IR 05000293/1998008
| ML20196C127 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 12/01/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20196C125 | List: |
| References | |
| 50-293-98-08, 50-293-98-8, NUDOCS 9812010252 | |
| Download: ML20196C127 (22) | |
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Enclosure U.S. NUCLEAR REGULATORY COMMISSION l REGION I .l License No.: DPR-35 Report No.: 98-08 !
Docket No.: 50-293
Licensee: BEC Energy 800 Boylston Street Boston, Massachusetts 02199 Facility: Pilgrim Nuclear Power Station Inspection Period: September 7,1998 - October 19,1998 Inspectors: R. Laura, Senior Resident inspector 1 R. Arrighi, Resident inspector l D. Dempsey, Reactor Engineer R. Ragland, Jr., Radiation Specialist E. Knutson, Resident inspector-Vermont Yankee Approved by: C. Cowgill, Chief Reactor Projects Branch No. 5 Division of Reactor Projects
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EXECUTIVE SUMMARY Pilgrim Nuclear Power Station i j NRC Inspection Report 50-293/98-08 <
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This integrat' enection included aspects of licensee operations, engineenng, l l maintenance, and piant support. The report covers resident inspection for the period of . !
- Septernber 7,1998, through October 19,1998;in addition, it includes the results of i
[ announced inspections by a health ' physics specialist and the Vermont Yankee resident !
inspector. Further, an engineering specialist performed an in-office review of a motor . .
operated valve open item.
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Operations -
Routine plant operations were well performed including shift turnover and pre-evolutionary briefings. (Section 01.1)
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- Operators performed well during a power reduction to clean two main condenser i waterboxes. Plant management developed a special procedure to provide specific criteria to address potential boundary valve leakage between the waterboxes. After mussel shells and other debris was removed from inside the condenser tubes, overall condenser performance significantly improved. (Sectio: 02.1) J
- ' A primary containment configuration control weakness was identified that'resulted in not entering the related LCO requirements during chemical'Jecentomination of the -
l RHR loops. Several opportunities existed for the licensee's staff to identify this - ;
problem. Contributing to this issue was the improper interpretation of an UFSAR '
table and inadequate communications between operations and systems engineering 1 personnel. (Section 04.1)
- A control room shift staffing issue was properly identified, resolved and reported to the NRC. (Section 08.1)
Maintenance I
i- * Routine surveillance and maintenance activities were generally well controlled.
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Supervisory presence in the field was noted. (Section M1.1)
- Two surveillance probNms were experienced by l&C technicians during this period.
The root cause ard corrective actions taken and planned in response to these issues i was determined to be good. (Section M1.1)
- Mechanical interference was identified between the control rod drive system solenoid directional control valve and the actuating arm of the valve position indicator for the inlet scram valve on four of the hydraulic control units (HCUs).
l' HCU operability was not affected, but the solenoids could have been damaged by l operation of the scram valve. (Section M2.1)
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An effective system has been implemented for processing degraded and non-conforming conditions from identification through correction. Review of open operability evaluations confirmed that individual conditions in the aggregate did not seriously degrade safety systems. (Section E2.1)
A LLRT of a primary containment penetration was well controlled by the system engineer with effective support from the I&C, operations and health physics groups.
(Section M1.1)
Effective engineering problem identification was indicated by subtle discrepancies involving the UFSAR drywell free volume and seismic classification of vacuum relief ,
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valves. These issues were properly evaluated and reported to the NRC. (Sections i E8.3 and E8.4)
Plant Suooort
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- The chemical decontamination of the RHR system included thorough and detailed i planning and preparation as evidenced by a generally thorough safety evaluation, clear and detailed procedural guidance, an equipment setup that guarded against leaks and minimized radiation exposures, and effective health physics access l controls. (Section R1.1) I
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-The chemical decontamination of RHR systems substantially reduced radiation dose rates in the RHR quadrants. This reduction in dose rates eliminated high radiation areas in both RHR quadrants and is expected to result in significant long term radiation exposure savings. (Section R1.2)
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Radiological boundaries were well defined and posted, and housekeeping in the reactor building was generally well maintained. The material condition of the trash compactor facility had improved in that the facility had been cleaned, painted, and color coded to implement human factors to aid waste handling. (Section R2.1) j
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- Overall response by the emergency response organization to the September 15, ;
1998, drill was good. No major concerns were noted. (Section P4.1) i
- Effective self-assessment oversight of the radiological controls program was evidenced by the identification of an inadequate radiological survey and prompt actions taken to investigate, evaluate, and implement corrective actions.
