded condition did not place the plant outside its design basis. The inadequate design of the original blowout panel mechanism is a violation of 10 CFR 50 Appendix B requirements for
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design control. This non-repetitive, licensee-identified, and corrected violation is being treated as a non-cited violation, consistent with Section Vll.B.1 of the NRC
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Enforcement Policy. (NCV 98-13-03: Inadequate Design Control for Blowout Panel Mechanism)
c.
Conclusions
VY identified that two reactor building blowout panels would relieve at, lower internal pressure than specified in the design for high energy line breaks and tornado
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mitigation. Since the deficiency resulted in a conservative response to these
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events, the condition did not place the plant outside its design basis. The licensee-identified and corrected deficiency design control error was dispositioned as a non-
cited violation, consistent with the NRC Enforcement Policy.
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IV. Plant Support
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R1 Radiological Protection and Chemistry (RP&C) Controls
' R1.1 Radioactive Liauid and Gaseous Effluent Control Proarams a.
Inspection Scope (84750-01)
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The inspector reviewed: (1) radioactive liquid and gaseous effluent release permits;
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(2) effluent control procedures; (3) unplanned or unmonitored release pathways; (4)
the 1997 annual effluent report; and (5) the Offsite Dose Calculation Manual (ODCM).
The inspection also included tours of the control room and other selected areas, and
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reviewed selected radioactive gas processing facilities and equipment, effluent and
process radiation monitoring system (RMS), verified the reactor building plant air balance, and air cleaning systems.
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Observations and Findinas All effluent radiation monitors and air cleaning systems were operable at the time of the plant tour. The reactor building was maintained at a negative pressure.
The ODCM providea descriptions of the sampling and analysis programs, which were established for quantifying radioactive liquid and gaseous effluent concentrations, and for calculating projected doses to the public. All necessary parameters, such as effluent radiation monitor s9tpoint calculation methodologies and site-specific dilution factors, were listed. Radioactive gaseous effluent release permits and monthly dose projections were complete. There were no unplanned and unmonitored radioactive gas releases in 1997 and 1998. The reviewed effluent control procedures were detailed.
The-1997 Annual Radioactive Effluent Report provided data indicating total released
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radioactivity for liquid and gaseous effluents. The assessment of the projected j
maximum individual doses resulting from routine radioactive airborne and liquid effluents were included, as required. Projected doses to the public were well below the Technical Specification (TS) limits. There were nn anomalous measurements, omissions, or adverse trends in this report.
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Conclusions The licensee maintained effective radioactive liquid and gaseous effluent control programs. The ODCM contained sufficient specification and instruction to acceptably implement and maintain the radioactive liquid and gaseous effluent control programs.
R2 Status of RP&C Facilities and Equipment R2.1 Calibration of Radiation and Hydroaen/Oxvaen Monitorina Systems a.
Inspection Scone (84750-01)
The most recent calibration results for the following selected effluent / process / area RMS, system flow rates, and hydrogen / oxygen monitors were reviewed:
Radiation Monitorina Systems Steam Jet Air Ejector Offgas Monitors
Main Stack Noble Gas Monitors (Normal and High Ranges)
Augmented Offgas (AOG) Building Noble Gas Monitors
Reactor Building Monitor
Spent Fuel Pool Floor Manitor
Liquid Radwaste Discharge Monitor
Service Water Discharge Monitor
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Flow Rate Measurement Systems AOG Flow Rate Monitor
Main Stack Flow Rate Monitor
Calibration of Hydroaen/Oxvaen Monitors AOG Hydrogen Monitor
Drywell Hydrogen / Oxygen Monitors
b.
Observations and Findinas All calibration results reviewed were within the acceptance criteria as defined by the licensee's procedures. The calibration data indicated that the RMS channels were responding in a linear manner. Chemistry staff tracking and trending efforts provided sufficient information to assess the RMS system performance. As a result of self assessment initiatives, the chemistry staff developed and implemented an action plan to improve RMS reliability.
Calibration results of flow rate measurement systems and hydrogen and oxygen monitors were within the licensee's acceptance criteria.
c.
Conclusions The licensee established, implemented, and maintained an effective radiation monitoring system calibration program, including flow rate measurement systems.
As a result of self-assessment initiatives, the licensee implemented efforts to improve radiation monitoring system reliability. The licensee also established and implemented an effective hydrogen / oxygen monitor calibration program.
R2.2 Air Cleanina Systems a.
