IR 05000280/1988041
| ML18152B294 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 11/30/1988 |
| From: | Cantrell F, Holland W, Larry Nicholson NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18152B293 | List: |
| References | |
| 50-280-88-41, 50-281-88-41, IEIN-85-091, IEIN-85-91, NUDOCS 8812080162 | |
| Download: ML18152B294 (20) | |
Text
Report Nos.:
UNITED STATES NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTA STREET, N.W.
ATLANTA, GEORGIA 30323 50-280/88-41 and 50-281/88-41 Licensee:
Virginia Electric and Power Company Richmond, Virginia 23261 Docket Nos.:
50-280 and 50-281 Facility Name:
Surry 1 and 2 License Nos.:
DPR-32 and DPR-37 Inspection Conducted: October 2 through November 5, 1988 D'a te S'i gned
// /)t1 kfl bate Signed Accompanying Inspector:
P. Fillion Approved by: ~~
F: T.tantre 11,2A!-~ ~ on Chief D1v1s1on of Reactor ProJects SUMMARY bate -Signed Scope:
This routine resident inspection was conducted on site in the areas of licensee actions on previous enforcement matters, plant operations, plant maintenance, plant surveillance, licensee event report_review, and design changes and modification Results:
One apparent violation was identified (280,281/88-41-0l) for failure to take appropriate corrective actions for identified deficiencies was noted as follows:
Failure to promptly identify a deviation to the shift supervisor and prepare a deviation report on August 29, 1988, that potential gas binding may adversely effect the operability of the high head safety injection pumps {paragraph 5).
'Failure to adequately evaluate *the adverse condition documented in station deviation Sl-87-946 from November 20, 1987, to April 11, 1988, with regards to control room chiller capacity {paragraph 3).
Failure to identify the potential control room envelope ventilation problem*at the time information was available to question the capability of the system {paragraph 7).
Failure to take appropriate corrective actions for an NRC identified violation with regard to inventory of special nuclear material which was discussed in inspection report 280,281/87-10 (paragraph 3).
These examples listed above indicate a weakness in past i~plementa-tion of the corrective action program at the Surry Power Statio In addition,_ an inspector followup item was identified in paragraph 5 for followup on licensee evaluation of technical issues identified during review of loss of P_ressure Relief Tank (PRT) water event (280,281/88-41-02).
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- Person Contacted Licensee Employees REPORT DETAILS J. Bailey, Superintendent of Operations
- D. Benson, Station Manager
- R. Bilyeu, Licensing Engineer H. Blake, Superintendent of Site Services
- R. Blount, Superintendent of Technical Services
- E. Grecheck, Assistant Station Manager
- G. Miller, Licensing Coordinator, Surry H. Miller, Assistant Station Manager
- J. Ogren, Superintendent of Maintenance
- J. Price, Site Quality Assurance Manager S. Sarver, Superintendent of Health Physics
- Attended exit meetin Other licensee employees contacted included control room operators, shift technical advisors, shift supervisori and other plant personnel.
The NRC Region II Section Chief, F. Cantrell, visited the Surry Power Station on October 12, 13, and 14, 198 M Cantrell 1s tours included the low level intake structure, service water system walkdown, auxiliary building, and the Unit 2 containmen On October 18~ 1988, the NRC NRR Director for Project Directorate II-2, H. Berkow, visited the Surry Power Station to review the current status of issues and to tour the station. Mr. Berkow was accompanied by the current Surry Project Manager, C. Patel; and also was accompanied by the oncoming Surry Project Manager, B. Buckle In addition, Mr. W. Troskoski of the NRC Executive Director's Staff visited the Surry Power Station on the same day and was briefed by the Resident Staf On October 20, 1988, one of the Commissioners of the Nuclear Regulatory Commission, Kenneth M. Carr, visited the Surry Power Station for a familiarization tour, to meet with licensee management and staff, and to review current station statu Commissioner Carr was accompanied by the following personnel:
M. L. Ernst, Acting Regional Administrator M. Federline, Technical Assistant to the Commissioner B. Wilson, Branch Chief, DRP, Region II NRC Resident Inspectors The Commissioner attended the morning management status meeting; met with the resident inspectors; was given a presentation of several areas by licensee management; and was taken on a tour of the service water
system, battery and switchgear rooms, control room, emergency diesel generator rooms, and the independent spent fuel storage installation.* Plant Status Unit 1 Unit 1 began the reporting period at cold shutdown with preparations being made to defuel -the reactor in order to find and replace leaking fuel assemb 1 i e Fue 1 offload had commence However, at the end of the inspection period, one fuel assembly in location G-6 in the reactor vessel became stuck and efforts were continuing to free the assembly from the manipulator crane gripper when the period ende Unit 2 Unit 2 began the reporting period at cold shutdown in day 21 of a scheduled 81 day refueling/maintenance outag Installation of new recirculation spray heat exchangers was completed and preparations were being made to defuel the reactor when the inspection period ende.