(Section R7)
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TABLE OF CONTENTS E X EC UTIVE S U M M A RY . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . ii
,- - Summary of Plant Status ............................................. 1
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l . O PE R AT I O N S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-O1 Conduct o f Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 4 L 01.1 General Comments (71707) ...........................1
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02 Operational Status of Facilities and Equipment ................... 1 02.1 Condenser Fouling Downpower . . . . . . . . . . . . . . . . . . . . . . . . . 1 04 Operator Knowledge and Performance . . . . . . . . . . . . . . . . . . . . . . . . . 2 04.1 Primary Containment Configuration Control Weakness . . . . . . . . . 2 08 Miscellaneous Operations issues (92901) . . . . . . . . . . . . . . . . . . . . . . . 4 08.1 (Closed) LER 5 0-2 9 3 /9 8-2 0 . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 o
! 11. M AI N T E N A N C E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 L :M1 Conduct of Maintenance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 l
M1.1 General Mainte nance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
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M2 Maintenance and Material Condition of Facilities and Equipment . . . . . . . 6 M2.1 Directional Control Valve / Scram Valve Mechanical Interference .. 6-M8 Miscellaneous Maintenance issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 M8.1 (Closed) LER 50-2 9 3/97-2 5 01 - . . . . . . . . . . . . . . . . . . . . . . . . . 7 l I l l . E N G I N E E R I N G . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 :
E2 Engineering Support of Facilities and Equipment . . . . . . . . . . . . . . . . . . 7 E2.1 Program for Evaluating Degraded, Non-Conforming Conditions ... 7 E8 Miscellaneous Engineering issuer; . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 E8.1 (Closed) IFl 50 2 9 3/9 7-1 J-05 . . . . . . . . . . . . . . . . . . . . . . . . . . 8 E8.2 (Closed) LER 5 0-293/9 7- 15-01 - . . . . . . . . . . . . . . . . . . . . . . . . . 8 E8.3 (Closed) LER 5 0-2 9 3/9 8-0 9 . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 E8.4 (Closed) LER 5 0-2 9 3 /9 7-30 . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 i
I V . PLA N T S U PPO RT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 )
R1 Radiological Protection and Chemistry (RP&C) Controls . . . . . . . . . . . . . 9 R1.1 Chemical Decontamination of the RHR System . . . . . . . . . . . . . . 9 R1.2 Chemical Decontamination Results Achieved .............. 11 R2 Status of RP&C Facilities and Equipment .......... ....... . . . 11 R2.1 Housekeeping and cleanliness . . . . . . . . . . . . . . . . . . . . . . . . . 11 R7 Quality Assurance in RP&C Activities . . . . . . . . . . . . . . . . . . . . . . . . . 12 R7.1 Identification of an inadequate Radiological Survey . . . . . . . . . . 12 R8 Miscellaneous RP&C lssues . . . . . . . . . . . . -. . . . . . . . . . . . . . . . . . . . 13 R8.1 (Closed) IFl 50-2 9 3/9 7-0 8-07 . . . . . . . . . . . . . . . . . . . . . . . . . 13 P4 Staff Knowledge and Performance in EP . . . . . . . . . . . . . . . . . . . . . . . 13 P4.1 Emergency Preparedness Training Drill . . . . . . . . . . . . . . . . . . . 13
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- V. M AN AG EM ENT M EETIN G S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14
' X1: Exit Meeting Summ ary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 INSPECTION PROCEDURES USED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 l l
l ITEMS OPENED, CLOSED, AND UPDATED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 l l
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LI ST O F AC RONYMS U SED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17
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i REPORT DETAll.S i
'Summarv of Plant Statum Pilgrim Nuclear Power Station (PNPS) began the period at approximately 100 percent
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power. There were two power reducti.ons during this inspection period. On September 18 l operators lowered power to 30 percent to clean the condenser water box due to repeated
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fouling. .The 1-3 and 1-4 water boxes were cleaned and the power was returned to full power on September 20. On September.21 operators reduced power to 80 percent to perform a rod pattern adjustment. The plant was at 100 percent power at the end of the report period.
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l. OPERATIONS 01 Conduct of Operations'
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.01.1 General Comments (71707)
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Using Inspection Procedure 71707,the inspector conducted frequent reviews of ongoing plant operations. The inspector observed proper control room staffing, I effective pre-evolutions briefings, and plant behavior was commensurate with the l. plant configuration and plant activities in progress.
l l[ Anomalies noted during plant tours were discussed with the nuclear watch ~
(: ' engineer. For example, the inspector identified shipping plugs were installed on the
[ atmospheric vent port of the salt' service water system pressure transmitters. The l. plugs were removed, instruments recalibrated and testing revealed that the shipping
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plugs did not affect the pressure readings. The licensee properly initiated a problem o report to document, evaluate and correct this condition.
02 : Operational Status of Facilities and Equipment
.02.1 Condenser Foulino Downoower l
l a. insoection Scooe (71707.93702)
During the previous inspection period, the licensee conducted several downpowers to backwash and heat treat the condenser to reduce biofouling. These actions were not fully effective in restoring the normal condenser operational parameters such as differential temperatures. This indicated potential blockage of the condenser tubes in the 1-3 and 1-4 waterboxes. During deep backshift inspection, the inspector observed portions of the downpower activities.
