IR 05000244/1998012
| ML17265A506 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 01/06/1999 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17265A505 | List: |
| References | |
| 50-244-98-12, GL-89-13, NUDOCS 9901190042 | |
| Download: ML17265A506 (33) | |
Text
U.S. NUCLEAR REGULATORY COMMISSION REGION I
License No.
Report No.
Docket No.
Licensee:
Facility Name:
Location:
DPR-1 8 50-244/98-1 2 50-244 Rochester Gas and Electric Corporation (RGSE)
R. E. Ginna Nuclear Power Plant 1503 Lake Road Ontario, New York 14519 Inspection Period:
Inspectors:
Approved by:
October 19, 1998 through November 29, 1998 P. D. Drysdale, Senior Resident Inspector C. C. Osterholtz, Resident Inspector K. S. Kolaczyk, Reactor Engineer D. A. Dempsey, Reactor Engineer L. J. Prividy, Sr. Reactor Engineer Lawrence T. Doerflein, Chief Projects Branch
Division of Reactor Projects 990ii90042 990i06 PDR ADDCK 05000244
.
~
~
~
EXECUTIVE SUMMARY R. E. Ginna Nuclear Power Plant NRC Inspection Report 50-244/98-12 This integrated inspection included aspects of licensee operations, engineering, maintenance, and plant support.
The report covers a 6-week period of resident inspection, and includes the results of announced inspections by regional specialists in the" inservice testing program and the service water system.
~Oerations Operations personnel performed well in response to an unexpected partial loss of offsite power.
The licensee effectively identified the cause of the loss, and restored power from the offsite circuit in a timely manner.
Operations personnel responded well to an inadvertent dilution when a mixed bed demineralizer with new resin was prematurely placed in service.
At the conclusion of the inspection period, the licensee was investigating the cause(s) of this event.
The requalification written examination for licensed operators was adequate.
Job performance measures reviewed had been recently enhanced and were of good quality.
Simulator scenarios were particularly effective in evaluating operator performance in that they exercised both the Functional Restoration and Emergency Contingency Action Procedures.
Operator performance during the requalification was good.
Maintenance Controlled maintenance procedures were used at job sites.
The procedures were up to date and were properly used by technicians involved in maintenance and surveillance work.
The inspectors observed good personnel and plant safety practices.
Equipment tested met the acceptance criteria specified for operability. The test acceptance criteria bases reviewed were adequate, with only minor discrepancies noted.
The licensee's efforts to troubleshoot and repair control room radiation monitors were initiallyunsuccessful and hindered by the lack of consideration of a previously discovered equipment non-conformance item and with difficulties in procuring spare parts.
The licensee's discovery of an improperly performed ITS surveillance reflected a good and detailed review of a previous outage work package.
Although both hydrogen recombiners were determined to be operable, the missed surveillance represented a condition prohibited by the ITS. This license identified and corrected violation of ITS surveillance requirements was of minimal safety consequence and is being treated as a Non-Cited Violation (NCV),
consistent with Section VII.B.1 of the NRC Enforcement Policy (NCV 50-244/98-12-02).
Executive Summary (cont'd)
~En ineerin The Inservice Test (IST) Program was well designed and implemented.
Program documents and implementing procedures were easy to use, technically correct, and updated to reflect industry developments.
Valves in the selected systems were tested in accordance with ASME Code requirements and NRC-approved relief requests.
Pre-conditioning of selected service water containment isolation valves prior to leakage testing may be inconsistent with industry guidance and is under review by the RGSE staff (IFI 50-244/98-12-03).
IST Program personnel were knowledgeable, and the condition monitoring program was effective.
Discrepancies regarding program scope and testing of relief valves were of minor importance and were appropriately addressed by RGSE.
The licensee established appropriate tests and maintenance tasks for service water system components consistent with the requested actions of NRC Generic Letter 89-13, "Service Water System Problems Affecting Safety-Related Equipment."
Technical items from a 1997 self-assessment of the service water system were appropriately prioritized and addressed in the licensee's corrective action program.
Radiological controls and postings in the auxiliary building were adequate.
Radiation Protection management tours were effective in the identification and resolution of radiological deficiencie TABLE OF CONTENTS EXECUTIVE SUMMARY TABLE OF CONTENTS IV I. Operations
02 Conduct of Operations........
~
~ ~....
~
.
01.1 General Comments...........
~..
~
........
Operational Status of Facilities and Equipment 02.1 Summary of Plant Status 02.2 Loss of Offsite Power on Circuit 751
~
~
Operator Knowledge and Performance......
05
04.1 Inadvertent Dilution of the Reactor Coolant System When Placing a Mixed Bed Demineralizer in Service
~.......3 Operator Training and Qualification..
~
~
~
~.
~
...............
05.1 Licensed Operator Requalification Program Inspection.......... 4 Miscellaneous Operations Issues..................
08.1 (Updated) LER 1998-003, Revision 1: Radon Build-up During Temperature Inversion Results in Actuations of Control Room Emergency Air Treatment System
~..
~
~ ~.......
II. Maintenance M1 Conduct of Maintenance............
~ ~....
M1.1 General Comments on Maintenance Activities
~
~ ~....
~ ~....
M1.2 General Comments on Surveillance Activities M2 Maintenance and Material Condition of Facilities and Equipment.. '...
M2.1 Corrective Maintenance for Control Room Radiation Monitor...
M4 Maintenance Staff Knowledge and Performance M4.1 Missed Technical Specification Surveillance on the Containment Hydrogen Recombiners....
...6
...6
... 6
~
~
I 7
..8
..8
. 9 III. Engineering.. ~... ~....... ~..........
~
~
~
~
~
~.
~
~..
E3 Engineering Procedures and Documentation E3.1 Inservice Testing Program Inspection
. ~...
E8 Miscellaneous Engineering Issues E8.1 (Closed) Unresolved Item 50-244/96-06-04:
Reliability Optimization Program.......
~
~
~
~
~
~
~
~
~
~
~
i ~
~
~
~
~
~ \\
~
~
~
~
~
~
~
~
~
~
~
~
~
~
6'ervice Water System
........
