IR 05000280/1994031
| ML18153B195 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 12/22/1994 |
| From: | Branch M, David Kern, Tingen S NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18153B194 | List: |
| References | |
| 50-280-94-31, 50-281-94-31, NUDOCS 9412290219 | |
| Download: ML18153B195 (20) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTA STREET, N.W., SUITE 2900 ATLANTA, GEORGIA 30323-0199 Report Nos.:
50-280/94-31 and 50-281/94-31 Licensee: Virginia Electric and Power Company Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060 Docket Nos.:
50-280 and 50-281 License Nos.:
DPR-32 and DPR-37 Facility Name:
Surry 1 and 2 Inspection Conducted:
November 6 through December 10, 1994 Inspectors:
M~W~B~enio~sident Inspector D~c_JKe~nt 'Y,;;pector S~~Ti~denfrn=spector Accompanybiyn:. Inspe~cto?r: \\* P. Kin Approved
~
~
~el isle, Section Chief Divisi n of Reactor Projects SUMMARY Scope:
I 1-r- 'l-Z...- 1 f Date Signe I 2-2-2.r- ?cf Date Signe 12.- i,z-9f Date Signed This routine resident inspection was conducted on site in the areas of plant status, operational safety verification, maintenance inspections, surveillance inspections, onsite engineering review, plant support and Licensee Event Report followu Inspections of backshift and weekend activities were conducted on November 6, 13, 22, 25, 26 and December 1 and 10, 199 PDR ADOCK 05000280 G
- Results:
Operations:
Operators conducted plant cooldown, loop isolation, and loop draindown evolutions in a controlled and professional manne Control room personnel demonstrated a thorough knowledge of plant conditions and available alternate indications of plant parameters (paragraph 3.2).
Management reevaluated the reliability of the containment sump isolation valves when outboard isolation valve failed to fully shut following sump pumpdown under administrative controls. The resulting Station Nuclear Safety and Operating Committee recommendation to maintain 1-DA-TV-1008 shut demonstrated a sound safety perspective (paragraph 3.4).
Containment sump isolation valve 1-DA-TV-IOOA failed to close following routine performance testing. Operations personnel promptly isolated and deactivated the second isolation valve in the line in accordance with Technical Specifications. Local leak rate testing subsequently verified that the valve was not fully closed. This testing was satisfactorily performed in that procedures were followed and no significant problems were encountered during the test (paragraphs 3.4 and 5).
Maintenance:
Installation and test instructions for the maintenance department to implement a design change were not specified in the work order task packag Temporary corrective actions implemented to prevent this problem from recurring were adequate (paragraph 4.1).
The licensee's on-line maintenance practices were acceptable per Temporary Instruction 2515/126, Evaluation of On-Line Maintenance (paragraph 4.2).
Outage planning management generally coordinated maintenance activities and outage schedule in an effective manne Emerging material problems such as a stuck open loop isolation valve were promptly addresse However, maintenance activities associated with Unit 1 Blow head safety injection pump breaker, charcoal media testing and IA station battery cell replacement were not as effectively coordinated. Discrepancies in scheduling these maintenance activities were identified by operations personnel and appropriate corrective actions were initiated. Station management's production perspective was well focused and did not permit outage schedule to interfere with safe maintenance and plant operation (paragraph 4.3).
Corrective actions implemented to correct spiking/moisture intrusion into ventilation vent and process ventilation radiation monitors resulted in improved system operation (paragraphs 8.1 and 8.2).
Engineering:
An unresolved item was identified involving design requirements for operation with Unit 2 containment reactor cavity drain valves shut (paragraph 3.3).
Engineering evaluations of a stuck open reactor coolant loop isolation valve and a failed containment isolation valve were comprehensive (paragraphs and 6.1).
The licensee implemented a temporary modification to the ventilation ~ystem to reduce the potential for chemical contamination of the filters during the Unit 1 Steam Generator Chemical Cleaning (SGCC) outage. Precautionary measures were prudent and represented a tangible product of the licensee's evaluation of a previous Unit 2 filter contamination event. Statfon Nuclear Safety and Operating Committee's review of the modification and supporting procedure revisions was detailed. Operators were knowledgeable of the revised procedures and closely monitored containment purge. Analyses of charcoal samples taken before and after the SGCC showed no degradation of the l-VS-FL-38 filter (paragraph 6.2).
Plant Support:
Security access controls during Unit 1 outage majntenance activities were good (paragraph 7.1).