Inspection Scope (84750-01)
The most recent surveillance test results (in-place HEPA and charcoalleak tests, air capacity tests, pressure drop tests, and laboratory tests for the iodine collection efficiencies) were reviewed web respect to the standby gas treatment system (SGTS).
b.
Observations and Findinos All surveillance results were within the TS acceptance criteria. The responsible individual had good knowledge of testing methodologies and acceptance criteri.
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Conclusions l
The licensee established, implemented, and maintained an effective Standby Gas l
Treatment System surveillance program with respect to charcoal adsorption surveillance tests, HEPA mechanical efficiency tests, and air flow rate tests.
R7 Quality Assurance (QA)in RP&C Activities R7.1 Radioactive Effluent Control and Effluent Samole Validation a.
Insoection Scooe (84750-01)
The inspection consisted of: (1) review of the 1998 audit; (2) QA policy of the measurement laboratory; and (3) implementation of the measurement laboratory quality control program for radioactive liquid and gaseous effluent samples.
b.
Observations and Findinas Audit findings from 1998 did not identify any significant regulatory or safety issues.
l However, findings and recommendations were identified to improve program performance. Response to audit findings was timely. The scope and technical depth of the audit were sufficient to assess the quality of the radioactive liquid and
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gaseous effluent control programs. Individuals with experience in radioactive l
effluents control and chemistry participated as audit team members.
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The QA Support Program consisted of the quarterly distribution of test or quality control samples and issuance of a performance evaluation report. Quality control charts for the gamma spectrometry and tritium measurement systems were frequently reviewed by licensee staff and used as a mechanism to assess laboratory
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c.
Conclusions The licensee established, implemented, and maintained an effective quality l
assurance program for the radioactive effluent control program with respect to audit l
scope and depth, audit team experience, and response to audit findings. The licensee also implemented an effective quality control program to validate measurement results for radioactive effluent samples.
R8 Miscellaneous RP&C lssues R8.1 Review of Open items (92902)
The following open items were reviewed for closure based on additional information from the licensee and sampling of the licensee's corrective actions where applicable.
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30 (Closed) IFl 97-08-02: Verification of Exhausting Actual and Maximum Fan Capacities Listed in Procedure OP-2611, " Gaseous Radwaste" The exhaust air from radwaste, turbine, AOG, and reactor buildings, and other sources (e.g., SBGT),is released to the environment through the main stack. The maximum exhaust air-flow from these buildings is about 200,000 cfm, as described by Procedure OP-2611. However, monitored air flow of the main stack was only about 150,000 cfm, indicating possible air balance mismatch.
The licensee conducted air flow tests which indicated that air was correctly directed from the above buildings to the main stack. The licensee, however, has not yet completed verification of actual fan capacities as stated in Procedure OP-2611, or verified total air flow rate to the main stack. The licensee indicated that air balance verification remains to be completed. The results of this evaluation will be reviewed in a subsequent inspection of the licensee program. (lFl 98-13 04: Plant Air Balance Verification)
(Closed) IFl 98-08-07: Storm Drain Contamination and Evaluation per Bulletin 80-10 The inspector reviewed the licensee's: (1) safety evaluation, 50.59(a)(2), including dose assessment results; (2) evaluation of radioactive materials discharged via the storm drain system; and (3) volume of sediment and water discharged from storm drains. The scope, depth, and conclusions of the above evaluations were very good. Storm drain contamination / discharged to the environment was well below regulatory requirements. This item is closed.
F8 Miscellaneous Fire Protections issues F8.1 Review of Ooen items (92904)
The following open item was reviewed for closure based on additional information provided by the licensee and a sampling review of corrective actions were appropriate.
(Closed) IFl 97-80-01: Penetration Barriers and Seal Design Verification As documented in NRC inspection 50-271/97-80,the licensee's special project team was performing a special walkdown of all plant penetration barriers and seals to establish a penetration seals database. Since this effort was incomplete at that time, this issue was left as an IFl to track the completion of this effort.
During this inspection, the inspector found that the licensee's contractor completed the walkdown and collected the relevant data to develop the database. Also, the licensee's contractor loaded the collected information into newly developed penetration and barriers seals database. The licensee was in process of revisino the existing procedure OP-0046 (installation and Repair of Fire Barrier Penetration Seals,
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Firebreaks, and Flood Seals) to include administrative control steps to make this
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database usable under their control program from the contractor. Per discussion with the licensee, this changeover was expected to be completed within a month.