IV. Plant Support
.
~
R1 Radiological Protection and Chemistry (RP5C) Controls R1.1 Radiological Controls in the Auxiliary Building
'I
~
~
~
~
~
~
18,
18 V. Management Meetings...
~
~
~
~
~
~
~
~
~
~
~
~
~
~
~
~
~
~
~
~
X1 Exit Meeting, Summary.................
18 IV
Table of Contents (cont'd)
ATTACHMENTS Attachment 1 - Partial List of Persons Contacted
- Inspection Procedures Used
- Items Opened, Closed, and Discussed
- List of Acronyms Used
Re ort Details I. 0 erations
Conduct of Operations'1
~ 1 General Comments Ins ection Procedure IP 71707 The inspectors observed plant operations to verify that the facility was operated safely and in accordance with licensee procedures and regulatory requirements.
This review included tours of the accessible areas of the facility, verification of engineered safeguard features (ESF) system operability, verification of proper control room and shift staffing, verification that the plant was operated in conformance with the improved technical specifications (ITS) and appropriate action statements for out-of-service equipment were implemented, and verification that logs and records accurately identified equipment status or deficiencies.
Operator performance throughout the inspection period was good.
Operational Status of Facilities and Equipment 02.1 Summar of Plant Status The plant was at full power at the beginning of the inspection period.
On October 20 and October 27, 1998, control room radiation monitors reached their alarm setpoint and actuated the Control Room Emergency Air Treatment System (CREATS) and isolated the control room from outside air (see section 08.1).. The licensee issued a four-hour notification to the NRC in accordance with 10 CFR 50.72 for the October 27 actuation.
The actuation on October 20 was determined
.to be invalid and was.therefore not reportable.
On November 12, 1998, plant operators noted that the "Twinko" voltage regulator output for instrument bus A (MQ-400E) had increased to 120.0 VAC and was slowly rising. The limitfor operability contained in the bases of the ITS was 120.4 VAC. The licensee performed an engineering analysis and determined that the ITS limit could be raised to 124.4 VAC based on the required operating voltages for the A-instrument bus loads.
Also, the licensee set a new administrative alert limit of 121.4 VAC for the output voltage.
Instrumentation and Control.(ISC) technicians subsequently placed a chart recorder on the Twinko regulator output to monitor its performance, and operators logged its voltage reading every four hours.
The o'utput, voltage was holding steady at approximately 120.3 VAC at the end of the inspection period.
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 On November 17, 1998, an inadvertent dilution of the reactor coolant system (RCS)
occurred after operators placed a mixed bed demineralizer with new resin in service.
Control rods automatically inserted nine steps, and reactor power had to be momentarily reduced by approximately 0.7% (see section 04.1).
On November 20, 1998, a loss of offsite power on circuit 751 occurred, causing a
momentary loss of power to the B-train safeguards buses and an automatic start of the B-Emergency Diesel Generator (B-EDG). The licensee made a four-hour notification to the NRC, in accordance with 10 CFR 50.72 (see section 02.2).
On November 27, 1998, main control board annunciator J-27, "Generator Voltage Regulator Alarm," was received when system operators (load dispatchers)
at the Energy Operations Center in Rochester, NY placed an offsite capacitor bank in service.
Control room operators noted electric power perturbations at the Ginna Station, but the J-27 alarm was cleared and no plant transient occurred.
The licensee initiated ACTION Report 98-1647 to investigate the problem.
The plant continued to operate at full power at the end of the inspection period.
Loss of Offsite Power on Circuit 751 Ins ection Sco e (71707)
The inspectors reviewed the licensee's response to a loss of offsite power on circuit 751.
Observations and Findin s On November 20, 1998 at 10:59 p.m., offsite power circuit 751 was unexpectedly lost, which momentarily de-energized the B-train 480 VAC vital buses and caused an automatic start of the B-EDG, Operations personnel appropriately responded by entering abnormal procedure AP-ELEC.1, "Loss of 12A and/or 12B Buses."
All equipment operated normally while B-train vital loads were supplied by the B-EDG.
Fifteen minutes after the loss occurred, control room operators established a 100%
offsite power lineup on circuit 767, which re-established offsite power to the B-train vital buses.
The B-EDG was subsequently secured at 11:19 p.m., and the licensee made a four-hour notification to the NRC in accordance with 10 CFR 50.72.
At the time of the loss, security personnel noted smoke coming from an underground manway outside the plant's protected area in the West end of the site parking lot. Electricians subsequently discovered that a spliced cable in the manway for circuit 751 had failed due to insulation wear.
The licensee repaired and successfully tested the splice on November 21, 1998, and it was successfully returned to service on November 22, 1998. The circuit 751 cable had been satisfactorily tested two months prior to the failure; however, the licensee indicated that small insulation degradations can fail quickly due to the high voltage involved.
The licensee informed the inspectors that the entire cable, under the parking lot would be replaced, and was evaluating the appropriate time for that to occu Conclusions Operations personnel performed well in response to an unexpected loss of,offsite power on circuit 751. The licensee effectively identified the cause of the loss, and restored circuit 751 in a timely manner.
Operator Knowledge and Performance 04.1 Inadvertent Dilution of the Reactor Coolant S stem When Placin a Mixed Bed Demineralizer in Service Ins ection Sco e (71707)
The inspectors reviewed the licensee's response to an inadvertent dilution of the reactor coolant system (RCS).
b.
Observations and Findin s On November 17, 1998, the licensee was placing the A-mixed bed demineralizer in service following a resin replacement in accordance with procedure S3.2B, "Placing A Mixed Bed Demineralizer in Service - Boron Concentration Different than RCS."
Procedure S3.28 required that the demineralizer be flushed for thirty minutes with RCS water, and then sampled to determine if the boron concentration had equalized between the demineralizer and the RCS.
If the sample indicated that the boron concentrations had equalized, the demineralizer was to be placed in service for five minutes to verify equalization had occurred.