Licensee Action on Previous Enforcement Matters (92702)
(Closed) Unresolved Item (URI) 280,281/88-12-01, Engineering Evaluation and Report of Control/Relay Room Chiller This item was identified in inspection report 280,281/88-12 during a region-based inspection of the performance of the control room and emergency switchgear room chiller syste The inspector noted that station deviation Sl-87-946, dated November 20, 1987, identified the fact that the subject chiller units did not meet the 90 ton capacity specified in the UFSA No formal evaluation or 10 CFR 50.59 review was performed until the inspector raised the issue during the week of April 11, 198 The subsequent analysis, dated April 19, 1988, justified continued operation as long as service water temperature remained below 70 degrees The 1 i censee has upgraded the chillers to meet the design capacit However, as discussed in paragraph 7 of this report, the entire system has since been determined to be degraded to the point of not being ab 1 e to meet performance specifictio The failure to adequately evaluate the adverse condition documented in station deviation Sl-87-946 is an additional example of violation 280,281/88-41-01 which is discussed in paragraphs 5 and 7 (control room ventillation) of this report. This unresolved item is therefore close (Closed) Unresolved item (URI) 280,281/88-18-01, Review of procedures for configuration control of piping blank This item was identified in inspection report 280,281/88-1 In that report the inspector had identified a concern with regards to control of piping blanks which are routinely removed by operations in order to establish temporary flowpaths for evolutions involving mid-nozzle operation during outage Since that time, the inspector was provided with the results of a review which was conducted at the direction of the operations superintendent in his are *
That review concluded that, for the most part, configuration control was being adequately maintained when piping blanks were removed to support operations evolution However, certain procedures did not adequately address removal and/or reinstallation-bf blank flanges to support operations or maintenance activitie Some of these procedures were:
Operations Procedure (OP)-19.2, Containment Vacuum System - Refueling Operations Maintenance Operating Procedure (MOP)-5.6, 5.7, 5.8; Reactor Coolant Sys tern Loop Fi 11 MOP-8.1, 8.4, 8.6; Return to Service of Charging Pump A, B, C MOP-14.1, 14.J; Remove RHR Pumps from Service MOP-14.3, 14.4; Return RHR Pumps to Service Corrective actions for the operations procedures were entered on the commitment tracking system by the Operations Superintendent on September 7, 198 The inspector reviewed the licensee actions in the operations area and determined that they were appropriat The inspector also reviewed the problem area with other department supervisory personnel and concluded that a similar problem does not appear to exist in other station area Technical Specification 6.4 requires that detailed.written procedures with appropriate check-off 1 i sts sha 11 be provided for norma 1 startup, shutdown, and operation of a unit and of all systems and components involving nuclear safety of the statio Failure to provide adequate procedure for installation or removal of piping blanks on safety-related systems is a violation of Technical Specification 6, The inspectors reviewed the findings and the corrective actions performed prior to the end of the inspection period and determined that they are acceptabl After discussion between the inspectors and NRC regional management, it was concluded that adequate corrective actions to prevent recurrence have been taken by the 1 i censee prior to the end of the inspection perio (Open)* Violation 280,281/87-10-0l, Failure to conduct an annual physical inventory for all special nuclear material. The violation was identified during a NRC inspection in May, 1987, and was discussed in inspection report 280,281/87-1 Licensee response, in part, to the violation on July 17, 1987, agreed that the violation was correct and stated that corrective steps which will be taken to avoid further violations included:
(1) preparation of a new procedure to perform the required inventories, (2) conduct of a physical inventory in accordance with the procedure on July 8, 1987, (3) conduct of the next physical inventory on September 30,
1987, and (4) conduct of future inventories on a semi-annual basis to be concurrent with the DOE reporting requirement During this inspection period, the inspector was informed that the 1 i censee commitment to perform i terns 3 and 4 above had not been accomplished as require This condition was identified during an audit by the licensee's quality assurance organizatio The failure to perform these requirements was identified in the licensee's corrective action program by a deviation report (Sl-88-1035) written on October 6, 198 This deviation report resulted in additional licensee review of the requirements of the regulations and concluded that their initial response to the violation was inadequate. Their conclusions were that (1) detector location was not clearly identified during past audits, (2) the most recently performed periodic audit procedures were signed off as complete, when in fact they were incomplete, and (3) the most recent deviated procedure which was prepared on October 6, 1988, to establish an accurate physical inventory of all detectors could not be accomplished due to personnel being unwilling to sign off verification steps based on memor The inspector reviewed the preceding conclusions with licensee management and was informed that they would be sending a revised response to the violation identified in inspection report 280,281/87-1 Failure to take appropriate corrective actions for an NRC identified violation is identified as an additional example of violation 280,281/88-41-01 which is discussed in paragraph 5 of this repor.
Unresolved Items Unresolved items are matters about which more information is required to determine.whether they are acceptable or may involve '{iolations or deviation No new unresolved items are identified in this inspection repor.