On November 25, the licensee declared a Notification of Unusual Event in response to an explosion in the station radioactive waste treatment facilit An unexpected rapid chemical reaction overpressurized a waste processing tan Personnel promptly verified that the overpressurizat1on did not cause a fire, release of radioactivity, or damage to safety related system The licensee's classification of this event was appropriate. The emergency plan was properly implemente An Inspection Followup Item was opened to review the work
controls that allowed the event to happen (paragraph 7.2) *
- REPORT DETAILS Persons Contacted.1 Licensee Employees
- W. Benthall, Corporate Licensing H. Blake, Jr., Superintendent of Nuclear Site Services
- R. Blount, Superintendent of Maintenance
- D. Christian, Station Manager J. Costello, Station Coordinator, Emergency Preparedness J. Downs, Superintendent of Outage and Planning
- D. Erickson, Superintendent of Radiation Protection
- A. Friedman, Superintendent of Nuclear Training D. Hayes, Superintendent of Administrative Services
- R. Hayes, Supervisor, Quality Assurance C. Luffman, Superintendent, Security
- J. McCarthy, Assistant Station Manager
- A. Meekins, Nuclear Site Services
- A. Price, Assistant Station Manager S. Sarver, Superintendent of Operations R. Saunders, Vice President, Nuclear Operations E. Smith, Site Quality Assurance Manager
- D. Sommers, Supervisor, Licensing
- T. Sowers, Superintendent of Engineering
- J. Swientoniewski, Supervisor, Station Nuclear Safety G. Woodiell, Nuclear Training Other licensee employees contacted included plant managers and supervisors, operators, engineers, technicians, mechanics, security force members, and office personne ~2 NRC Personnel M. Branch, Senior Resident Inspector
- D. Kern, Resident Inspector
- S. Tingen, Resident Inspector
- Attended Exit Interview Acronyms and initialisms used throughout this report are listed in the last paragrap Plant Status Unit I operated at reduced power in order to minimize SG level oscillation On November 28, the unit was shutdown from 96% power for a planned outage to repair the C RCP seal and chemically clean all three
SG At the end of the inspection period, the unit was in cold shutdow Unit 2 operated at power for the entire inspection perio.
Operational Safety Verification (71707} Biweekly ESF Inspections.1.1 Unit 1 Inside Containment Recirculation Spray Pumps On November 29, the inspectors walked down the recirculation spray pumps located inside the Unit 1 containment. The containment sump was verified to be clean and the pump discharge test spool pieces were verified to be removed and properly store.1.2 Units 1 and 2 AVEFTs On December 5, the inspectors walked down the Units 1 and 2 AVEFTs located in the upper level of the auxiliary buildin Equipment was in good overall condition and housekeeping was also goo Manual dampers were verified to be in the proper positions, motor operated dampers were not leaking oil, oil reservoirs were full and the ventilation system modification to support unit 1 SGCC was properly installed. The TM is further discussed in paragraph Unit 1 Cooldown and Loop Isolation The Unit 1 reactor was shut down on November 28 for a planned maintenance outage. The inspectors observed plant cooldown and loop isolation activities from the control room and various locations within the plant. Station procedures limit RCS cooldown rate to 50 degrees F per hour, which is half of that permitted by T RCS pressure and temperature were continuously recorded on the analog trend recorder. Operators manually plotted RCS pressure and temperature at 25 degree intervals to verify that the RCS was maintained within required operating parameter Pressure, temperature, and integrated cooldown rate data points were also monitored on the P250 process computer screen. The inspectors discussed RCS TS requirements and methods of trending the cooldown with control room personne Licensed operators demonstrated thorough knowledge of plant conditions and available alternate indications of plant parameter The SS provided close oversight throughout the evolution. Operators effectively controlled plant equipment which could adversely effect RCS pressure or temperature during the cooldow The inspectors concluded that operators conducted plant cooldown evolutions in a deliberate and professional manne ~-~~-
The inspectors reviewed procedures l-GOP-2.4, RCS Cooldown From HSD to 345 degrees F, revision 6, and l-GOP-2.6, RCS Cooldown from 195 degrees F to Ambient, revision These procedures provided good detail and were clearly understood by control room operator TS requirements for the plant cooldown were clearly stated in the procedures. Procedure l-GOP-2.4 was written assuming the MS trip valves were open and directed operators to adjust the steam header pressure control valve to control RCS cooldown rate. However, the inspectors observed that the MS trip valves were shut *and that operators used the MS trip valve bypass valves to control the RCS cooldown rate. The SS had directed operators to use the MS trip valve bypass valves in accordance with l-OP-MS-005, Administrative Control Of MS Trip Bypass Valves, revision 0, to adjust the cooldown rate. Both methods of controlling cooldown were effective, but cooling down using the bypass valves was not addressed in procedure l-GOP-The inspectors discussed this observation with the Operations Superintendent who initiated appropriate action to improve the procedur The C RCS loop hot leg isolation valve (l-RC-MOV-1594) failed to move when operators attempted to shut the valve to isolate the C RCS loo Electricians monitored motor current and thrust as operators attempted to close the valve agai The operating history of the valve suggested that the valve may have been previously backseated with sufficient force to cause the disk to be mechanically stuck above the seat. Operations and maintenance personnel coordinated effectively to gather information to support technical assessment of stem integrity (Section 6.1). The inspectors observed that several of the subsequent procedural steps for loop isolation were completed despite l-RC-MOV-1594 failing to shut. Completion of these additional steps with the loop isolation valve open did not adversely effect the evolutio