If the boron concentrations had not equalized, the demineralizer was to be flushed and re-sampled at thirty minute intervals until boron equalization was achieved.
Following the initial RCS flush, the demineralizer sample taken by Radiation Protection (RP) personnel indicated that the boron concentration was 259 parts per million (ppm)
~ The corresponding RCS boron concentration was 265 ppm.
Consequently, operators placed the A-demineralizer in service for five minutes, and
.then placed the demineializer in bypass.
Operators subsequently noted that the RCS average temperature increased by approximately 0.6 degrees Fahrenheit, and that control rods automatically stepped inward nine steps in response.
Operators reduced turbine load by approximately 0.7% to avoid exceeding 100% reactor power, as indicated on the nuclear instruments.
Operators performed a second thirty-minute flusli of the demineralizer and its subsequent sample indicated a boron concentration of only 10 ppm. This caused the licensee to believe that the demineralizer resin had removed an excessive amount of boron and caused an inadvertent RCS dilution of appi'oximately 200 gallons.
Consequently, operators added approximately six gallons of boric acid to the RCS to restore its normal boron concentration.
Operators then repeatedly flushed and sampled the demineralizer several additional times, and successfully placed it in service after a normal boron concentration was achieved.
The immediate causes of the event were not readily apparent; however, the licensee
initiated ACTION Report 98-1617 to investigate the event, and to identify and correct the related causes.
The licensee also indicated that a human performance evaluation would be conducted to identify and resolve possible human performance issues.
C.
Conclusions Operations personnel responded well to an inadvertent dilution of the reactor coolant system when a mixed bed demineralizer with new resin was prematurely placed in service.
At the conclusion of the inspection period, the licensee was investigating the cause(s)
of this event. An inspector follow-up item was initiated to review the completed event investigation and human performance evaluation (IFI 50-244/98-12-01).
Operator Training and Qualification 05.1 Licensed 0 erator Re uglification Pro ram Ins ection a.
Ins ection Sco e (71707)
The inspectors reviewed and evaluated the licensee's annual requalification examination for licensed operators.
b.
Observations and Findin s The inspectors reviewed the written requalification examination that the licensee administered on November 12, 1998,.and determined that the test items were in accordance with the standards for the development of open book multiple choice questions prescribed in NUREG-1021, "Operator Licensing Examiner Standards."
The inspectors noted'that several test items could have been enhanced to provide a better overall assessment of operator knowledge, and discussed these items'with the licensee, who indicated that written examination test items would be reviewed for improvement.
The inspectors also reviewed the Job Performance Measures (JPMs) that the licensee administered on November 11, 1998. The tasks reviewed were of sufficient depth to identify potential operator performance weaknesses in procedural usage, and to test operator knowledge of the location of safety-related plant equipment.
The inspectors also noted that the licensee recently updated JPMs to enhance the cues for evaluators who administer the JPMs.
The inspectors observed an operating test that the licensee administered in the training simulator on November 10, 1998. The inspectors considered that the scenario set administered was challenging, and that it went into significant depth in the Emergency Operating, Functional Restoration, and Emergency Contingency Action Procedures.
The observed crew successfully accomplished all identified critical tasks.
Supervisory oversight of the crew was particularly noteworthy, as crew briefings were thorough and included appropriate guidance at operational
transition points.
The inspectors noted some minor modeling deficiencies between the'"control room and the simulator that were discussed with the licensee for resolution.
Some examples included 1) the maintenance rule equipment status posted in the simulator was of a different revision than that posted in the control room, 2) the Operations Plan in the simulator was not current, and 3) the Operator Aid tags posted in the simulator were not consistent with those in the control room.
C.
Conclusions The requalification written examination for licensed operators was adequately prepared.
Job performance measures reviewed had been recently enhanced and were of good quality. The observed simulator scenarios were particularly effective in evaluating operator performance, in that they exercised both the Functional Restoration and Emergency Contingency Action Procedures.
Observed operator performance during the requalification was good.
IVliscellaneous Operations Issues 08.1 U dated LER 1998-003 Revision 1: Radon Build-u Durin Tem erature Inversion Results in Actuations of Control Room Emer enc AirTreatment S stem On October 5, 1998, the licensee submitted LER 1998-003 to the NRC following three successive actuations of the Control Room Emergency AirTreatment System (CREATS) causing the control room ventilation system to isolate from the outside air and to reconfigure to its recirculation mode (see IR 50-244/98-11).
The licensee submitted Revision 1 to the LER on November 24, 1998, after two additional actuations signals were generated on October 20 and 27, 1998.,
Revision 1 stated that the October 20 actuation was an invalid signal due to a possible faulty detector and/or cable, and that the October 27 actuation was due to a temperature inversion.
The inspectors reviewed the data for the air samples the licensee took following each actuation, and agreed that the results confirmed the licensee's conclusions.
The inspectors also observed the licensee's troubleshooting and maintenance activities to correct problems associated with the control room radiation monitors.
These activities were still ongoing at the end of the inspection-period.(see section M2.1). The inspectors concluded that the LER adequately described the details of each actuation; however, the licensee indicated that the need for another revision to the LER would be determined after additional analysis and maintenance was performed.
Pending completion and NRC review of the root causes and corrective actions, this LER remains open (LER 1998-003, Revision 1).
II. Maintenance M1 Conduct of Maintenance M1.1 General Comments on Maintenance Activities a.
Ins ection Sco e (62707)
The inspectors observed portions of plant maintenance activities to verify that the correct parts and tools were utilized; the applicable industry codes 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.
Component failures or problems that affected systems included in the RGSE Maintenance Rule Program were assessed to determine if the maintenance was effective.
b.
Observations and Findin s The inspectors observed all or portions of the following maintenance work activities:
WO¹ 19800485, "Calibration of RHR Outlet Valve HCV-624" (observed on November 11, 1998)
WO¹ 19802567, "RHR Discharge MOV to Sl Pump - 857B-Replace 7.5 ft/Ib Motor with 10 ft/Ib Motor" (observed on November 11, 1998). An electrician performing the work identified that the replacement motor had a large dent on the motor casing.