Plant Operations Operational Safety Verification (71707)
The inspectors conducted daily inspections in the following areas:
control room staffing, access, and operator behavior; operator adherence to approved procedures, technical specifications, and limiting conditions for operations; examination of panels containing instrumentation and other reactor protection system elements to determine that required channels are operable; arid review of control room operator 1 ogs, operating orders, plant deviation reports, tagout logs, jumper logs, and tags on components to verify compliance with approved procedure The inspectors conducted weekly inspections in the following areas:
verification of operability of selected Engineered Safety Feature (ESF)
systems by valve alignment, breaker positions, condition of equipment or component(s), and operability of instrumentation and support items essential to system actuation or performanc **
Pl ant tours included observation of genera 1 pl ant/equipment conditions, fire protection and preventative measures, control of activities in progress, radiation protection controls, physical security controls, plant housekeepirig conditions/cleanliness, and missile hazard The inspectors routinely monitor the* temperature of the auxi 1 i ary *
feedwater pump discharge piping to ensure steam binding is prevente The inspectors conducted biweekly inspections in the following areas:
verification review and walkdown of safety-related tagout(s) in effect; review of sampling program (e.g., primary and secondary coolant samples, boric acid tank samples, plant liquid and gaseous samples); observation of control room shift turnover; review of implementation of the plant problem identification system; verification of selected portions of containment isolation lineup(s); and verification that notices to workers are posted as required by 10 CFR 1 Gerta in tours were conducted on backshifts or weekend Backshift or weekend tours were conducted on October 3, 5, 6, 8, 9, 10, 11, 12, 13, 14, 15, 17, 18, 19, 20, 22, 23, 24, 29; and November 3, 4, and Inspections included areas in the Units 1 _and 2 cable vaults, vital battery rooms, steam safeguards areas, emergency switchgear rooms, diesel generator rooms, control room, auxiliary building, Units 1 and 2 containments, cable penetration areas, independent spent fuel storage facility, low level intake structure, and the safeguards valve pit and pump pit areas. Reactor coolant system leak rates were reviewed to ens*ure that detected or suspected leakage from the system was recorded, investigated, and e_valuated; and that appropriate actions were taken, if require The inspectors routinely independently calculated RCS leak rates using the NRC Independent Measurements Leak Rate Program (RCSLK9).
On a regular basis, radiation work permits (RWPs) were reviewed and specific work activities were monitored to assure they were being conducted per the RWP Selected radiation protection instruments were periodically checked, and equipment operability and calibration frequency were verifie In the course of monthly activities, the inspectors included a review of the licensee I s phys i ca 1 security progra The performance of various shifts of the security force was observed in the conduct of daily activities to include: protected and vital areas access controls; searching of personnel, packages and vehicles; badge issuance and retrieval; escorting of visitors; and patrols and compensatory post POTENTIAL FAILURE OF SPENT FUEL POOL DOOR SEAL The inspectors monitored the licensee actions regarding the discovery that a potential exists for a failure of the spent fuel pool door sea The licensee identifiea, via station deviation report S1~88-1027, that during certain evolutions the only barrier that prevents emptying the spent fuel inventory into containment is an inflatable seal around the transfer canal doo Prior to moving activated fuel, the licensee opens the fuel transfer tube gate* valve and moves a dummy fuel assembly through the
containment fuel pool penetratio It is during this evolution that the water in the spent fuel pool is retained only by an inflated seal around the transfer canal doo A loss of instrument air that inflates the door seal could also render the transfer system inoperable and prevent closing the transfer tube gate valv The licensee was in the process of testing the fuel transfer system when an operator noted an air leak in the air supply to the inflatable door sea Investigation revealed a spot light had been placed adjacent to the hose from the air system regulator to the seal, and it had burnt a small hole in the airhos The airhose was temporarily repaired using tape, and the gate valve was closed until a permanent repair could be performe The door *Seal remained inflated during this event with no loss of water past the sea However, at the same time this condition occurred, an operator was preparing to enter the refueling canal to aid in positioning the dummy fuel assembl In addition, as documented in station deviation Sl-88-1012, the fuel transfer system conveyor air motor became mechanically boun This air motor is supplied from the same instrument air supply as the door sea A work order was issued and the motor was subsequently repaire After repair, the checkout evolutions involving transfer of the dummy assembly were completed.
The above event caused the senior reactor operator in charge of refueling to evaluate the scenario and submit a station deviation identifying the overall concer Given the worst case of a fuel transfer system failure with the dummy assembly stuck in the tube, coupled with a failure of the door seal to retain the spent fuel pool water, the spent fuel pool would drain to within 13 inches of the top of the fuel assemblies. This water would drain inside containment and flood the basement since the reactor cavity seal is not normally installed during this evolutio The licensee mandated that the fuel transfer gate valve remain closed until an investigation into the event could be performe This investigation, dated October 10, ~988, concluded the following:
Loss of instrument air to the seal could result in a significant leakrate into the spent fuel building transfer canal and containmen The gate valve (22 inches diameter) could be closed.during canal door seal leakage, if no obstructions blocked the closur With a total loss of air and the transfer cart protruding through the transfer tube, it may not be possible to reinflate the seal or isolate the transfer tube (cJose the gate valve).