However, procedure l-OP-RC-006, Isolation and Drain of One Reactor Coolant Loop, revision 4, does not provide leeway for subsequent steps to be completed out of orde The inspectors discussed implementation of the loop isolation procedure with the SS and questioned how the position of C RCS components would be confirme The SS acknowledged that the steps should not have been performed out of order and verified that the valves and breakers were subsequently returned to the unisolated positio Additionally, based upon the scope of planned maintenance, Operations management had directed that a system lineup of the C RCS loop be performed prior to reactor startup. The inspectors observed isolation and drain down of the A RCS loop and noted that the procedure was performed in the correct sequenc The inspectors determined that performance of steps out of order during the attempted C RCS loop isolation was an isolated occurrence and did not indicate a programmatic breakdow Operators effectively quantified loop isolation valve leakage and determined it to be acceptable. Operators demonstrated professional communications and effectively monitored RCS parameters during the loop draindown evolution *
4 Reactor Cavity Drainage Valves On November 14, the inspectors conducted a review of the Unit 2 PSL. * The purpose of this review was to identify any ESF systems or components that may be in a degraded or inoperable conditio The PSL is divided into system groups and covers both safety-related and non-safety system The inspectors' review identified an issue associated *with the posifion of two normally open reactor cavity fuel transfer canal drain valves (2-RL-11 and 2-RL-12). These two valves were closed and the PSL contained an April 26, 1993, memo from the system engineer to the Superintendent of Engineering describing the situation which required this valve configuration. According to the memo, during the spring 1993 RFO, a higher than expected amount of radioactive material was deposited in the reactor cavity transfer canal during fuel movemen The inspectors held discussions with engineering personnel and were provided the following informatio Engineering had evaluated the impact of operating Unit 2 with two reactor cavity fuel transfer canal drain lines closed. After consultation with NA&F, site engineering concluded that the potential water holdup in the transfer canal during an accident would be inconsequential.
NA&F reconfirmed this conclusion in that, the potential holdup would equate to less than 2 inches of water in the containment sump, with an upper bound effect on the order of 0.1 foot on available NPSH for the safety-related pumps taking suction from the sum Additionally, the licensee concluded that the maximum impact would be expected to occur well after minimum NPSH becomes a concern since collection of spray water in the canal would be a gradual process. The licensee held discussions with Stone and Webster who confirmed that the design function of 2-RL-11 and 2-RL-12 is normal operational drainage, not post accident drainag Based on the information provided by the licensee, the inspectors had additional questions associated with design features of the containmen The inspectors questioned the licensee as to the design flow paths available to the reactor coolant sump from the refueling and reactor cavity areas; specifically, what flow paths are available once the transfer canal fills and begins to spill over into the refueling cavit The licensee provided the inspectors the following information in a memo from K. L. Basehore to T. B. Sowers dated November 22, 199 The memo stated the following:
Design basis LOCA calculations conservatively assume that approximately 70,000+ gallons are instantaneously transferred to and retained in the reactor cavity area.
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Accumulation or holdup of water in the reactor cavity would be limited by several leakage paths to the containment sump area as follows: There is a ventilation system distribution header that discharges into the incore sump room next to the access ladder. Should the incore sump room fill to the top of the access ladder well, the ventilation system provides a flow path through the 60 by 60 inch ventilation plenum which encircles the biological shield wall, and downward through the three containment fan cooler supply headers and to the sum.
There is an access door through the shield wall into the incore room at CL elevation. This door is normally closed but is not leak tight, as evidenced by the experience with the 1988 loss of the reactor cavity sea.
There is a steel pipe penetration at CL elevation that extends through the shield wall and transitions upward to a vertical duct. According to specifications, this duct is to be supported for 6500 pounds flooded weigh.
Postulation concerning the unavailability of all of the above paths and any other miscellaneous paths not cited here, would still lead to the conclusion that the water level could rise to the elevation of the six hot and cold leg penetrations through the primary shield wall without significant additional inventory holdu The licensee concluded, based on their investigations into this issue and discussion with Stone and Webster, that holdup of significant amounts of water in the reactor cavity above the-7' 2" elevation is not a realistic postulate. They concluded also that there are numerous pathways for drainage to the sump, and the expected rates of accumulation of spray water in the reactor cavity region would not be so large as to overwhelm this drainage capabilit The inspectors' review of this issue identified that the licensee does not have a well documented design basis for this important safety feature of the containment. Additionally, the inspectors question the licensee's conclusion that valves 2-RL-11 and 2-RL-12 are only for draining the transfer canal after refueling since plant drawing ll548-FM-1I8A, revision 1, depicts two paths for draining this are One normally shut path goes to the suction of the reactor cavity purification pump and the other normally open path goes directly to the containment accident sum Pending additional review by the NRC, this item is identified as URI 50-281/94-31-01, Containment Design Features.