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Topical headings such as 01, M8, etc., are used in accordance with the NRC y standardized reactor inspection report outline. Individual reports are not expected to l address all outline topics.
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b. Observations and'Findinas l
On September 19, operators lowered reactor power to facilitate a condenser bay entry to open, inspect and clean condenser waterboxes 1-3 and 1-4, one at a time.
Close coordination was observed between reactor engineering and control room operators in reactivity management. The licensee developed special procedure TP l 98-032," Temporary Procedure For Performing On-Line inspections of Condenser l Waterboxes," to provide specified criteria for potential loss of water level in the adjacent water box.
The operations staff isolated and released the 1-3 condenser water box for maintenance. The maintenance staff removed mussel shells from inside the condenser tubes and cleaned minor debris from the tube sheet. After returning the 1-3 water box to service, the operators repeated the process for the 1-4 water box j with similar good results. The inspector locally observed backwash and heat t treatment of the waterboxes in the intake structure. The operator stationed in the intake structure was very experienced and performed the required actions in a proper manner. No problems were noted by the inspector. Subsequently, control room operators returned the unit to full power operations with no significant problems. Overall condenser efficiency improved significantly with the water box
. differential temperatures returning to normal. ~
c. Conclusions Operators performed well during a power reduction to clean two main condenser waterboxes. Plant management developed a special procedure to provide specific criteria to address potential boundary valve leakage between the waterboxes. After mussel shells and other debris was removed from inside the condenser tubes, overall condenser performance significantly improved.
1 04 Operator Knowledge and Performance l
04.1 Primary Containment Confiauration Control Weakness a. Insoection Scoce (71707)
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During the chemical decontamination of the RHR loops, the inspector reviewed that the proper technical specification (TS) requirements were entered and action statements followed. Further details on the decontamination process are located in Section R1.1 of this report, b. Observations and Findinos Operators entered four different TS limiting conditions for operation (LCO) that addressed the RHR LPCI function, RHR containment spray, RHR torus cooling and containment integrity. The chemical decontamination process required the j installation of flanges and hoses into several RHR system piping penetrations. The l
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decontamination process was performed one loop at a time; the "B" loop first followed by the "A" loop.
During the "A" RHR loop chemical decontamination process, the inspector noted that only one RHR piping containment penetration (i.e.,23A and 26A valves) was listed as inoperable for TS 3.7.A.2, Primary Containment Integrity. The inspector identified that another ponetration (i.e.,28A and 29A valves) should also have been considered inoperable. This penetration had a temporary hose connected to the drain line downstream of two manually operated valves located between the 28A/29A valves. As part of the chemical decontamination process, these manual 1
- valves were opened to allow process flow. The opening of the manualisolation l valves in the drain line, in effect, rendered the outer containment isolation valve l (28A) inoperable. The inspector determined that the intent of TS 3.7.A.2 was met i since valve MO-1001-29A was deenergized and tagged shut as part of a general chemical decontamination boundary valve tagout. Although the 28A/29A containment penetration was isolated, surveillance requirement 4.7.A.2.b.2, which records the position of the isolated penetration valve each day, was not being performed. The inspector informed the licensee of this observation and they subsequently declared the 28A/29A penetration inoperable and initiated action to comply with the TS surveillance requirement. Since this was done prior to exceeding the TS surveillance interval, no TS violation resulted for the "A" loop.
However, the chemical decontamination had already been completed on the "B" loop and the operators had missed TS surveillance 4.7.A.2.b.2 for that loop. This l was considered a minor violation not subject to formal NRC enforcement. The
! licensee initiated a problem report to document, evaluate and take corrective actions, as needed.
The inspector interviewed operations and system engineering personnel to determine how the relevant TS LCOs were missed for the 28A/29A penetration.
Operations management referenced the RHR injection line check valve located downstream of the 28A/29A valves which is listed in UFSAR Table 5.2-4, Containment and Reactor Vessel Isolation Valves. Although the RHR injection line check valve (i.e., 1001-68) was listed in the table, the inspector identified that the valve was only tested as part of the IST program and not tested pursuant to Appendix J requirements. Hence, the check valve cannot be credited as a l containment isolation valve. The misinterpretation occurred because the table
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contains both containment isolation valves and also reacto: vessel isolation valves.
l Discussions with the system engineer revealed that engineering department personnel were aware of the need to enter the LCO for the 28A/29A penetration.
l The inspector noted that several opportunities existed for the licensee to detect the problem including: the LCO review board for online maintenance, the special test procedure developed for the chemical decontamination and the onsite review committee evaluation. The inspector had no further questions or concern.