It would take approximately 85 minutes to drain to the top of the weir (approximately 13 inches over the fuel assemblies) with no makeup.
Since fuel pool cooling would be lost, maximum heatup rate would be
- less than 4.5 degrees F/hour, which would allow approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> before bulR pool boiling would occu **
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Assuming 13 inches of shielding above the top of the fuel assemblies, a dose rate of approximately 50 R/hour at the edge of the spent fuel pool would be expecte The report recommendations included a requirement that the reactor cavity seal assembly be installed prior *to opening the transfer tube gate valve for testing the* conveyor syste This would limit a drain down of the spent fuel pool to approximately 14 feet over the top of the spent fuel racks, thus providing adequate shielding to perform recovery operations in the fuel building.
. The inspectors continued to review the licensee corrective actions regarding this postulated scenari An action plan, dated October 12, 1988, was developed by the licensee to implement both short term and long
- range corrective action Identification of the potential problem by the refueling SRO indicates a increased sensitivity to safety issues that is commendabl The extensive investigation into this event after a-station deviation was submitted was also commendabl However, it is should be noted that the failure of the reactor cavity seal on May 17, 1988, had not resulted in a generic review of similar seal configurations at the station until the spent fuel pool seal potential problem was highlighted by the above occurrence.
LEAKAGE OF PRT WATER INSIDE CONTAINMENT The inspectors investigated the circumstances that resulted.in approximately 250 gallons of water leaking from a pressurizer safety valve flange on Unit This event occurred on October 4, 1988, and was identified in station deviation report S2-88-52 The licensee had been venting the reactor for several days by maintaining the reactor coolant system (RCS) at mid-nozzle and degassing through an empty pressurizer, into the pressure relief tank (PRT), and out the process ventilation syste The pressurizer code safety valves were removed and a temporary cleanlines~ cover (]asket material) was placed over the flange opening The pressurizer power-operated relief valves were being maintained open providing a vent path from the pressurizer to the PR _The event was initiated when the control room operator attempted to vent and depressurize the safety injection accumulators utilizing operating _
procedure 2-0P-7.7.4, Venting Safety Injection Accumulator Step 5.1.3:4 of the above procedure opens valve HCV-2936 and vents the accumul~tors into the process vent syste It was at this point that the control -room operator noticed fluctuations in the RCS standpipe level and an increase in PRT pressur The PRT level was noted to decrease from 13 to 10 percent*
during the event, thus translating to a loss of inventory from the PRT of approximately 250 gallon The operator secured the accumulator vent and noted that the RCS standpipe level indication returned to norma It was also reported that water was flowing from the pressurizer safety valve flanges that would indicate that the PRT water was being displaced back toward the pressurizer through the discharge lines from the primary relief valves which had been removed for maintenanc **
The inspector reviewed this event with the station staff performing an
.investigation into the inciden The event team also researched the possibility of losing RCS inventory out the incore thimble guide tubes that are disconnected at the seal table with their low pressure seals installed. This scenario was deemed to be the worst case condition since it was determined that enough pressure could develop in the reactor.
coolant system to blow the low pressure seals bn the thimble guide tubes and therefore constitute a cold leg loss of inventor The above event signifies the complexity of problems when using a shared system, such as the process vent system, to perform concurrent function Th~ licensee event team identified several weaknesses with procedures, system interfaces, valve lineups, and level indication It was also recommended that an alternate method of venting the accumulators, such as*
to the containment atmosphere, be evaluate The licensee reported this event via the INPO Network System on October 13, 198 The inspector reviewed the results of the investigation and agreed that significant questions remain to be answered regarding the impact of this even These questions include possible overpressurization of piping, inaccurate level indication, inaccurate sparger location, and the effects of the boric acid that was spilled on adjacent component The licensee was researching the above items when the inspection period ende Therefore, this will be identified as an inspector followup item for followup on licensee evaluation of technical* issues identified during review of loss of PRT water event (280; 281/88-41-02).
INSIDE RECIRCULATION SPRAY PUMP DEGRADATION - UNIT 2 On October 7, 1988, the licensee made a 10 CFR 50. 72 ca 11 to the NRC informing us of degradation which has been discovered during disassembly and inspections of the Unit 2 inside recirculation spray pump The disassembly and inspections were scheduled during the current Unit 2 outage due to similar degradation of the same pumps on Unit 1 which was identified in June 198 Additional followup of this area is discussed in paragraph 6 of this -repor *
POTENTIAL GAS BINDING OF HIGH PRESSURE SAFETY INJECTION PUMPS On October 12, 1988, the licensee reported, pursuant to the requirements of 10 CFR 50.72, that an evaluation had identified a gas accumulation in the suction of the high pressure safety injection pump (HPSI) that could possibly gas bind the pump during a loss-of-coolant acciden The report further stated that vents will be installed on the high points during the present outage The cause of this was attributed to a design deficiency of the system.