6 Monthly Verification of Containment Isolation Valves Lineups The inspectors walked down containment penetrations and verified that ~ontainment isolation valves were properly aligned, vent/drain connections were capped and/or t~at there was no leakage from the piping. Unit 1 containment penetrations 76, 77, 78, 87 and 88 were inspected inside and outside of containmen Also, Unit 2 containment penetrations 76, 77, 78, 87 and 88 were inspected outside of containment. These penetrations contain MFW and AFW supply piping to the SG On November 16, containment sump inboard isolation valve 1-DA-TV-lOOA failed to close during routine performance testin Operations personnel promptly isolated and deactivated the second isolation valve in the line, 1-DA-TV-lOOB, in accordance with TS 3.8.C. The inspectors independently verified that 1-DA-TV-lOOB was closed and that the actuating air supply had been properly isolated. The valves were properly tagged in the isolated position to assure positive control. The licensee established administrative controls (AC# Sl-94-1116) to enable periodic opening of 1-DA-TV-lOOB for pumpdown of the containment sum TS 3.8.C permits intermittent repositioning of a deactivated isolation valve when proper administrative controls are established. The inspectors reviewed SE 94-201, Administrative Control For 1-DA-TV-lOOB, dated November 17 and AC# Sl-94-1116 and determined that administrative controls for 1-DA-TV-lOOB were adequately evaluated and safely establishe On November 19, 1-DA-TV-lOOB failed to indicate fully closed after pumping the containment sump under administrative control The licensee promptly entered a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> LCO to reestablish containment integrity. Operations personnel pumped the sump again to flush the valve seat clear, successfully closed the valve, and reestablished containment,integrity. Management reevaluated the reliability of the containment sump isolation valves and directed that 1-DA-TV-lOOB.remain shut until after the planned Unit 1 shutdown on November 2 The inspectors observed SNSOC discussion of the supporting SE and concluded that the decision to maintain 1-DA-TV-lOOB shut demonstrated a sound safety perspectiv.5 Monthly Review of Safety-Related Tagouts The inspectors reviewed the following safety-related tagouts and verified that the tagouts were prepared and implemented in accordance with OPAP~OlOO, Tagouts, revision 4, and that the tagged components were in the required positio Tagging Record No.: l-94-VS-0227, Implement DCP 90-08 Tagging Record No.: l-94-SI-0154, Implement DCP 93-088 Within the areas inspected,.one URI was identifie *
~---------- Maintenance Inspections (62703)
During the reporting period, the inspectors reviewed the following maintenance activities to assure compliance with the appropriate procedure.1 Installation of Unit 1 SI Test Jacks On May 11, 1994, the Unit 1 reactor trip breakers were manually tripped open by operators due to low SG level Low SG levels occurred during surveillance testing when an l&C technician inadvertently bumped a relay that caused the B MFW Pump to tri The event was discussed in NRC Inspection Report Nos. 50-280/94-12 and 50-281/94-12 and LER 50-280/94-00 One of the corrective actions to prevent recurrence of a similar event was to relocate the test points for the train A and B master relays, SIA-A and SIA-B, to test jacks located on the front of the SI master relay the logic cabinet doo On November 30, the inspectors witnessed the installation of a dual post test jack on SI train A master relay SIA-A logic cabinet doo The inspectors also witnessed the installation of wiring from the test jack to relay SIA-A coil terminal points. This work was accomplished in accordance with WO 38298912-01 *
The installation of the test jacks was a permanent hardware change and therefore considered a design chang DC 93-088-03, SI Racks Test Jacks/Surry/1&2, dated July 26, 1994 provided the specifications, drawings, and bill of materials required to accomplish this design change. The DC did not or was not required to provide work instruction I&C technicians performed the design change via skill of the craft and DC 93-088-03 was used for guidance. Technicians documented the work and testing performed to accomplish the design change in the W VPAP-2002, Work Request and Work Order Tasks, revision 5, step 6.12.3, states that when implementing a DC the planner shall assure all installation or testing requirements are specified in the work order task packag The inspectors concluded that installation and test instructions to implement DC 93-088-03 were not specified in the work order task package; therefore, the DC was accomplished via skill of the craft in lieu of obtaining supplemental work instruction In the previous inspection period, Violation 50-280, 281/94-28-01 was identified for failure of personnel to stop work when instructions could not be followe The cause of this violation was attributed to inadequate work instructions for implementation of a D The licensee stated that the response to this violation will address corrective actions to ensure that adequate work and test instructions are specified in work order task packages or work plans for accomplishing DC * *
DCs can be accomplished by NSSs or the maintenance departmen The inspectors have previously reviewed DCs performed by NSSs and concluded that installation and testing procedures were properly implemente As temporary corrective actions for the violation and this recent event, the Superintendent of Maintenance issued instructions to maintenance personnel informing them that DC implemen~ation will be by procedure or supplementary work instructions. Only the pre-job portions of DCs listed on the plan of the day would be worked until the procedure or supplementary work instructions are prepare The inspectors considered the temporary corrective actions to be adequat Evaluation of On-Line Maintenance (Tl 2515/126}