The licensee obtained another replacement motor, and initiated an ACTION Report (98-1593) to investigate the damage on the first motor.
~
WO¹ 19803454, Inspect B-RHR Pump Circuit Breaker in accordance with Procedure "GME-50-02-DBINSPECT" (observed on November 12, 1998).
~
Station Modification SM95-098.1, spent fuel pool re-rack.
The modification also accomplished a reconfiguration of the spent fuel pool cooling system suction strainer because it interfered with full access to one of the new spent fuel storage racks (observed on November 13, 1998).
c.
Conclusions The inspectors observed that controlled procedures in use at the job sites were up to date and were properly utilized by technicians involved in the work. The inspectors observed good personnel and plant safety practices during the maintenance wor M1.2 General Comments on Surveillance Activities a.
Ins ection Sco e (61726)
The inspectors observed selected surveillance tests to verify that approved procedures were in use, procedure details were adequate, test instrumentation was properly. calibrated and used, technical specifications were satisfied, testing was performed by knowledgeable personnel, and test results satisfied acceptance criteria or were properly dispositioned.
The inspectors also reviewed selected acceptance criteria bases (ACBs) for consistency with the ITS, the UFSAR, and related design basis documents.
b.
Observations and Findin s The"inspectors observed portions of the following surveillance activities:
~
PT-9.1.17, "Undervoltage Protection - 480 Volt Safeguards Bus 17,"
(observed November 16, 1998).
~
PT-9.1.18, "Undervoltage Protection - 480 Volt Safeguards Bus 18,"
(observed November 16, 1998)
~
ACB 97-0071:
ACB 97-0072:
ACB 97-0073:
Degraded Voltage for "27" Relays Dropout Time for "27" Relays Loss of Voltage for "27D" Relays All undervoltage relays on buses 17 and 18 performed within the acceptance criteria limits designated by the PT. The ACBs listed above were consistent with the current engineering design analyses for relay setpoints, and ITS limits on degraded voltage, relay drop out times, etc.
The inspectors noted minor administrative and typographical inconsistencies between the ACBs and their referenced documents, eg, ACB 97-0073 stated what the ITS specification was for the loss of voltage setpoint on the "27D" relays, but it also indicated that the ITS bases were not applicable.
~
PT-2.3, "Safeguard Motor Operated Valve Operation," for MOV-704B, B-RHR pump suction valve; and MOV-857B, RHR pump discharge to Sl pump Suction (observed on November 12, 1998).
ACB 94-0245:
ACB 94-0220:
Maximum stroke time for MOV-704B Maximum stroke time for MOV-857B Both valves achieved stroke times within the acceptance criteria specified in the licensee's inservice test program (ASME Section XI) document, and both of the ACBs were consistent with the supporting ASME requirements for maximum allowable stroke times for these valves.
However, both ACBs referenced an out-of-date document (ME 280) that was superseded by interface procedure IP-IIT-2,
"Inservice Testing Program for Pumps and Valves."
The inspectors discussed the inconsistencies in all of the above ACBs with the licensee's inservice test (IST) coordinator, who acknowledged the need to update the ACBs with current information, and to include other related references which establish the design basis functional requirements.
This area was already identified as a necessary effort in order for the licensee to complete its 10 CFR 50.54 (f)
project (see IR 50-244/98-08).
The licensee indicated that the ACBs used for technical specification surveillances and IST program equipment will be upgraded to controlled documents to assure that acceptance criteria limits remain current with program requirements.
Conclusions The inspectors confirmed that procedures used during surveillance tests were current and properly followed. The equipment test results satisfied the acceptance criteria specified in the procedures for operability.
The acceptance criteria bases reviewed were adequate, with only minor discrepancies noted.
Maintenance and Material Condition of Facilities and Equipment Corrective Maintenance for Control Room Radiation Monitor Ins ection Sco e (62703)
The inspectors reviewed maintenance activities performed to correct problems with the control room radiation monitors.
Observations and Findin s Maintenance activities to troubleshoot and repair control room radiation monitors were ongoing throughout the inspection period.
This effort was initiated after radiation monitors R-36 (control room noble gases)
and R-37 (control room particulates) generated actuation signals, causing the control room emergency air treatment system (CREATS) to isolate the control room from outside air. A total of six actuation signals occurred from September 4, 1998, through October 27, 1998.
Four were attributed to increased radon levels caused by temperature inversions and two were considered invalid actuation signals caused by instrument spiking. The four valid actuations were considered reportable in accordance with the requirements of 10 CFR 50.72 (see section 08.1).
The licensee left control room ventilation in the recirculation mode after the actuation signal that occurred on October 20, 1998.
During initial troubleshooting activities, Instrumentation and Control (IRC)
technicians, diagnosed that a short in the detector and a short in the detector cable for R-37 contributed to inadvertent CREATS actuation signals.
The licensee ordered a new cable be manufactu'red in accordance with the vendor manual wiring diagrams, and procured a new detector.
The procured detector cable was discovered to have an incorrect placement of its alignment key for matching pins to sockets and could not be used as a replacement.
The licensee generated ACTION
Report 98-1635 to investigate the problem.
A second procured detector had the alignment key in the correct position, but did not have the associated paperwork to demonstrate that it was suitable for safety-related use.
The radiation monitor system engineer subsequently discovered a non-conformance report from 1991 (91-520) that indicated the pin configuration for the R-37 detector was different than that specified in the vendor manual drawings (i.e., pin "H" and pin "F" should be shown as switched).
The switched pin configuration had not been incorporated into the vendor manual drawings, and thus adversely affected the initial troubleshooting activities.
The licensee re-evaluated the original R-37 detector's operability using the correct configuration information and discovered that the detector was not shorted and still acceptable for use.
Also, since the new cable for R-37 was ordered using the wrong configuration, it could not be used as a replacement.
The licensee generated ACTION Report 98-1643 to investigate the issue.