The inspector's review of the database found that the licensee had assigned a unique number to each of the plant penetration fire barriers, along with information included such as, location and type of barrier. Additional critical characteristics of each penetration, such as applicable design draw'.ng, type of barrier or seal material, depth of barrier, type of configuration as qualificd per test report, past history of repair, and its originalinstallation date and asse,ciated design documentation cross-references were also planned to be captured as a part of the instruction manual that was being developed.
The inspectors concluded that the licensee had taken adequate steps to ensure the development of an appropriate penetration and barrier seal database for the station.
Therefore, this item is closed.
V. Management Meetings X1 Exit Meeting Summary l
The resident inspectors met with licensee representatives periodically throughout the inspection and following the conclusion of the inspection on December 16, 1998. At that time, the purpose and scope of the inspection were reviewed, and the preliminary findings were discussed. NRC region based inspectors debriefed VY management at the conclusion of their respective on-site inspections. The results of an in-office review of select engineering open items was communicated to VY management during a telephone debrief on November 18,1998. The licensee
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acknowledged the findings presented.
The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.
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ITEMS OPENED, CLOSED, AND DISCUSSED
OPENED URI 98-13-02:
ASME Pre-service Inspection of MSIV Not Resolved Prior to Plant l
Startup IFl 98-13-04:
Plant Air Balance Verification CLOSED
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l VIO 95-22-01:
Inservice Test (IST) Program Scope VIO 95-22-02:
Valve Inservice Test Deficiencies VIO 95-22-03:
Inadequate Testing of Minimum Flow Check Valves VIO 95-23-01:
Inadequate Inservice Testing of Stop Check Valves
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LER 97-013-00:
Unknown Cause(s) Result in Extended Actuation Times for Three-Out-of-Four Safety Relief Valves LER 98-009-00:
Main Steam isolation Valve Leakage Exceeds Technical Specification Limit Which Could Have impacted the Ability of a System to Mitigate Consequences of an Accidant LER 98-011-00,01: High Pressure injection and Reactor Core Isolation Cooling Systems Low Steam Supply Pressure Isolation Function Bypassed During Start-up Contrary To Technical Specification Requirements VIO 97-531-01013: Maximum Torus Temperature (also reference eel 97-10-01 a)
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VIO 97-531-08014: RHR Minimum Flow Protection (also reference eel 97-10-01 c)
VIO 97-531-06014: Failure to the Correct Calculation Assumption for the Room Cooler Thermal Performance (also reference eel 97-10-01 e)
VIO 97-531-01023: Maximum Torus Temperature (also reference eel 97-10-03a)
eel 97-10-03b:
Failure to enter a Limiting Condition for Operation (LCO) Statement During RHR System Testing eel 97-10-03c:
Service Water Pump Operability
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VIO 97-531-09014: EDG Susceptibility to Tornado Effects and Use of PRA to Justify EDG l
Operability (also reference eel 97-10-03d)
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l eel 97-10-03e:
Standby Battery Charger
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VIO 97-10-06:
Inadequate Test Control A-1
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VIO 98-80-02:
Inadequate Implementation of the Corrective Action Program URI 97-10-14:
Delayed Access Power Supply
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VIO 98-06-01:
10 CFR 50.73(A)(2), Failure to issue an LER IFl 97-04-04:
RHRSW Flow Requirements LER 98-010-00:
Failure to Foresee System Interactions Allows the installation of a Wetwell Hardened Vent System Which Presents Conditions Which Challenge Containment Systems in the Event of a Postulated LOCA LER 98-012-00:
Failure to Adequately Define Stendards for Reducing Contact Forces Caused by Sharp Edges on MOV Internals Results in Six Primary Containment isolation Vaives Caing Declared inoperable LER 98-018-00:
ASME Section XI Code VT-3 Examination not Completed Following Repair of Main Steam Isolation Valve due to the Omission of Relevant Information from a Plant Procedure URI 98-08-06:
Impact of Low 8lowout Panel Set Point on Secondary Containment IFl 97-08-02:
Verification of Exhausting Actual and Maximum Fan Capacities Listed in Procedure OP-2611, " Gaseous Radwaste" IFl 98-08-07:
Storm Drain Contamination and Evaluation per Bulletin 80-10 IFl 97-80-01:
Penetration Barriers and Seal Design Verification NON-CITED VIOLATIONS OPENED / CLOSED NCV 98-13-01:
Bypass of HPCI and RCIC Low Pressure Isolation Signals l
NCV 98-13-03:
Inadequate Design Control for Blowout Panel Mechanism ITEMS UPDATED l
VIO 97-531-05014: Residual Heat Removal Net Positive Suction Head (also reference eel l
97-10-01 b)
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VIO 97-531-07014: Design Calculations (also reference eel 97-10-01 h)
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l ITEMS WITHDRAWN VIO 98-80-01:
Failure to Perform Adequate Safety Evaluation VIO 98-80-03:
Failure to Issue Required Reports to the NRC L
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l LIST OF ACRONYMS USED A/E Architect / Engineer GMO Basis for Maintaining Operation
'BWR Boiling Water Reactor CAM Containment Air Monitor CFR Code of Federal Regulation CMIP Configuration Management improvement Project CS Core Spray DBD Design Basis Document DES-Duke Engineering & Services DI Department Instruction EDCR Engineering Design Change Request EDG Emergency Diesel Generator ER Event Report FSAR Final Safety Analysis Report GE General Electric GL Generic Letter
'GPM Gallons Per Minute HPCI High Pressure Coolant injection HVAC Heating, Ventilation and Air Conditioning IFl Inspection Follow-up ltem
- lST Inservice Testing LCO Limiting Condition for Operation LER Licensee Event Report LOCA Loss Of Cooling Accident
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LPCI Low Pressure Coolant Injection MSLB Main Steam Line Break NCV Non-Cited Violation NPSH Net Positive Suction Head NRC Nuclear Regulatory Commission ODCM Offsite Dose Calculation Manual
' PCIS Primary Containment isolation System PRA Probabilistic Risk Assessment QA Quality Assurance RB Reactor Building RHR Residual Heat Removal RMS Radiation Monitoring System RRU Room Refrigeration Unit SCE Safety Class Electrical SFPC Spent Fuel Pool Cooling-SGTS.
Standby Gas Treatment System SOV Solenoid-operated Valves SSPV Scram Solenoid Pilot Valve
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SW Service Water TS-Technical Specification L
URI Unresolved item VIO Violation VY Verrnont Yankee B-3
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ENGINEERING INSPECTION OPEN ITEM CROSS REFERENCE This matrix provides a cross reference between open items covered by the following NRC docketed activities: Engineering Team inspection (A/E Team) 50-271/97-201, Engineering Followup Inspection 50-271/97-10,and the Escalated Enforcement Action Letter 97-531, dated April 14,1998.
Inspection Report
50-271/97-201 Open item Title Replaced By Escalated Enforcement issue Closed By Open items (IR 97-201section reference)
IR 97-10 item Violation Number inspection Report URI 97-201-01 Suppression Pool Water Temperature - Past eel 97-10-01a VIO 97-531-01013 50-271/98-13 Operation Outside Design Basis (E1.1.2.2.a)
URI 97-201-02 Untimely Actions to Resolve Suppression Pool Water eel 97-10-03a VIO 97-531-01023 50-271/98-13 Temperature issue - 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action (E1.1.2.2.a)
IFl 97-201-03 Failure to issue an LER (E1.1.2.2.a)
eel 97-10-09a VIO 97-531-01033 50-271/98-12 URI 97-201-04 Clarification of RHR Pump NPSH Margins and eel 97-10-01b VIO 97-531-05014 open Correction of Calculation Errors (E1.1.2.2.b)
URI 97-201-05 Non-conservative LPCI Flow Valves Used in LOCA NCV 97-10-02 50-271/97-10 Analyses - 10 CFR Part 50, Appendix B, Criterion 111, Design Control (E1.1.2.2.c)
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URI 97-201-06 Insufficient technical basis as requested by Bulle'in eel 97-10-01c VIO 97-531-08014 50-271/98-13 88-04 for existing minimum flow, IEB 88-04 (E1.1.2.2.d)
URI 97-201-07 Change to LPCI Mode of RHR Operation as 50-271/98-11 described in FSAR (E1.1.2.2.d)
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Inspection Report
50-271/97-201 Open item Title Replaced By Escalated Enforcement issue Closed By Open items (IR 97-201section referenc.