The resident inspectors, accompanied by their section chief, toured the piping spaces and noted the* high points in questio The licensee discovered this situation during a review in response to NRC Information Notice IN 88-23, dated May 12, 198 The internal response to this IN was
assigned to a system engineer who requested that t"he site nondestructive test group ultrasonically inspect selected high points for absence of wa'te The results of the ultrasonic testing were compiled and transmitted to the station engineer o~ August 23, 1988, indicating actual voids did exist in the piping; The engineer concluded in a memorandum to his supervision, dated August 29, 1988, that 11the operability of the HPSI pumps during an emergency is in question 11 *
. The above memorandum stated that the worst known pipe voiding at the time of the examination was in the area of Unit 2 LCV-21158 and LCV-2115D..
These valves open to provide a flowpath from the refueling water storage tank (RWST) to the suction of the HPSI pump The approximate volume o this void was stated to be 3.8 cubic feet, thus considerably greater than the 2.2 cubic feet that could be:s~fely passed as stated in a tel~phone conversation by the pump vendor (Byron Jackson).
The memorandum went on to state that the installation* of vents is recommended to ensure that the HPSI pumps are provided with a full suction in all situation The exact size of the gas voids cannot be determined from the data obtaine The engineer stated that he suspended any further examinations as soon as he was convinced that the pump~ could not safely pass the known amount of ga The licensee has decided not to pursue a determination of an acceptable amount of voiding and instead is proceeding with the installation -0f high-point vent The question of pump operability was raised on August 29 in the engineer's memo and copies were addressed to several tiers of supervision. No deviation report was prepared by the engineer when he identified the potential problem; nor did any of the supervision identify the problem in the licensee's corrective action progra It was not until the work to add the vents was discussed in a scheduling meeting at the assistant station management level on October 12, that the significance of the issue was recognize CFR 50, Appendix B, Criterion XVI, states that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and correcte In the case of significant conditions adverse to quality, the measures shall assure that the cause of the condition is determined, and corrective action taken to preclude repetitio The identification of significant conditions adverse to quality, the cause of the condition, and the corrective action taken shall. be documented and reported to the appropriate levels of managemen These requirements are implemented by the licensee's QA Topical Report VEP 1-S That report states, in part, that adverse conditions significant to quality, the cause of the conditions, and the corrective action taken are reported to appropriate levels of both offsite and onsite management by the use of a deviation repor The topical report also requires review of each deviation report for reportabi l i,ty of the condition to the NR The topical report
requirements are implemented at the station by Surry Power Station Administrative Procedure SUADM-0-12, Operations Department Notification This procedure defines a deviation as a significant difference between the expected value or conditions and the actual value or conditio It further requires that the shift supervisor shall be informed of all deviations, or nonconformances which may be deviations and also requires the individual identifying the deviation to complete a deviation repor The failure to promptly identify a deviation to the shift supervisor and prepare a deviation report on August 29, 1988, that potential gas binding may adversely effect the operability of the high head safety injection pumps is identified as a violation (280,281/88-41-0l).
Station* management expressed their concern that the above item was identified on August 29, 1988, and that it was not properly evaluated at that tim Unit 1 continued to operate until September 14, when it was shutdown for EOG concerns, and Unit 2 operated until the refueling outage ihat began September 1 This problem could potentially have resulted in all the high head safety injection pumps being unable to perform as required during an acciden The inspectors routinely review each station deviation report and have noted a large increase in the numbers of reports submitte The station deviation report submitted on the spent fuel pool door does indicate an increased awareness to identify and act on safety concern The licensee has taken some interim corrective actions, such as reading each station deviation report in the daily management meetin This process will continue until a formal program can be developed to adequately identify and evaluate safety concern This station has historically addressed discrepancies and concerns in a somewhat informal manner, with no formal mechanism in place, for example, to generate written justification for continued operations (JCOs).
The inspectors are continuing to evaluate the effectiveness of a program to identify and evaluate safety concerns as they aris CONTAINMENT SPRAY NOZZLE BLOCKAGE On October 19, 1988, a station deviation report (Sl-88-1157) was submitted identifying several spray nozzles in the Unit 1 containment that were covered with tape (tape wrapped around the nozzle).
The shift technical advisor subsequently performed an inspection of all the spray nozzles and documented the results via a memorandum dated October 19, 198 The results identified eight nozzles covered with tape and several nozzles oriented incorrectl The inspectors discussed the results of this inspection with the station staff and monitored the station evaluation and corrective actions.