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c. Conclusion A primary containment configuration control weakness was identified that resulted in not entering the related LCO requirements during chemical decontamination of the RHR loops. Several opportunities existed for the licensee's staff to identify this l problem. Contributing to this issue was the improper interpretation of an UFSAR
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08 Miscellaneous Operations issues (92901)
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08.1 (Closed) LER 50-293/98-20: Operatina with less Than Minimum Ooeratino Shift
! Crew Composition This LER documented that on July 30,1998, there was no shift technical advisory (STA) coverage for approximately two hours. The on-coming senior licensed operator, who was STA qualified, called in stating that he would be late relieving the shift. An off-going licensed operator supplemented the on-coming shift but failed to recognize that there would be no STA coverage.
The inspector conducted an on-site review of the LER and verified that the licensee generated a problem report to document and correct this condition. The failure to provide adequate shift coverage was considered a minor violation not subject to formal NRC enforcement. This LER is closed. ;
11. MAINTENANCE M1 Conduct of Maintenance
! M1.1 General Maintenance a. Insoection Scope (62707/61726)
i l The inspector observed all or portions of LCO maintenance and surveillance l activities. An emphasis was placed on verifying adherence to maintenance rule data including system unavailability times and component failures. Portions of the
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MR 19802185 High pressure coolant injection (HPCI) logic relay troubleshooting 8.5.2.2.2 "LPCI Loop B Pump and Valve Quarterly Operability" 8.3.2 " Control Rod Exercise" 3.M.3.25.10 " Weekly Pilot Cell and Charger inspection" 8.C.14 " Weekly Pilot Cell, Overall Battery Check, and Battery Charger Test" I 8.M.2-2.5.1 "HPCI Steam Line High Flow Isolation" l 8.7.1.5 Local Leak Rate Testing of Primary Containment Isolation l Valves"
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l b. Observations and Findinas The licensee experienced two surveillance test problems during this inspection period. The first problem occurred on September 15,1998, during the conduct of surveillance 8.M.2-2.5.1, "HPCI Steam Line High Flow Functional Test." The HPCI system operated as designed; however, a digital timer did not start and the time delay of a relay could not be obtained. The instrumentation and controls (l&C) I technicians checked the equipment settings, test connections, and reset the HPCI i system logic. The surveillance was performed again with similar results. During the substitution of the digital timer, the I&C technicians shorted the test connections together causing a 125 volt DC fuse to blow. The fuse was replaced and the I surveillance was repeated two additional times with the new test equipment with j similar results. The test equipment was changed back to the original timer and l retesting performed with no success. Upon removal of the test leads to check the I timer, operations personnel received the "HPCI Power Failure" alarm. The !
surveillance test was stopped, the l&C supervisor and system engineer were notified, and a problem report was written.
The inspector observed the troubleshooting of the HPCI logic relay and investigation l of the power failure alarm. Investigation revealed no damage to the relay, however, !
a second 120 volt DC fuse had blown and pressure switch DPIS-2352 was damaged. The l&C supervision and technicians reviewed electrical drawings and verified that the procedure was correct in how to set up the test equipment, and the digital test equipment was satisfactorily bench tested. The damaged components were repaired and the digital test equipment was replaced with a strip chart recorder. The inspector monitored the subsequent performance of the surveillance; no problems were experienced. The inspector also attended the critique and reviewed the corrective actions for this issue and determined them to be thorough.
No violations were identified by the inspector, l
A second surveillance test problem experienced by the licensee occurred on October 1 2,1998, and resulted in the inadvertent start of the "A" emergency diesel generator l (EDG). The licensee was performing a logic system functional test (LSFT) that required l&C technicians to insert a small rubber boot between energized relay
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j contacts. The l&C technician inadvertently bumped the contact which completed
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the logic and started the EDG. Operators secured the EDG with no problems.
Subsequently, the licensee initiated a problem report and conducted a lessons l learned meeting. The inspector discussed this event with the maintenance master process owner and also conducted in-office review of related LER 98-022. The l licensee performed a detailed root cause evaluation and initiated review of other surveillance procedures that require the use of rubber boots to be placed over relay contacts. No violations were identified by the inspector. LER 98-22 is closed. The licensee corrective actions taken and planned were appropriate.
After completion of the chemical decontamination of the "A" loop of RHR, the inspector observed local leak rate testing (LLRT) of penetration X-39A for the 23A and 26A valves. A LLRT was required since this penetration was used as a tie-in point for the chemical decontamination equipment. The containment system i
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engineer acted as the test director with support from an l&C technician and a licensed reactor operator. Equipment on the LLRT test cart was verified to be in proper calibration. The leakage rate was 46 scem which was significantly less than the acceptance criteria of 7890 secm. Good communication and teamwork were noted between the various work groups. No problems were noted by the inspector, c. Conclusions ' ,
Routine surveillance and maintenance activities wer 7enerally well controlled.