1R 97-10 item Violation Number Inspection Report URI 97-201-08 Inappropriate operatmg instructions regarding RHR EEi 97-10-01d VIO 97-531-02014 50-271/98-12 pump motor starts,10 CFR Part 50, Appendix B, Criterion llL Design Control (E1.1.2.2.e)
URI 97-201-09 FSM1 Not Updated to incorporate Reduced RHR NCV 97-10-50-271/97-10 Heat Exchanger Capacity - 10 CFR 50.71(e)
12a (E1.1.2.2.f)
URI 97-201-10 Entry into TS LCO conditions when equipment is 50-271/97-10 rendered inoperable - TS 3.5.A.2 and TS 3.5.A.3 (E1.1.2.2.g)
URI 97-201-11 Timeliness to Follow-Up Self-Assessment of TS eel 97-10-03b 50-271/98-13 Surveillance Requirements (E1.1.2.2.g)
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URI 97-201-12 Instrument Accuracy DO T 71/98-11 IFI 97-201-13 RRU 7 & 8 - Calculation Methodology, incorrect eel 97-10-01e VIO 97-531-06014 50-271/98-13
,4ssumption,10 CFR 50, Appendix B, Criterion Ill,
'asign Control (E1.2.2.2.b)
Uiti 97-201-14 Downgrading RRUs 5 and 6 Without a Safety VIO 97-10-07 50-271/98-80 Evaluation,10CFR50.59, Changes, Tests, and Experiments (E1.2.2.2.c)
URI 97-201-15 Deleting RRus 5 and 6 from the GL 89-13 program, 50-271/97-10 Cor mitment in Response to Generic Letter GL 89-13 (E1.2.2.2.e)
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Inspection Report
50-271/97-201 Open item Title Replaced By Escalated Enforcement issue Closed By Open items (lR 97-201section reference)
lR 97-10 item Violation Number Inspection Report URI 97-201-16 RHR Heai Exc..2
- - Analysis of Tests VIO 97-10-06 50-271/98-13 Measurements Coi.e
- ted and Recorded with inaccurate / uncalibrated instruments,10CFR50, Appendix B, Criterion XII, Control of Measuring and Test Equipment (E1.2.2.2.f)
URI 97-201-17 GL 89-13 Commitment - Not inspecting RHR Heat 50-271/97-10 Exchanger 18, Commitment in Response to Generic Letter GL 89-13 (E1.2.2.2.g)
URI 97-201-18 Common Mode Failure of Non-Safety regulators eel 97-10-01f VIO 97-531-03014 50-271/98-12 Affecting Safety Related Diesel Generators, 10CFR50 Appendix B, Criterion lit, Design Control (E1.2.2.2.h)
URI 97-201-19 Effectively Changing a Technical Specification by a eel 97-10-03c 50-271/98-13 Revision to an Operating Procedure, NUREG 1606, Proposed Regulatory Guidance Related to implementation of 10CFR50.59 (E1.2.2.2.j)
URI 97-201-20 Use of PRA to address tornado missilea in lieu of eel 97-10-03d VIO 97-531-09014 50-271/98-13 providing positive protection as described in licensing documentatien. NRC acceptance of PRA as design and licensing basis may be necessary.
(E1.2.2.2.k)
IFl 97-201-21 Vernon 69ky switchyard low voltage - 10 CFR Part NCV 97-10-10 50-271/97-10 50, Appendix B, Criterion Ill, Design Control (E1.3.3.2.a)
URI 97-201-22 Main Station Battery Service Test-10CFR50 VIO 97-10-08 50-271/98-12 Appendix B, Criterion XVil, OA Records (E1.3.3.2.b)
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inspection Report 50-271/97-201 Open item Title Replaced By Escalated Enforcement issue Closed By Open items (IR 97-201section reference)
lR 97-10 item Violation Number inspection Report IFl 97-201-23 Standby Batte.y Charger CAB Single Failure, eel 97-10-03e 50-271/98-13 (E1.3.3.2.c)
IFl 97-201-24 Unapproved Use Offsite Vernon Tie as Station URI 97-10-14 50-271/98-13 Blackout AAC Power Source and as Offsite Delayed Access Source of Power (E1.3.3.2.e)
IFl 97-201-25 Lack of documentation within separation criteria 50-271/97-10
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regarding NEDO 10139 for analysis and documentation of separation exceptions (E1.3.3.2.b)
IFl 97-201-26 Lack of provisions for instrument drift in Instrument open Uncertainty Calculation Methodology (E1.3.3.3.d)
IFl 97-201-27 Excessively high Uncertainty for RHR Flow eel 97-10-01g VIO 97-531-04014 50-271/98-12 Indication and Recording Loop (E1.3.3.3.e)
eel 97-10-09b VIO 97-531-10014 50-271/98-12 URI 97-201-28 FSAR deficiencies and errors (E1.4.3)
NCV 97-10-50-271/97-10 12b URI 97-201-29 Design Control Weakness, (E1.5.3)
eel 97-10-01h VIO 97-531-07014 open C-4
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