STATION TAGGING PROGRAM On October 24, the inspectors discussed with the superintendent of operations the findings and reply to a liGensee quality assurance audit*
(Report S88-24) regarding the station tagging progra Based on concerns from a previous INPO audit, management requested the Quality Assurance
{QA) audit that was conducted during the Spring 1988, Unit 1 refueling outag The QA auHit report identified a ~otal of nine findings and ten
- observations relating to failures to follow procedures, inadequate procedures, and a lack of attention to detai The specific details of this audit were presented to the inspectors by the Quality Assurance Manage The findings and responses seem to indicate a tagging system that is basically sound, but confusing at times in that it does not adequately handle the abnormal tagging situations that arise during a major outag The concept of using blanket tagouts to isolate entire systems and include all the work under this single tagging order was introduced prior to the Spring 1988, refueling outag Although the audit results did not identify major problems with tagging and isolations in the field, it is evident that some specific details are needed in the area of blanket tagouts to correct the confusio The licensee agreed with this observation and stated that plans are being implemented to install a computerized tagging system that would improve on the method of establishing isolatio The inspectors will continue to monitor the 1 icensee actions on this subject as part of the routine inspection progra Within the areas inspected, one apparent violation was identifie.
Maintenance Inspections (62703)
During the reporting period, the inspectors reviewed maintenance activities to assure compliance with the appropriate procedure INSPECTION AND REPAIR OF INSIDE RECIRCULATION SPRAY PUMPS 2-RS-P-lA & B During this inspection period, the inspectors monitored the work associated with the disassembly, inspections, and repair of the Unit 2
- nside recirculation spray (IRS) pumps IA and 1 The inspectors observed selected disassembly evolutions of the pump casings in the shop and noted the following
2-RS-P-lA had indication of damage due to wear ring rotation of the casing wear ring and the first stage impeller upper wear ring between the fixed ring support area and the rings, respectivel Also, the first stage impeller could not be removed from the shaft as designed and showed rota ti ona 1 movement between the shaft keyway and the impeller keywa Additional internal parts including shaft sleeve snap rings, impeller lock collar bolts, and lock wire were either loose or missin *
2-RS-P-lB had indication of damage due to wear ring rotation of the casing wear ring between its fixed ring support area and the rin Also, the pump.had damage associated with failure of a stabilizing bearing sleev Parts of the sleeve appeared to have passed through the pump internals during some period prior to disassembl Some parts were also found in the pump operating sump after the pump had been removed from its normal locatio The licensee 1s inspection results generally agreed with that of the inspector The pumps were disassembled in the presence of a field service representative from the pump vendor, Bingha This representative concluded that although the pumps were degraded, they were still operable and capable of producing a pressure and flow; however, no estimate was provided with regard as to how long the pump(s) would run in their 11as found 11 conditio The inspectors will continue to monitor pump repairs and testing as part of the regular inspection progra Within the areas inspected, no violations or deviations were identifie.
Surveillance Inspections (61726)
Dufing the reporting period, the inspectors reviewed various surveillance activities to assure compliance with the appropriate procedures as follows:
Test prerequisites were me Tests were performed in accordance with approved procedure Test procedures appeared to perform their intended functio Adequate coordination existed among personnel involved in the tes Test data was properly collected and recorde Inspection areas included the following:
TESTING OF THE UNIT 1 REACTOR CAVITY SEAL On* October 15, 1988, the inspectors witnessed testing of the J-Seal portion of the reactor cavity seal assembly for Unit The test was being conducted in accordance with Spec i a 1 Test ST -224, Operability Reactor Cavity J-seals dated October 11, 198 The purpose of the test was to verify the ability of the reactor cavity J-seals to perform their intended function by comparison of actual leakage with expected values.
This test was.accomplished by installing the reactor cavity seal assembly in the Unit 1 containment, flooding the cavity to different specified levels above the reactor vessel flange, deflating the inflatable *seal at
. p
the specified levels, and monitoring for leakage past the J-seal Potential leakage rates were specified in the procedure past the J-seal with maximum leakage anticipated at 26'- 611 to be approximately 150 gp The inspectors reviewed a copy of the test procedure, the radiation work permit associated with_ the test, and attended the pretest briefing on the morning of October 1 The inspector witnessed the initial raising of cavity level from inside the Unit 1 containment and independently verified leakage from the J-seals to be less than 0.05 gallons per minute at the first test point (1'- 611 ).
The inspector then exited containment and observed filling of the cavity to the 16'- 011 level from the control roo The inspector noted that the licensee was still having communications problems due to having to use hand held radios in the noisy containment while wearing respirator The inspector continued the monitoring of the test from the control room until J-seal leakage at the second test point (16'- 011 ) was determined.. That leak rate was determined to be less than 0.10 gallons per minut The test was completed on October 17, when the leakage* past the J-seal at the third test point (26' - 611 ) was determined to be less that 0.30 gallo~s per minut After completion of testing and approval of the test results by the station safety committee, the inspectors reviewed the completed test procedur Several small discrepancies were noted and identified to the license However, the inspectors consider that the completed test procedure did adequately document testing of the cavity sea CONTROL ROOM & EMERGENCY SWITCHGEAR ROOM VENTILATION The inspectors followed testing and evaluation of the ventilation that provides cooling to the control room and emergency switchgear room (ESR).