The objective of TI 2515/126, Evaluation of On-Line Maintenance, was to evaluate the licensee's procedures and practices regarding the removal of equipment for on-line scheduled maintenanc The inspectors questioned the licensee to determine what procedures were in place to assess the risks associated with doin maintenance on equipment while the plant was on-line. The inspectors reviewed a memorandum to all supervisors from the Assistant Plant Manager, Operations and Maintenance, dated March 13, 199 The memorandum documented the station philosophy regarding maintenance and TS action statements and required that:
Safety related equipment not be removed from service and TS action statements not be entered for nonessential maintenanc PM be considered essential maintenance and should be scheduled to coincide with surveillance tests where possibl CM work which can be accomplished concurrently with PM work without significantly increasing out-of-service time be scheduled as suc Up coming entries into TS action statements be specifically addressed in planning meetings to maintain awareness and sensitivity to this issue. Management be notified when expected durations for LCO maintenance are exceeded, including PM' As much CM work as possible be performed during outages to eliminate unnecessary maintenance risks while the reactor is operatin The licensee also provided the inspectors with a memorandum sent to all superintendents by the Assistant Plant Manager, Safety and
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Licensing, dated November 9, 1994. This memorandum summarized the NRC's concerns that were addressed in the T The inspectors met with the Superintendent of Outage and Planning to ascertain what procedures or steps are tdken during the work planning stages to address the issues in the TI. The superintendent stated that the planned maintenance was on a 52 week cycl EDGs and safety related equipment such as_ the LHSI pumps maintenance was scheduled on a quarterly basi For example, No. 1 EDG PM would be scheduled for the first quarter and the other EDGs during succeeding quarters. The maintenance associated with that particular component is packaged to be completed at the same time when possible. The time to do the maintenance is evaluated and if it exceeds 50% of the allowed LCO time, SNSOC approval is required. The Superintendent of Outage and Planning also stated that trains of safety systems are not removed from service when a unit is operating. The inspectors concluded that the licensee's current practices for performing on-line maintenance were acceptable per TI 2515/12 Outage Scheduling and Work Control The licensee held detailed work status meetings twice per day and staffed an outage work control center 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per day throughout the present Unit 1 SG chemical cleaning outag The inspectors reviewed the outage schedule, attended status meetings, and observed work control center activities frequentl Planned work, emerging issues, and lessons learned from the previous Unit 2 SG cleaning outage were efficiently integrated into the outage schedul Outage planning management generally coordinated maintenance activities and outage schedule in an effective manner. A stuck open reactor coolant loop isolation valve (I-RC-MOV-I594)
necessitated significant rearrangement of planned work activitie Schedulers addressed emergent material problems such as I-RC-MOV-I594 promptly, with minimal impact on resources or outage duration. However, emergent maintenance activities associated with B LHSI pump breaker, AVS charcoal filter media testing, and planned maintenance associated with replacement of six IA station battery cells were not effectively coordinate The SS recognized the AVS conflict and postponed IA station battery maintenance to assure at least one AVS train remained operable at all time Operators also identified that the B LHSI pump was tagged out for breaker maintenance at the same time as the IA station battery and backup power supply to the A LHSI pump were inoperable. Although permitted by TS, this was inconsistent with VPAP-2805, Shutdown Risk Progra This condition existed for only a short period of time before the SS recognized the conflict and restored the B LHSI pump to service. The licensee identified both of these occurrences and initiated corrective action to determine why the schedule conflicts were not identified at an earlier stag *
The inspectors observed strong management involvement in reviewing the DRs at the morning meeting and identified no repeat occurrences during this inspection perio Both schedule conflicts were identified by the licensee and neither resulted in a significant reduction to safet Station management clearly emphasized personnel and plant safety at the daily work status meeting Personnel were sp~cifically reminded that Unit 2 remained on-line, and to be alert to ensure Unit 1 outage activities did not adversely challenge Unit 2 operation. The inspectors concluded that station management production perspective was well focused and did not permit outage schedule to interfere with safe maintenance and plant operatio Within the areas inspected, no violations or deviations were identifie.
Surveillance Inspections (61726)
During the reporting period, the inspectors reviewed the following surveillance activity to assure compliance with the appropriate procedure and TS requirement Unit 1 Containment Isolation Valve Local Leak Rate Testing On November 29, the inspectors observed the testing of Unit 1 containment isolation valves 1-DA-TV-lOOA and These valves were tested in accordance with l-OPT-CT-201, Containment Isolation Valve Local Leak Rate Testing (Type C Containment Testing), revision The purpose of this test was to obtain the as-found leak rates of 1-DA-TV-lOOA and B following the failure of 1-DA-TV-lOOA to fully close (see paragraph 3.4).