Supervisory presence in the field was noted. ;
Two surveillance problems were experienced by l&C technicians during this period. l The root cause and corrective actions taken and planned in response to these issues !
was determined to be good.
A LLRT of a primary containment penetration was well controlled by the system engineer with effective support from the l&C, operations and health physics groups.
M2. Maintenance and Material Condition of Facilities and Equipment M2.1 Directional Control Valve / Scram Valve Mechanical Interference a. insoection Scope (62707)
On September 23, the inspector observed a mechanical interference problem on four hydraulic control units (HCUs). Specifically, the electrical connector for the directional control valve was rotated and positioned directly over the position indicator for a scram valve. Because the scram valve stem moves upward when the valve opens, it would have hit the solenoid's electrical connection, were a scram to occur. The inspector was concerned that this mechanical interference might slow the opening time of the scram valves and adversely affect the scram time of the associated control rods, b. Observations and Findinas The inspector informed the nuclear watch engineer of the condition. In response to this concern, a problem report was written and the solenoids were rotated to clear the interference. During subsequent investigation, it was determined that the vendor (General Electric) had identified the condition in 1973 and had informed BWR owners via a service information letter (SIL-3). The SIL indicated that the condition could result in damage to the solenoid connector, and recommended rotating the solenoids to clear the interference. In addition, the Sllindicated that the potential for the condition could be eliminated by tying off the solenoid cable to the HCU frame, thereby restricting rotation of the solenoid.
The licensee concluded that the interference condition would not have adversely affect the control rod scram time. The opening force of the scram valve is significantly greater than what would be required to bend the solenoid cable
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connection out of the way. Additionally, only a very small amount of stem travelis required to provide sufficient flow through the scram valve to achieve the required control rod scram time. The inspector reviewed the control rod scram times for the December 6,1997, reactor scram and noted that the scram time for the four HCU's in question were consistent with the other 141 rods, it should also be noted that the directional control valves do not perform a safety function, and therefore the possibility of damage to the solenoid was not a safety concern.
c. Conclusions Mechanical interference was identified between the solenoid directional control valve and the actuating arm of the valve position indicator for the inlet scram valve on four hydraulic control units (HCUs). HCU operability was not affected, but the solenoids could have been damaged by operation of the scram valve.
M8 Miscellaneous Maintenance issues (92902)
M 8.1 (Closed) LER 50-293/97-25-01: Shutdown due to Two Inocerable Main Steam Lsplation valves The inspector conducted an in-office review of LER 97-25-01. This issue was previously reviewed and documented in Section M2.1 and E1.1 of NRC Inspection Report No. 50-293/97-13as a violation of NRC requirements. The violation was subsequently closed out in Section E8.1 of NRC Inspection Report No. 50-293/98-06. This LER is closed.
Ill. ENGINEERING E2 Engineering Support of Facilities and Equipment E2.1 Proaram for Evaluatina
Dearaded,
Non-Conformina Conditions The inspector reviewed Pilgrim's program for justifying continued operation pending corrective action for structures, systems, and components with identified degraded, non-conforming conditions. The inspector reviewed the licensee procedures and determined that they established appropriate guidance for dispositioning degraded, non-conforming conditions. The inspector determined that the use of probabilistic risk assessment (PRA)importance measures (Fussell-Vesely and Risk Achievement Worth) to prioritize corrective actions was an appropriate application of PRA to the process. The inspector reviewed the active operating evaluations for possible cumulative effects within and/or between systems and noted no problems. In addition, the inspector examined several recently identified degraded, non-conforming conditions from another nuclear facility to determine if they may be applicable to Pilgrim, and concluded that they were not applicable. The inspector concluded that Pilgrim had established an effective system for processing degraded and non-conforming conditioas from identification through correction.
E8 Miscellaneous Engineering issues (92903)
E8.1 (Closed) IFl 50-293/97-13-05: Complete Draft Evaluation of Motor-Operated Valve Coefficients of Friction During the Generic Letter (GL) 89-10 motor-operated valve (MOV) prograrn closeout inspection, the licensee had not completed its internal review and design verification of calculation M-772, " Evaluation of MOV Coefficients of Friction." The calculation subsequently was approved on December 11,1997. The inspector performed an in-office review of the final calculation. The licensee performed a statistical analysis of the data from 68 static and 31 dynamic tests. Applying a 2-sigma sample standard deviation of the mean of the data, the analysis justified the Pilgrim GL 89-10 program assumption of 0.2 for stem friction coefficient. This item is closed.
E8.2 (Closed) LER 50-293/97-15-01: Incorrect Salt Service Water Pumo Overload Settina The inspector conducted an in-office review of LER 97-15-01. This issue was previously reviewed and documented in Section E.2.3 of NRC Inspection Report No.
50-293/97-13as an apparent violation of NRC requirements. This issue was subsequently non-cited in accordance with Section Vll.B.4 of the Enforcement Manuel in NRC letter dated April 27,1998. This LER is closed.