The licensee documented via station d~viation report (Sl-88-937} dated September 9, 1988, that this ventilation system can no longer maintain normal design room temperatures when operating -in the designed configuratio Previous operating experience has proven that the operation of two chillers and both trains of air handling units has been
- required to maintain acceptable room temperature The system was designed to maintain the control room at 75 degrees F and 50% relative humidity during either normal or emergency conditions, with the ESR maintained at 80 degrees F and 40% relative humidity during normal operations, and 87 degrees F and 35% relative humidity during emergency operation The system consists of three chiller units with three service water pumps (1-VS-P-lA, B & C) that supply cooling to each chiller condense Each chiller has a chilled water pump {1-VS-P-2A, B & C) that circulates chilled water to a set of air handling unit The air handling units utilizes a fan to pass air over the chilled water coils and out a system of ductwor A normal train consists of a chiller providing water to one of two sets of air handling units. The third chiller functions as a swing chiller to supply either train. Each train in itself should be capable of maintaining design temperature **
The station deviation report referenced above concluded that the ability of the existing equipment to perform its intended function is indeterminate at this time based on the following:
The actual chiller capacities were significantly less than the_
design/procurement documentatio This issue was addressed in inspection report 88-12 as unresolved item 88-12-0 The upgrade of the chiller motors to meet original design capacities was completed
- in May 198 The emergency switchgear room heat loads have increased by the installation of new electrical equipment over the year It is apparent that the design change process did not track and account for the additional of heat load The material condition of the Heat,. Ventilation/Air Conditioning (HVAC) equipment has significantly degraded since original installa-tio *
The licensee performed special test ST-220, Control Room Envelope Air Conditioning System, to record data for determining the ability of the system to perform as designe Prior to performing this test, the air handling units were cleaned to the maximum extent possibl The filters in these units have traditionally been routinely replaced, but the fans and.cooling coils were found to be extremely dirt The construction of these uni ts precludes easy access to the fans and coi 1 s for routine cleanin The service water valves that modulate and short-cycles service water through the chiller condensers were removed and the piping blanked to maximize chi 11 er performanc The 1 i censee stated that these va 1 ves were known problems and would be repaired prior to declaring the system fully operab 1 The results of the above test were sti 11 under development as the inspection period ende A preliminary evaluation of the raw data was performed by the resident inspector and confirmed the licensee initial findings that the chilled water flow was less than the design value and the air flow through the air handlers did not meet design criteria *. The inspector considers the collection of data to be adequat On November 2, 1988, the licensee's evaluation of the test data had progressed to a point that it was determined that the capacity of the chiller system is inadequate to maintain design room temperatures during a loss of coolant accident with a loss of offsite powe In addition, the non-safety related central chiller system, which is to be used as a backup during certain high energy line break and Appe.ndix 11R11 scenarios, appears to be inadequate to perform as required. This information was reported to the NRC as a 4-hour, non-emergency call~
The resident inspectors attended a briefing of station management near the end of the inspection perio In that meeting it became apparent that the capability of the control room envelope ventilation system to perform its
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design function as specified in the UFSAR was questionable as far back as two years ag However, the information that the Architect/Engineer (A/E)
provided with regard to calculated base heat load was unrealistic which resulted in the present licensee actions to confirm actual system loadin This process did not include identification of the suspect condition by the station corrective action program (writing of a deviation report)
until September 9, 198 Discussions with engineering personnel involved in the ventilation upgrade program indicated that the recent sensitivity to documentation of deviations after the'Unit 1 reactor cavity seal event resulted in the preparation of the deviation repor The failure to identify the control room envelope ventilation potential problem at the time information was available to question the capability of the system is an additional example of violation 280,281/88-41-0 The resident inspectors will continue to monitor the effort to correct this problem since it is identified as a requirement for unit restar EMERGENCY DIESEL GENERATOR SPECIAL TESTING On October 22, the inspector witnessed testing of the No. 2 Emergency Diesel Generator (EOG) in accordance with Special Test ST-225, EDG(S) Load Reject Tes tin£.