The inspectors attended the prejob brief, witnessed the performance of the test in the Unit 1 containment, verified that the test equipment was calibrated and that procedures were followed and also reviewed the test data. The leak rate through 1-DA-TV-lOOA exceeded the band of the flow meter utilized for the test. This confirmed that the valve was stuck in the mid positio DA-TV-lOOA was scheduled to be repaired prior to starting up Unit 1 following the outag The leakage rate through 1-DA-TV-lOOB was within acceptance criteria specified by the procedur The inspectors concluded that this testing was satisfactorily accomplished in accordance with l-OPT-CT-20 Within the areas inspected, no violations or deviations were identifie.
Onsite Engineering Review (37551) RCS Loop Isolation Valve Stem Integrity On November 30, the C RCS loop hot leg isolation valve (l-RC-MOV-1594) failed in the open position (see paragraph 3;2).
Maintenance personnel monitored stem thrust and motor current as
- operators repeated their attempts to cycle the valve both electrically and manuall The maximum thrust applied to the valve stem during these activities was 150,000 pound Due to the large size of the valve body, radiography of the valve to confirm integrity and position was impractical. Engineers, maintenance personnel, and vendor representatives discussed their observations and concluded that the valve disc was most likely wedged between the outer surface of the seat ring and the backsea The valve disk remained completely out of the RCS loop flow pat~.
The licensee developed a contingency valve repair plan which required the reactor vessel to be drained to reduced inventor Licensee management recognized that with fuel in the vessel, plant conditions for this repair inherently reduced overall safety margin from core damag Therefore management requested engineering to perform a SE to evaluate the option of leaving l-RC-MOV-1594 in its current condition and scheduling the repair for the next RFO in the fall of 199 The inspectors determined that evaluation of this option demonstrated a good safety perspective and was appropriat Valve components may have been weakened due to stresses _exerted during the unsuccessful attempts to close the valv The inspector expressed concern that the stem may have been weakened sufficiently to cause the disk to separate from the stem during power operation. Engineers prepared SE 94-208, Operation of Unit 1 with l-RC-MOV-1594 Stuck in the Open Position, dated December 6, to evaluate l-RC-MOV-1594 valve integrity during plant heatup and power operation with the valve disc stuck in its current position until the next RF Ultrasonic testing of the valve confirmed that the valve stem had not been damaged or deforme The inspectors reviewed SE 94-208 and vendor documentation of allowable thrust for the various valve component The weak link was the trunnion pin which can accept over double the 150,000 pound thrust applied when attempting to close the valv The SE addressed valve component material composition, allowable thrust, heatup and cooldown stresses, and assessments of two previous stem-disc disk separations on similar valve The licensee determined it was safe to leave 1-RC-MOV 1594 backseated in the open position until the next refueling outage. The inspectors concluded that the engineering evaluation of the reactor coolant loop isolation valve was comprehensiv.2 Containment Purge Ventilation Lineup Modification The licensee implemented a TM to the AVS to reduce the potential for chemical contamination of the filters during the Unit 1 SG cleaning outage. A jumper was installed to divert containment purge flow, which could contain potentially damaging chemical vapors, to the category II ventilation filter (l-VS-FL-14). This path bypassed safety related charcoal filters (l-VS-FL-3A and 38)
which had been damaged during the Unit 2 SG chemical cleaning
earlier in the yea The jumper was installed in accordance with l-TOP-4047, Installation of Temporary Flanges and Ventilation Duct Work to Permit Purging Unit 1 Containment Using the Auxiliary Building General Area Filter Exhaust Fan (l-VS-F-59) During Unit 1 SG Chemical Cleaning Outage, Revision The inspectors attended SNSOC review of the supporting SE (SE 94-198) and concluded that the modification was well thought out and reviewed in detai The inspectors walked down the installed modification*and observed operation of the containment purge system with the ventilation jumper in place. Procedure lOOP-VS-001, Containment Ventilation, revision 6-T2, was properly modified to incorporate operation with the jumpe The inspectors observed that operators were knowledgeable of the revised procedure and closely monitored containment purg The licensee was not able to specifically identify the source of the damaging chemicals from the Unit 2 filter contamination (see NRC Inspection Report Nos. 50-280/94-24 and 50-281/94-24).
Corrective actions resulting from that event included data collection regarding filter performance during this Unit 1 outag Charcoal filter samples were analyzed on l-VS-FL-3A, 38, and 14 prior to the start of Unit 1 SG chemical cleaning. This confirmed the as found condition of the filters. Station management directed that as left samples be obtained for all three filters after SG chemical cleaning is complete The l-VS-FL-3B filter was used briefly during the Unit 1 outage in support of fuel cask movement operations, but did not process containment purge flo On December 9, charcoal from filter l-VS-FL-3B was sampled and analyze The sample results verified that the methyl iodide removal rate was with acceptance criteria specified in TSs and that the charcoal media in l-VS-FL-3B was not degraded during the outag The inspectors determined that installation of the ventilation jumper and subsequent activities including charcoal filter analysis were a tangible and positive product of the licensee evaluation of the earlier ventilation filter contamination even Within the areas inspected, no violations or deviations were identifie.