E8.3 (Closed) LER 50-293/98-09: Incorrect Seismic Values Used for Vacuum Relief Valves This LER documented that incorrect seismic values were used for the main steam safety relief valve line vacuum breakers when they were originally purchased. An engineering evaluation was performed that concluded that the valves were operable.
The valve supplier reviewed the valve design calculations and indicated that the valves could be qualified to the higher seismic requirements. The licensee committed that they would determine the seismic capability of the valves and modify, as required, to assure full seismic qualifications before startup from the cycle 12 refueling outage.
The inspector conducted an on-site review of the LER and reviewed the corresponding engineering evaluation and proposed corrective actions and found them to be appropriate. The long term corrective actions to restore full qualification of the vacuum breakers were being pursued via problem report 98.9208. The lack of seismic qualification is considered a violation of NRC design control requirements.
This non-repetitive licensee-identified and corrected violation is being treated as a Non-Cited Violation (NCV 50-293/98-08-01), consistent with Section Vll.B.1 of the NRC Enforcement Policy. LER is closed.
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l 9 E8.4 (Closed) LER 50-293/97-30: Misaonfication of FSAR Drvwell Free Volume This LER documented a discrepancy between the value for the drywell free volume j stated in the FSAR (147,000 cubic feet) and that in a draft engineering calculation ,
(137,000 cubic feet). The 147,000 cubic feet value was used as a design input for j a variety of desian basis calculations and analyses. '
The inspector conducted an on-site review of the LER and reviewed the licensee's engineering evaluation. There were no identified operability concerns for the analyses as a result of the reduction of drywell free air volume. The licensee committed to revise the FSAR to clarify the drywell free volume value and complete a review of the affected drywell volume calculations. Other calculations and analysis affected by the inappropriate drywell value will be corrected and/or reconciled, as necessary, as part of the Design Basis Recovery program. 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 50- l 293/98-08-02). This LER is closed. i IV. PLANT SUPPORT R1 Radiological Protection and Chemistry (RP&C) Controls R1.1 Chemical Decontamination of the RHR System The residual heat removal (RHR) system rooms (" quads") located within the reactor building at Pilgrim Nuclear Generating Station (PNGS) contain major components of the RHR system including piping, valves, pumps, and heat exchangers. Due to a combination of corrosion product buildup and space limitations, a significant amount of personnel radiation exposure has historically been received at PNPS from tours, inspections, and maintenance activities in the RHR quads. As part of an overall dose reduction strategy, plans were made to chemically decontaminate both loops of the RHR system. The licensee used a three step (ClTROX-NP-CITROX) chemical decontamination process. The first step applied CITROX, a combination of citric acid and oxalic acid, which was continually regenerated by passing it through a cation resin bed. Next, nitric acid and potassium permanganate (NP) were applied to condition the chromium rich oxide layer to allow its removal during the cleanup step. This was followed by an oxalic acid rinse to dissolve manganese dioxide.
Finally, another CITROX step was applied.
The decontamination effort required significant planning and coordination and was performed while the plant was on-line. This required a technical specification revision to allow isolation of an RHR loop and entrance into a 7-day limiting condition of operation (LCO) for decontamination of each loop, temporary power supplied by diesel generators, and temporary connections to turbine building closed cooling water (TBCCW), condensate water storage and transfer (CST),
demineralized water, and instrument air systems.
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10 l a. Inspection Scone (86750)
A review was performed of the preparation, planning, and implementation of chemical decontamination of residual heat removal (RHR) systems. Information was gathered by a review of applicable documents including the safety evaluation and procedural guidance, through tours of the facility to observe equipment setup and l health physics access controls, and through interviews with key individuals
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responsible for planning and implementation of chemical decontamination activities.
b. Observations and Findinas Safety evaluation No. 3172," Chemical Decontamination of the Residual Heat ,
Removal System," included a description of the chemical decontamination process, I a system / component failure consequence analysis, and an evaluation of the impact and potentialimpact of the chemical decontamination on affected systems including mechanical, structural, electrical, instrumentation and control, radiological, l radwaste, chemistry, and the safety function of affected systems. This included an j evaluation of the impact if decontamination chemicals were introduced into the l reactor vessel, torus, and RHR system components. The safety evaluation i concluded that chemical decontamination activities would have a minimalimpact on vital system components, events were bounded by the FSAR, and changes associated with the chemical decontamination would not involve an unreviewed safety question.
Procedural guidance developed for chemical decontamination activities was clearly l written, included signoffs for key steps / activities, contained appropriate details, and I provided evidence of thorough planning.
Equipment setup was well controlled and thought out as evidenced by comprehensive and detailed procedural guidance, use of camlock adapters and plastic to guard against leaks, positioning of equipment and hoses outside of high traffic areas, and properly sized hose lengths to minimize clutter.