The licensee performed an operational analysis to raise the alarm setpoints for the control room radiation monitors to prevent spurious actuation signals being generated due to temperature inversions.
The minimum analyzed air flow through the detectors for the new setpoints was determined to be 3.0 standard cubic feet per minute (scfm).
However, only 2.7 scfm could be measured using the installed rotameter on the radiation monitor skid. The licensee generated ACTION Report 98-1646to investigate the problem.
IRC technicians measured air flow using test equipment with increased sensitivity, and determined that the actual flow was approximately 11 scfm and that the rotameter was inaccurate.
The licensee indicated that the rotameter would have to be disassembled, cleaned, and inspected to further investigate the problem.
The control room ventilation system remained in the recirculation mode at the end of the inspection period, as corrective maintenance activities were still being performed.
Conclusions The licensee's efforts to troubleshoot and repair control room radiation monitors were initiallyunsuccessful and hindered by the lack of consideration of a previously discovered equipment non-conformance item and with difficulties in procuring spare parts.
Maintenance Staff Knowledge and Performance Missed Technical S ecification Surveillance on the Containment H dro en Recombine rs Ins ection Sco e (62703)
The inspectors reviewed the licensee's actions in response to their discovery that ITS surveillances on both containment hydrogen recombiners were not properly performed during the last refueling outag Observations and Findin s On November 17, 1998, the licensee identified that the calibration of flow transmitters FT-3-1A and FT-3-1B on the containment hydrogen recombiners, performed during the last refueling outage (October 1997), were not accomplished in accordance with procedural requirements.
The calibrations were performed to satisfy ITS surveillance requirement SR 3.6.7.2, which requires a channel calibration of the hydrogen recombiner actuation and control circuits at least once every 24 months.
Flow transmitters FT-3-1A and FT-3-1B sense the flow of combustion air when the recombiner blower motor is operating, and enable the recombiners'ontrol circuits to initiate combustion. Calibration procedures CPI-CNMT-INSTR-398Aand CPI-CNMT-INSTR-398B both required that a 0-2 inch incline manometer, with an accuracy of a0.05 inches of water, be used to calibrate the flow switches.
However,'the licensee identified that a Heise gage, with an accuracy of a0.4 inches of water, was actually used for both flow transmitter calibrations.
Based on this discovery, the licensee initiallydetermined that the operability of both hydrogen recombiners could not be assured, and entered limiting condition for operation (LCO) 3.6.7, Condition B, for two inoperable hydrogen recombiners.
LCO 3.6.7, Condition B, allows seven days to restore one recombiner to operability.
However, upon further review of the calibration data, the licensee concluded that both recombiners were operable, because no adjustments were made to their October 1997 calibration "as-found" data, and because these flow transmitters had historically not exhibited a drift problem.
Consequently, the licensee determined that Surviellance Requirement (SR) 3.6.7.2 was improperly performed since test equipment with an inappropriate accuracy was used.
The licensee also determined that SR 3.6.7.2 should be declared "missed" under the provisions of LCO SR 3.0.3, which allowed deferral of an operability determination, but required the surveillance be re-performed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of discovery.
The following day, IKC technicians entered the containment, re-performed the calibrations on both flow transmitters, and verified that the flow transmitters functioned within their required tolerances using test equipment with the proper accuracy.
The licensee then declared both hydrogen recombiners operable and properly exited LCO SR 3.0.3 within the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> limit. Subsequent to the end of the inspection period, the plant operations review committee (PORC) concluded that the missed surveillance was reportable, in accordance with 10 CFR 50.73.
Conclusions The licensee's discovery of an improperly performed ITS surveillance reflected a good and detailed review of a previous outage work package.
Although both hydrogen recombiners were determined to be operable, the missed surveillance represented a condition prohibited by the ITS. This license identified and corrected violation of ITS surveillance requirements was of minimal safety consequence and is being treated as a Non-Cited Violation (NCV), consistent with Section VII.B.1 of the NRC Enforcement Policy (NCV 50-244/98-12-02).
III. En ineerin E3 Engineering Procedures and Documentation E3.1 Inservice Testin Pro ram lns ection (73756)
a.
Ins ection Sco e
Inspectors assessed the adequacy of the Ginna Inservice Test (IST) Program via review of the IST Program document, its implementing procedures, and IST test results.
The inspectors witnessed IST test activities conducted by RG5E and performed walkdowns of safety-related equipment.
Test equipment was examined during IST tests of pumps to verify that ASME Boiler and Pressure Vessel Code (Code) range and accuracy requirements were also satisfied.
The IST program document and procedure reviews included a verification that components were tested in accordance with the Code, and the conditions outlined in the NRC safety evaluation report (SER) that approved the Ginna IST Program.
The inspectors placed a particular focus on Code relief requests, cold shutdown justifications, and surveillance procedures for components in three systems identified as higher contributors to overall risk in the Ginna probabilistic risk assessment (PRA). The three systems of concern were the component cooling water (CCW), service water (SW) and residual heat removal (RHR) systems.
The inspectors also determined if RGRE had monitored industry developments in the IST area, and reviewed how RG&E dispositioned the following NRC Information Notices (INs) that discussed potential IST program weaknesses:
IN 89-32, "Surveillance Testing of Low-Temperature Overpre'ssure Protection Systems" IN 94-30, "Leaking Shutdown Cooling Isolation Valves at Cooper Nuclear Station" IN 97-09, "Inadequate Main Steam Safety Valve Setpoints and Performance Issues Associated With Long MSSV Inlet Piping" IN 97-90, "Use of Non-conservative Acceptance Criteria in Safety Related Pump Surveillance Tests" b.
Observations and Findin s Pro ram Sco e
During this inspection, the Ginna IST Program was in its third ten-year interval, which started on January 1, 1990, and was due to expire on December 31, 1999.
Testing requirements were based upon the 1986 edition of the ASME Code including subsections IWP, "Inservice Testing of Pumps in Nuclear Power Plants" and IWV "Inservice Testing of Valves in Nuclear Power Plants."