Plant Support (71707, 71750)
The inspectors conducted facility tours, work activities observations, personnel interviews, and documentation.reviews to determine whether licensee programs were effectively implemented to comply with regulatory requirements in the areas of radiological protection, security, emergency preparedness, and fire protectio.1 Plant Tour Observations The inspectors observed radiological control practices and radiological conditions throughout the plant. Radiological
posting and control of contaminated areas was goo Workers complied with radiation work permits and appropriately used required personnel monitoring device Selected aspects of plant physical security were reviewed during regular and backshift hours to verify controls were in accordance with the security plan and implementing procedures. This review included security measures, vital and protected area barrier integrity, maintenance of isolation zones, personnel a*ccess control, searches of personnel, packages and vehicles, and escorting of visitor No discrepancies were note The inspectors observed security personnel control access to the Unit 1 containment while the equipment access hatch was removed to support various outage maintenance activities. Security personnel were alert and knowledgeable of their dutie On one occasion, the security access key card system malfunctioned due to moisture intrusion at outdoor key card readers and cable runs. Access alarms remained operable and doors remained locke However, key card readers did not process key entry requests and did not unlo*ck doors to controlled areas. Security force personnel promptly established compensatory measures to provide access to the protected area and to the control roo Additionally, vital area keys were available to the shift supervisor to assure operator access to perform assigned duties during security system failure The inspectors concluded that compensatory access control measures for key card system failures were acceptabl.2 Notification of Unusual Event At 9:13 p.m., on November 25, 1994, the licensee declared a NOUE in response to an explosion in the SR An unexpected rapid chemical reaction overpressurized a waste processing tan One person was injured and evacuated to a local hospital for treatmen No one was contaminate Personnel promptly verified that the overpressurization did not cause a fire, radioactive release, or damage to safety related system The licensee classification of this event as a NOUE was appropriat The emergency plan was properly implemented and the NOUE was terminated at 11:45 The licensee initiated a two part review of the even The first review was performed by the contractor who designed the SR This report postulated that hydrogen peroxide reacted with residual material in the bottom of the treatment tank including EDTA and copper sulfate to generate a large amount of oxygen gas. This theory was confirmed through analytical and experimental recreation of the reaction. The report concluded that the overpressurization event resulted from operating the waste processing system in a mode not covered by existing procedure Part two of the event review, performed by the Virginia Power corporate staff, was in progress at the close of this inspection period. Control of work activities which led to the
overpressurization remains under NRC review. This is identified as IFI 50-280, 281/94-31-02, SRF Overpressurization NOUE - Control of Work Activitie The inspectors reported to the station after being notifi~d of the event. The inspectors independently assessed the situation, reviewed the event declaration and verified the timeliness of the licensee's report Within the areas inspected, no violations or deviations were identifie.
Licensee Event Report Followup (92700)
The inspectors reviewed the LERs listed below and evaluated the adequacy of the corrective action~
The inspectors' review also included followup of the licensee's corrective action implementatio.1 (Closed) LER 50-280, 281/93-11, Radi~tion Monitors Inoperable Due to Detector Ground Referenc Gaseous vents system ventilation vent radiation monitors, l-VG-RM-131-1 and 2, were inoperable from, September 16 through October 9, 199 TS Table 3.7-6, Item 12 requires that an alternate method for monitoring the ventilation system for radiation be established and that a special report be submitted to the NRC if l-VG-RM-131-1 and 2 are inoperable for more than seven days. These radiation monitors were declared inoperable due to repetitive spiking. This spiking was not attributed to release of radioactive gases. Subsequent investigation revealed that spiking was caused by electrical disturbances in the syste The inspectors verified that EWR 93-057, Kaman Radiation Monitors Grounding Enhancement Installation, revision B, implemented hardware changes to prevent electrical disturbances. The inspectors also reviewed DRs for gaseous vents ventilation vent radiation monitors