Health physics access controls included establishment of an outer boundary to limit access to " essential" personnel only, use of radiation work permits and thorough l radiological work briefings, use of water and lead shielding around cation resin I columns and the decon skid, and positioning of work stations and cation resin columns to minimize personnel exposure.
c. Conclusions l The chemical decontamination of the RHR system included thorough and detailed l planning and preparation as evidenced by a generally thorough safety evaluation, clear and detailed procedural guidance, equipment setup that guarded against leaks and minimized radiation exposures, and effective health physics access controls.
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R1'.2 Chemical Decontamination Results Achieved a. Inspection Scope (83728)
A review was performed of the results of the RHR chemical decontamination
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project. Information was gathered through a review of radiological surveys and through discussions with cognizant personnel.
i b. Observations and Findinas j The effectiveness of the chemical decontamination was evaluated and characterized I with decontamination factors (DFs). DFs were calculated by first selecting l representative dose rate measurement base points based on knowledge of personnel I occupancy during inspection and maintenance activities, and then obtaining and j dividing pre-decontamination dose rates by post-decontamination dose rates. l l Although specific component base points were continually monitored with teledosimetry during the decontamination effort, general area base points were used to characterize the effectiveness of the chemical decontamination. A shielded l E-530 (directional) probe was used to establish component DFs and an RO-2 ion l chamber was used to establish the general area DFs.-
l Radiological surveys showed that general area dose rates in the RHR quads were
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reduced by a factor of approximately three (3). This eliminated high radiation areas l in both RHR quads allowing both RHR quads to be down-posted to radiation areas.
Based on a review of occupancy, radiation dose savings were estimated to be 36 person-rem per outage year and 9 person-rem per operating year.
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The chemical decontamination of RHR systems substantially reduced radiation dose rates in the RHR quads. This reduction in dose rates eliminated high radiation areas in both RHR quads and is expected to result in significant long term radiation i exposure savings.
! R2 Status of RP&C Facilities and Equipment R2.1 Housekeeoina and cleanliness a. Insoection Scoce (86750)
Plant tours were conducted to evaluate housekeeping and cleanliness, material conditions, and maintenance of radiological boundaries. Areas inspected included 23' of the reactor building, -13' of the radwaste building, and the trash compactor j facility.
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b. Observations and Findinas Radiological boundaries were well defined and posted, and housekeeping in the reactor building was generally well maintained. The material condition of the trash compactor facility had improved in that the facility had been thoroughly cleaned, painted, and color coded to implement human factors to aid waste handling.
c. Conclusion Radiological boundaries were well defined and posted, and housekeeping in the reactor was generally well maintained. The material condition of the trash compactor facility had significantly improved in that the facility had been thoroughly cleaned, painted, and color coded to implement human factors to aid waste handling.
R7 Quality Assurance in RP&C Activities R7.1 Identification of an inadeauate Radioloaical Survev a. Inspection Scope (92904)
A onsite and in-office review was performed of the use of problem reports to identify, evaluate, and resolve radiological control deficiencies. Information was gathered by a review of Problem Report No. 98.1990.00," Radiological Survey Not Reflective of Actual Conditions" and interviews with cognizant personnel, b. Observations and Findinas Title 10CFR20.1501,Subpart F, indicates that each licensee shall make or cause to be made surveys necessary to comply with regulations and are reasonable under the circumstances to evaluate the extent of radiation and radiological hazards present.
On August 13,1998, the acting Radiation Protection Manager (RPM) determined that a radiological survey map for an outside radioactive materials storage area adjacent to the radwaste truck lock (RWTL) did not reflect the presence of or dose rates associated with an onsite storage container (OSSC) seated on a 40' flatbed trailer that had recently been placed in the area, and was not accurate and complete.
Licensee corrective actions included an immediate resurvey of the affected area; verification of other surveys and work activities to validate performance of the RP technicians; formal documentation of the investigation in the problem reporting system; disciplinary action and remedial training for the responsible technician, and notification of the NRC. The licensee's investigation concluded that this was an isolated event with low risk significance due to the low radiation levels (e.g.,3 mR/h) present in areas adjacent to the OSSC. This 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 50-293/98-08-03)
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Effective self-assessment oversight of the radiological controls program was evidenced by the identification of an inadequate radiological survey and prompt .
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actions taken to investigate, evaluate, and implement corrective actions.
R8 Miscellaneous RP&C lasues R8.1 (Closed) IFl 50-293/97-08-07:Lav up of Radwaste Floc Reevele Tank This item was originally opened because the use of the floc recycle tank was !
discontinued and it could not be determined if the tank had been properly drained and isolated from plant systems for long term lay up. During this inspection, a I September 24,1998, video tape of the floc recycle tank was reviewed. The video showed that the stainless steel tank was intact and appeared structurally sound,
- with no signs of active or recent leakage, and there was no accumulation of resins on the floor. The tank was vented and reportedly full of powdex resin. Results of l acoustical testing performed by a vendor indicated that the tank contents were dry.