Using RGSE's IST Program submittals, the Updated Final Safety Analysis Report (UFSAR), improved technical specifications (ITS), design basis documents, system drawings, and operating and surveillance procedures, the inspectors verified that the pumps and valves that performed a safety, function in the CCW, SW, and RHR systems were included in the IST program.
During the review, the inspectors identified one program scope issue regarding the SW system isolation valves on the spent fuel pool (SFP) heat exchangers.
Section 9.2.1.4 of the UFSAR stated that emergency operating procedures provided guidance to operators for the isolation of the heat exchangers, if necessary, to ensure that adequate SW flow was provided to the CCW heat exchangers during the recirculation phase of an accident.
Under some post-accident conditions, procedure ES-1.3, "Transfer to Cold Leg Recirculation," directed closure of SFP heat exchanger service water discharge manual isolation valves 4622 and 8689 to obtain total service water flow of greater than 5000 gallons per minute to the CCW heat exchangers.
However, these valves were not included in the IST Program.
As a result, the valves may not have received the applicable tests, which would have included periodic cycling and position indication verification.
In a November 19, 1998 telephone conversation, RGRE resolved the testing concern on these valves by placing the valves within the IST Program.
The failure to include the valves in the program and conduct the required testing as described in Articles IWV-1100 and IWV-3410 of the Code constituted a violation of minor significance and was not subjected to formal enforcement action.
Pum Testin The inspectors found that RG5E had properly identified and tested the CCW, SW, RHR, and containment spray (CS) pumps.
Measured IST parameters and associated reference values were placed into a computerized database that trended test results automatically.
Component condition summary reports and ACTION Reports had been initiated by the condition monitoring and performance trending group whenever a degrading trend was noted.
For example, ACTION Report 97-1448 documented a declining trend in the C-standby auxiliary feedwater pump differential pressure while still well above its lower alert range limit. RGRE's engineering evaluation found that the trend was a cyclical phenomenon and attributed the decrease to seasonal changes in storage tank water temperature changes.
The licensee's condition monitoring and performance trending program also included vibration spectrum analysis, lubricating oil sampling and analysis, and routine bearing temperature monitoring.
Valve Testin Component test frequencies, methods, and acceptance criteria for valves in selected systems were in accordance with Code requirements, except for three examples discussed in greater detail in the following paragraphs.
Containment isolation valves were included in the IST Program as Category A or A/C valves and leakage
rates were tested pursuant to 10 CFR 50, Appendix J, as required by relief request GR-2 and the March 1991 NRC Safety Evaluation Report (SER) for the Ginna IST Program.
The licensee's IST procedures were adequate.
For example, procedure PT-8.0,
"RHR System Valves - Seat Leakage Test Valves 852A/B, 853A/B, 700, 701, 720, and 721," which established the method for testing reactor coolant system pressure isolation valves in the RHR system per the Ginna ITS, satisfied the requirements of the ITS Program and the Code, by correcting the results for tests performed at less than nominal function differential pressure.
The valve test program was thorough and acceptable; however, the inspectors identified three issues concerning valve testing as follows:
Relief request GR-3, was applicable to instrument air system containment
isolation check valve 5393.
The request was changed incorrectly (and contrary to the March 1991 SER) to defer the exercise test required by Article IWV-3522, in accordance with the five-year schedule of 10 CFR 50, Appendix J, Option B. However, the error was administrative only, and did not carry-over into the test schedule for the valve, which continued to be tested at the proper refueling outage frequency.
RGRE agreed to revise the relief request accordingly.
2)
Cold shutdown justification CS-10 stated that power-operated relief valves 430 and 431C would be tested quarterly when in the cold shutdown condition, and when cooling down from power operations when it has been greater than 92 days, plus 25% since the last test.
The inspectors noted that in accordance with Section 6.2 of NUREG-1482, "Guidelines for Inservice Testing at Nuclear Power Plants," the 25% extension was not intended to be used repeatedly and as an operational convenience to extend surveillance intervals beyond those specified.
RGRE agreed to remove reference to the 25% exterision from CS-10.
3)
RGKE seat leak tested four six-inch, manually-operated containment isolation valves that were locked open during plant operation.
The valves of concern are located in the service water system return lines from the four containment recirculation fan cooler (CRFC) units.
The valves can provide operators the capability to isolate a leaking cooler if the containment becomes pressurized above service water system pressure during an event, and thereby prevent the escape of radioactivity to the environment through the service water system.
The valves were classified as containment isolation valves, and were seat leak tested, but not in accordance with the criteria outlined in Appendix J of 10 CFR 50, based on the lack of an automatic isolation function and the closed loop nature of SW inside containment.
For example, the seat leak test medium was water, in lieu of air, and the measured leak rate was not applied to the overall containment leak rate.
This departure from the
Appendix J test requirements was approved by the NRC in December 1982, as part of the Ginna Systematic Evaluation Program (SEP), after which seat leak testing began.
The inspectors'oncern involved how the seat leak test was conducted.
Specifically, Ginna seat leak test procedure RSSP-2.4, "CNMT Recirculation Fan Service Water Valve Leak Check" allowed operators to cycle the isolation valves up to five times before commencing a seat leak test, to ensure positive seating characteristics.
However, as described in NRC Information Notice (IN) 97-16, "Preconditioning of Plant Structures, Systems, and Components before ASME Code Inservice Testing or Technical Specification Surveillance Testing," the NRC has determined valve cycling before testing to be an unacceptable preconditioning activity, since it prevents assessment of how a valve would operate during a plant event.
RGRE personnel indicated the valves were cycled prior to testing because, they would likely fail the as-found test due to corrosion build-up on the valve seats.
However, the inspectors noted system operating procedures made no provision of the need to cycle the valves to ensure positive seating du'ring an abnormal event.
To resolve this issue, RGRE committed to review the valve seat leak test, operating procedures, and the valve design/licensing basis prior to start-up from the next refueling outage, and to modify them, as necessary, to ensure appropriate valve testing.
This issue will be an inspection follow-up item (IFI 50-244/98-12-03).