issued since the completion of EWR 93-057 and verified that spiking has not occurred since implementation of this modificatio The inspectors concluded that corrective actions implemented to prevent ventilation vent radiation monitors from spiking were effective in improving system operatio.2 (Closed) LER 50~280/94-07~ Process Vent High Range Accident Monitor Inoperable Greater Than Seven Day Process vent radiation monitors l-GW-RM-130-1 and 2 were inoperable from June 14 through August.IO, 199 TS Table 3.7.6, Item 11 requires that an alternate method for monitoring the process vent system for radiation be established and that a special report be submitted to the NRC if l-GW-RM-130-1 and 2 are inoperable for more than seven days. These radiation monitors were declared inoperable when system flow rate could not be maintained in the proper band. Accumulation of moisture in the iodine filter cartridge was determined to be restricting system flo The moisture was attributed to condensation of humid gas in the pipin As corrective action, heat tracing was installed on the
- radiation sample lines. The inspectors walked down the sample lines and verified that the heat tracing was installed. The inspectors also reviewed DRs initiated since August IO, 1994 and
- verified that the system has operated satisfactoril The inspectors concluded that corrective actions implemented to prevent moisture from accumulating in the process radiation monitors piping were effective in improving system operatio (Closed) LER 50-281/94-002, Both Trains of Auxiliary Ventilation Exhaust Inoperabl On March 7, 1994, AVEFT fan l-VS-F-58B tripped and was declared inoperabl When l-VS-F-58B tripped, the emergency power supply to AVEFT fan l-VS-F-58A was inoperable and therefore l-VS-F-58A was declared inoperabl TS 3.0.2 prohibits operation with l-VS-F-58A and B inoperable and a six-hour LCO to hot shutdown was entere The emergency power supply to l-VS-F-58A was returned to service, l-VS-F-58A was declared operable and the six-hour LCO was exited. Subsequent investigation revealed that l-VS-F-58B tripped because the manual damper in the suction line was not properly locked open and vibrated shu As corrective action procedures were revised to provide instructions to ensure that the manual dampers are properly locked open when operated to return the system to service following maintenanc The inspectors reviewed procedures O-MOP-VS-004 and 005, Return to Service of Auxiliary Ventilation Exhaust Train A and B, revision I, and verified that the procedures contained provisions for locking manual dampers ope Within the areas inspected, no violations or deviations were identifie.
Exit Interview The inspection scope and findings were summarized on December 13, 1994, with those persons indicated in paragraph The inspectors described the areas inspected and discussed in detail the inspection results addressed in the Summary section and those listed belo Item Number URI 50-281/94-31-01 IFI 50-280, 281/94-31-02 LER 50-280, 281/93-11 Status Open Open Closed Description/(Paragraph No.}
Containment Design Features (paragraph 3.3).
SRF Overpressurization NOLIE -
Control of Work Activities (paragraph 7.2).
Radiation Monitors Inoperable Due to Detector Ground Reference (paragraph 8.1).
r
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Item Number LER 50-280/94-07
Status Closed Description/(Paraqraph No.)
Process Vent High Range Accident Monitor Inoperable Greater Than Seven Days (paragraph 8.2).
LER 50-281/94-002 Closed Both Trains of Auxiliary Ventilation Exhaust Inoperable (paragraph 8.3).
Proprietary information is not contained in this report. Dissenting comments were not received from the license.
Index of Acronyms and Initialisms AFW AVEFT AVS CL CM DC DCP DR EOG EDTA ESF EWR F
HSD I&C IFI LER LCO LHSI LOCA MFW MS NA&F NOUE NPSH NRC NSS PSL PM RCP RCS RFO SE SG SGCC SI AUXILIARY FEEDWATER AUXILIARY VENTILATION EXHAUST FILTER TRAIN AUXILIARY VENTILATION SYSTEM CENTERLINE CORRECTIVE MAINTENANCE DESIGN CHANGE DESIGN CHANGE PACKAGE DEVIATION REPORT EMERGENCY DIESEL GENERATOR ETHYLENE DIAMINE TETRA-ACETIC ACID ENGINEERED SAFETY FEATURE ENGINEERING WORK REQUEST FAHRENHEIT HOT SHUTDOWN INSTRUMENTATION AND CALIBRATION INSPECTION FOLLOWUP ITEM LICENSEE EVENT REPORT LIMITING CONDITIONS OF OPERATION LOW HEAD SAFETY INJECTION LOSS OF COOLANT ACCIDENT MAIN FEEDWATER MAIN STEAM NUCLEAR ANALYSIS AND FUEL NOTICE OF UNUSUAL EVENT NET POSITIVE SUCTION HEAD NUCLEAR REGULATORY COMMISSION NUCLEAR SITE SERVICES PLANT STATUS LOG PREVENTATIVE MAINTENANCE REACTOR COOLANT PUMP REACTOR COOLANT SYSTEM REFUELING OUTAGE SAFETY EVALUATION STEAM GENERATOR STEAM GENERATOR CHEMICAL CLEANING SAFETY INJECTION
SNSOC STATION NUCLEAR SAFETY AND OPERATING COMMITTEE SRF SURRY RADWASTE FACILITY SS SHIFT SUPERVISOR TI TEMPORARY INSTRUCTION TM TEMPORARY MODIFICATION TS TECHNICAL SPECIFICATION URI UNRESOLVED ITEM VPAP VIRGINIA POWER ADMINISTRATIVE PROCEDURE WO WORK ORDER