The nuclear services master process owner indicated that a long term floc recycle tank inspection frequency was being evaluated and a revised description of the status of the floc recycle tank was being prepared for a future revision to the FSAR.
This item is closed. l P4 Staff Knowledge and Performance in EP P4.1 Emeroency Preoaredness Trainina Drill a. Insoection Scoce (71707)
The inspector observed the performance of the licensee's emergency response organization during the September 15,1998, emergency preparedness training drill in the technical support and operations support center (OSC). The inspector also attended the licensee's post-exercise drill critique to evaluate the licensee's self-assessment of the exercise.
b. Observation and Findinos All of the licensee's training objectives were met; no major deficiencies were noted.
Simulated events were properly diagnosed, proper mitigation actions performed, and emergency declarations were timely and accurate. The inspector did identify one concern regarding an OSC team being sent out to the reactor water cleanup area, which had an obvious (simulated) airborne radiation condition, without self-contained breathing apparatus (SCBA). The inspector informed the licensee of this observation, and verified that the OSC team members were SCBA qualified.
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Overall response by emergency response organization to the September 15,1998, drill was good. No major concerns were noted. (Section P4.1)
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V. MANAGEMENT MEETINGS X1 Exit Meeting Summary The inspector presented the findings of this inspection to members of the licensee management on November 12,1998. The licensee acknowledged the findings presented.
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INSPECTION PROCEDURES USED
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IP 37551: Onsite Engineering )
IP 40500: Effectiveness of Licensee Controls in Identifying, Resolving, and Preventing l Problems ;
IP 61726: Surveillance Observation i IP 62707: Maintenance Observation !
IP 71707: Plant Operations
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IP 71750: Plant Support Activities - i IP 82301: Evaluation of Exercises for Power Reactors '
IP 83728: Maintaining Occupational Exposures ALARA IP 92700: Onsite Followup of Written Reports of Nonroutine Events at Power Reactor Facilities IP 92901: Followup - Operations i
IP 92902: Followup - Maintenance '
lP 92903: Followup - Engineering j IP 92904: Followup - Plant Support i IP 93702: Prompt Onsite Response to Events at Operating Power Reactors '
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ITEMS OPENED, CLOSED, AND UPDATED l
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l IFl 50-293/97-08-07 Lay up of Radwaste Floc Recycle Tank IFl 50-293/97-13-05 Complete draft evaluation of motor-operated valve coefficients l of friction I l
LER 50 293/97-15-01 incorrect Salt Service Water Pump Overload Setting
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- LER 50-293/97-25-01 Shutdown due to Two Inoperable Main Steam isolation Valves -
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- LER 50-293/97-30 Misapplication of FSAR Drywell Free Volume l
l- LER 50-293/98-09 incorrect Seismic Values Used for Vacuum Relief Valves I
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l LER 50-293/98 20 Operating with less Than Minimum Operating Shift Crew Composition .l l
LER 50-293/98-22. Inadvertent actuation of the "A" Emergency Diesel Generator
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l NCV 50-293/98-08-01 Incorrect Seismic Values Used for Vacuum Relief Valves t
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l NCV 50-293/98-08-03 Inadequate Radiological Survey'
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! 17 LIST OF ACRONYMS USED
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ALARA- As Low As is Reasonably Achievable BECo Boston Edison Company _
' CFR~ Code of Federal Regulations CST- ' Condensate Water Storage and Transfer 4 DF decontamination factor: (pre-decontamination measurement)/(post-decontamination measurement) expressed as a numerical value DRP Oivision of Reactor Projects i EP Emergency Preparedness FSAR Final Safety Analysis Report i HCU ' Hydraulic Control Unit I l&C - Instrumentation and Controls !
IFl inspection Follow-Up item ,
IR inspection Report l LCO Limiting Condition of Operation LER Licensee Event Report MOV- Motor-operated Valve i MR _ Maintenance Request I mR/h - milliroentgen per hour i I
NCV' Non-Cited Violation NOV Notice of Violation
'NRC Nuclear Regulatory Commission NRR- Office of Nuclear Reactor Regulation 1 OSC- Operations Support Center I
~OSSC Onsite Storage Container I PDR Public Document Room !
PNPS Pilgrim Nuclear Power Station l
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PR Problem Report -l PRA Probabilistic Risk Assessment i RFO Refueling Outage RHR Residual Heat Removal RP. . Radiological Protection RPM radiation protection manager RWTL radwaste truck lock SCBA Self Contained Breathing Apparatus SIL Service information Letter STA Shift Technical Advisory TBCCW. Turbine Building Closed Cooling Water UFSAR Updated Final Safety Analysis Report VIO Violation A
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