Power-0 crated Valves The inspectors considered that the licensee's testing of power-operated valves was in accordance with the Code.
For example, relief request GR-7 permitted RG5E to establish stroke time limits based on reference values, in addition to limiting values of full stroke time.
RGSE had established appropriate reference values based on the average of three strokes measured when the valves were in good working order.
Stroke time measurements were taken to the nearest tenth of a second, which exceeded the Code requirement of IWV-3413(b). Fail position tests of air-operated valves were performed properly, and remote valve position indicator verification was performed in accordance with IWV-3300 and Section 4.2.5 of NUREG-1482 (positive verification of obturator position).
The inspectors noted that RGSE had considered design information contained in the UFSAR when developing valve stroke times and had dispositioned inconsistencies when appropriate.
For example, the limiting values of full stroke time established for motor-operated containment spray isolation valves 860A/8/C/D exceeded the opening stroke time assumed in the containment integrity steam line break and loss of coolant accident recirculation analyses.
However, this condition had been evaluated in design analysis DA-ME-95-160, "Stroke Time Limit Evaluation for Containment Spray Discharge Valves," dated December 5, 1995. The analysis demonstrated that sufficient flow was provided through the valves at the stroke time assumed in the accident analyses, assuming that the valves were stroking at
'he limiting rate established in the IST program.
The inspectors found RGRE's evaluation to be acceptable.
A'SME Class 2 and 3 Relief Valves The IST program invoked OM-1-1987 for testing safety/relief valves.
Sub-tier maintenance procedures for each type of relief valve were clearly written with appropriate quality control hold points.
The sections pertaining to visual inspection, seat tightness testing, valve sample inspection, consecutive openings and minimum waiting period, and acceptance criteria met the requirements of OM-1. However, the provisions of OM-1 regarding temperature stability, ambient and process fluid temperatures were not addressed.
Specifically, sections 8.1.3.1, 8.1.3.4, and 8.1.3.5 of OM-1 required relief valves to be tested with the normal system operating fluid and temperature for which they were designed.
To meet this requirement, the temperature of the valve bodies must be known and stabilized before commencing set pressure testing, with no change in measured temperature of more than 10 degrees Fahrenheit in 30 minutes.
Further, the ambient temperature of the operating environment must be simulated during the set pressure test.
If the effect of ambient temperature on the set pressure was known for a particular valve type (viz. established by documented tests) an ambient temperature different from the operating ambient temperature may be used.
RGSE's procedures did not require documentation of ambient or valve body temperatures or verification of ambient temperature stability during testing.
Where applicable, "cold" set pressure settings of valves in high temperature service were specified by the valve vendor.
However, RGRE did not have information to verify that the correlations were certified by test stated in OM-1. The ASME has found
,that some relief valve. manufacturers have no engineering or test bases for their correlations, and has established a task force to determine standardized criteria for the correlations.
Since the difference between the "cold" and operating set pressures typically are small, the inspectors concluded that no immediate safety concern existed regarding RGSE's current practice.
Failure to address the temperature requirements of OM-1 was a violation of minor significance and is not.
subject to formal enforcement action.
The inspectors reviewed the test results of 16 relief valves in various safety-related systems.
Several service water system valves failed primarily due to internal corrosion or excessive seat leakage.
RGRE had implemented an adequate corrective action plan for these valves that consisted of testing at an increased frequency and replacement with corrosion resistant valves.
Test E ui ment During an IST Program self-assessment conducted in March 1997, RGSE identified several installed gauges and instruments that did not meet the range or accuracy requirements of ASME Article IWP-4120. As a temporary measure, test procedures were revised to install suitable instruments prior to each test.
The inspectors noted
that RGS.E was installing (or had installed) permanent test connections to allow the use of hand held digital Heise gauges having an accuracy of 0.25% of full scale.
While observing an IST test of a standby auxiliary feedwater pump, the inspectors noted that the licensee had provisions in place to ensure that the difference in elevation between the test connection and the Heise gage was properly accounted for.
Use of Industr 0 eratin Ex erience RGSE was aware of industry developments in the IST area described in NRC INs and industry IST workshops, and had modified the IST Program where appropiiate to ensure it met industry standards and NRC requirements.
Recent program changes included installing the portable digital Heise pressure gages discussed in the previous paragraph and changing test procedures to more clearly identify when partial flow testing of check valves was required, C.
Conclusions The Inservice Test (IST) Program was well designed and implemented.
Program documents and implementing procedures were easy to use, technically correct, and updated to reflect industry developments.
Valves in the selected systems were tested in accordance with ASME Code requirements and NRC-approved relief requests.
Pre-conditioning of selected service water containment isolation valves prior to leakage testing may be inconsistent with industry guidance and is under review by the RG5E staff (IFI 50-244/98-12-03).
IST Program personnel were knowledgeable, and the condition monitoring program was effective.
Discrepancies regarding program scope and testing of relief valves were of minor importance and were appropriately addressed by RGSE.
E8 Miscellaneous Engineering Issues E8.1 Closed Unresolved Item 50-244 96-06-04: Service Water S stem Reliabilit 0 timization Pro ram The inspectors reviewed the licensee's corrective actions pertaining to the resolution of several issues associated with the service water (SW) system.
The inspectors independently evaluated the following items:
Revision of the Service Water System Reliability Optimization Program (SWSROP) to reflect the current system configuration; Completion of the initial thermal performance test program for the safety related heat exchangers cooled by the SW system and establishment of test and/or cleaning frequencies; Confirmation that the action items from a 1997 SW system self assessment, such as updating the system hydraulic model, were adequately addresse The inspectors verified the existence of periodic testing and maintenance requirements, included the basis for all safety related heat exchangers cooled by the SW system.
The inspectors found that the frequencies for such testing and maintenance were based on sound engineering evaluations.
An updated and simplified SWSROP document was also developed which identified specific test and maintenance procedures to implement program requirements.
The inspectors noted that the licensee had completed about two thirds of 'the original 100 open items in the Commitment and Action Tracking System (CATS)
that were attributed to the 1997 SW system self assessment.
These items were appropriately prioritized and addressed.
For example, the inspectors found that a good plan was in place to address Action Report 97-2149, issued in December 1997, which involved the observation that the A-CCW heat exchanger was exhibiting higher tube-side fouling than the B-CCW heat exchanger.
The plan included a flush prior to a thermal performance re-test of the heat exchangers in the upcoming March 1999 outage.
A number of items that affected the SW hydraulic model, such as the need to clarify the design basis screen house bay levels used in design analysis, had been adequately addressed.
Items still open were of minor significance and were being tracked in the CATS. The inspectors considered these actions to have been adequate.
The licensee's commitment to improved SW system performance for the long term was evident by a number of initiatives:
Effective chlorination, inspection, and maintenance of the SW system in
, controlling zebra mussel activity
~ The completion in 1997 and 1998 of certified pump performance tests with new stainless steel impellers 'at the vendor's test facility
~
Completion in August 1998 of a system hydraulic model developed using conservative conditions to appropriately account for pump perfo'rmance, and macro-scopic fouling or tube plugging at the heat exchangers
'I
~
Planned modifications at the 1999 refueling outage include re-tubing both CCW heat exchangers and addition of instrument taps to improve online performance monitoring of various heat exchangers The inspectors determined that the outstanding issues concerning Unresolved Item 50-244/96-06-04had been adequately resolved.
No violations of regulatory requirements were observed.
The licensee established appropriate tests and maintenance tasks for system components consistent with the requested actions of Generic Letter 89-13, "Service Water System Problems Affecting Safety-Related Equipment." This unresolved item is closed. (URI 50-244/96-06-04).
IV. Plant Su ort R1 Radiological Protection and Chemistry (RP&C) Controls R1.1 Radiolo ical Controls in the Auxiliary Buildin a.
Ins ection Sco e (71750)
The inspectors performed plant tours of the auxiliary building to verify proper radiological controls.
b.
Observations and Findin s The inspectors performed tours periodically throughout the inspection period.
Contaminated areas, radiation areas, and locked high radiation areas were clearly identified and properly posted.
RP staff monitored personnel entering and exiting the auxiliary building, including discussions on radiation work permit requirements.
The inspectors noted that documented periodic RP management tours resulted in the identification and resolution of a few deficiencies in radiological controls, such as the repair of worn contamination boundary tape.
Management tours also resulted in the periodic verification that radiation monitoring equipment was properly source checked and calibrated.
C.
Conclusions Radiological controls and postings in the auxiliary building were adequate.
Radiation Protection management tours were effective in the identification and resolution of radiological deficiencies.
V. Mana ement IVleetin s X1 Exit Meeting Summary The inspectors presente'd the inservice testing inspection results to members of licensee management at the conclusion of the inspection on October 30, 1998.
The inspectors presented the results of the service water follow-up inspection on November 20, 1998. At the end of the inspection period, the inspectors presented the overall results to members of licensee management on December 3, 1998. The licensee acknowledged the findings presented.
The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary.
No proprietary information was identifie ATTACHMENTI PARTIALLIST OF PERSONS CONTACTED
.
Licensee B. Flynn J. Pascher
.G. Graus J. Hotchkiss G. Joss
'R. Popp R. Ploof J. Smith B. Thomson J. Widay T. White G.,Wrobel Primary Systems Engineering Manager Electrical Systems Engineering Manager ISC/Electrical Maintenance Manager Mechanical Maintenance Manager Results and Test Supervisor Production Superintendent Secondary Systems Engineering Manager Maintenance Superintendent Chemistry & Radiological Protection Manager Plant Manager Operations Manager Nuclear Safety 5 Licensing Manager INSPECTION PROCEDURES USED IP 37551:
IP 61726:
IP 62703:
IP 62707:
IP 71707:
IP 71750:
IP 73756:
IP 92901:
Onsite Engineering Surveillance Observation Maintenance Observation Maintenance Observation Plant Operations Plant Support Inservice Testing of Pumps and Valves Follow-up - Operations ITEMS OPENED, CLOSED, AND DISCUSSED
~Oen ed IFI 50-244/98-12-01 Inadvertent RCS Dilution NCV 50-244/98-12-02 Improperly Performed ITS Surveillance Test IF I 50-244/98-1 2-03 Inspector follow-up of licensee's actions to address pre-conditioning of valves Closed URI 50-244/96-06-04 Service Water System Reliability Optimization Program NCV 50-244/98-12-02 Improperly Performed ITS Surveillance Test
Attachment
Discussed LER 1998-003, Rev.
Radon Buildup During Temperature Inversion Results in Actuations of Control Room Emergency Air Treatment System
Attachment
LIST OF ACRONYMS USED SWSRO
.
VAC ACB ASME CCW CW CFR CREATS CRFC EDG ESF IFI IN IR ISI IST ITS JPM LCO LER MOV MDAFW MSSV NCV NRC ppfn PRA PT RCS RFC RGRE RHR RP RPRC scfm SER SFP SI SW Acceptance Criteria Basis American Society of Mechanical Engineers Component Cooling Water Circulating Water Code of Federal Regulations Control Room Emergency Air Treatment System Containment Recirculation Fan Coolers Emergency Diesel Generator Engineered Safeguards Feature Inspector Follow-up Item Information Notice
Inspection Report
Inservice Inspection
Inservice Test
Improved Technical Specification
Job Performance
Measure
Limiting Condition for Operation
Licensee Event Report
Motor-Operated Valve
Motor-Driven Auxiliary Feedwater
Noncited Violation
Nuclear Regulatory Commission
parts per million
Probablistic Risk Assessment
Periodic Test
Recirculation Fan Cooler
Rochester Gas and Electric Corporation
Radiation Protection
Radiological Protection and Chemistry
standard cubic feet per minute
Safety Evaluation Report
Spent Fuel Pool
Safety Injection
Service Water System Reliability Optimization Program
Updated Final Safety Analysis Report
Unresolved Item
Volts Alternating Current