The purpose of the test was to evaluate the emergency diesel engine governor transient response capability by instantaneously reducing load on the generato The test required that the No. 2 EOG be 1 oaded to approximately 2. 75 MWe on the 2H emergency bus and then disconnect the load from the EOG by opening its output breaker to the emergency bu The test was conducted satisfactorily and resulted in minima 1 change in speed when tbe output breaker was opene No discrepancies were note On October 24 and 25, the inspectors witnessed testing of the number 3 EOG in accordance with Special Test ST~227, Emergency Diesel Generator (EOG)
NO. 3 Load Sequence Tes The purpose -of the test was to obtain the voltage and frequency response of the isolated emergency diesel generator subjected to load profiles which bound worst case scenario The results of these tests will be analyzed to determine the required number of load blocks and the times between each load block to ensure that the diesel can start and accelerate all load The design considerations are addressed in paragraph 9 of this repor The test required that the associated emergency bus be loaded with the required pumps and resistive load bank to simulate the initial load block to which the EOG would be subjecte The offsite source breaker to the bus is then opened, and the EOG breaker will connect the EOG to the bus if the EOG is already running; or, if the EOG is not running, the diesel will start, obtain rated speed and voltage, and will then connect to the bu Additional loads will be added to the bus as directed in the test to evaluate their effect on the diese The inspectors witnessed both types of test No discrepancies were noted.
Within the areas inspected,.one additional example of a violation identified in paragraph 5 was noted regarding the adequacy of control room ventilation syste.
Licensee Event Report (LER) Review (92700)
The inspectors reviewed the LER 1s listed below to ascertain whether NRC reporting requirements were being met and to determine appropriateness of the corrective action(s). The inspector's review also included followup on implementation of corrective action and review of licensee documentation that all required corrective action(s) were complet (Closed)
LER 280/88-05, Inoperable Heat Tracing Due to Inadequate Procedure The issue involved identification on inoperable technical specification heat trace circuits due to inadequate su-rvei 11 ance and maintenance procedure The concern of operability of the heat trace was initially identified by the NRC resident inspectors and resulted in a violatio Corrective actions included extensive maintenance and calibration of the heat trace syste Additional corrective actions included operator walkdowns of the associated local panels each shif The inspectors verified that corrective actions were implemente This LER is close.
Design Changes and Modifications (37700)
During this inspection period, the inspectors selected several design change packages which were being implemented for revie These ch~nges were:
EMERGENCY DIESEL GENERATOR MODIFICATIONS On October 11-14, 1988, an inspector from the Region II Office was on site to review the purpose of and procedure for special tests to be conducted on the emergency diesel generator While reviewing Information Notice -
85-91, Load Sequencers for Emergency Diesel Generators, the licensee identified that a problem similar to that described in the Notice-existed at Surry Power Statio At Surry, should a loss of offsite power (LOOP)
occur before or simultaneously with a loss of coolant accident (LOCA),
accident mitigation equipment will be sequenced onto the diesel generators (D/G) in blocks that can easily be accepted by the D/ The resulting consequences is that limiting safeguards or safety injection signals clear the emergency bus and initiate load sequencin However, should a LOOP occur after a LOCA, the bus is not cleare When the D/G output breaker closes onto the bus, the D/G - sees an instantaneous load change of greater magnitude than it is designed to handl This conclusion is based on the 12.5 Mega Vol ts/Amps (MVA) ultimate step increase limit given on the manufacturer's Dead Load Capacity Curve ACD 6Z-4 Motor start MVA-values at full voltage given on the motor data sheets were summed and the total was compared with the 12.5 limit. Motor start Mega Watt (MW) values were determined by multiplying the start MVA by the start power factor (from
.the motor data sheets).
Since this calculation showed that the 12.5 MVA limit was exceeded, it was postulated that both D/Gs on the accident unit may fail for the LOOP after LOCA scenari A study was performed by the licensee 1s Nuclear Engineering Department (NED)* to define the problem and propose solution After this study was complete, a team of engineers from Stone & Webster Engineering Corporation were brought to the NED office in Innsbrook; Virginia, to independently review the problem definition and proposed solutio They also re-established the relevant origina*1 design basi Meetings with engineers from the generator vendor, Morrison-Knudsen Company, were held to help resolve the proble Engineers from other utilities were also consulte Unit 1 was shutdown in September 1988, as a result of the identified proble Unit No. 2 had been shutdown earlier in September 1988, for a refueling outag Essentially the proposed solution was to install timing relays to sequence blocks of load onto the D/G for the LOOP after LOCA scenario. The timing sequence for this scenario must be faster than for the other scenarios in order to meet the accident analysis constraint An objective was to make the timing sequence ~s fast as possible to obtain the best possible margin of safet *
The NRC inspector reviewed the 70 percent complete draft modification package at the sit The fi.nal package will include a complete safety evaluation wherein the proposed sequencing scheme (to be validated by test) is shown to be consistent with any accident analysi Computer codes will be utilized in making this determinatio The present sequencing scheme for the LOOP before LOCA scenario will not chang The purpose of the diesel generator tests will be to demonstrate the D/G 1s ability to accept relatively large instantaneous load increases while maintaining acceptable voltage levels, thus validating the proposed sequencing schem The licensee believes that the ultimate short time Kilo Watt (KW) ability of the D/Gs may be limited by the turbo charger performance in the first few minutes of operatio The turbo charger may not achi.eve maximum efficiency with the relatively cool exhaust gases in the first few minutes of operation. The 11Cold Load Capability 11 test will define this power limiting effect in terms of magnitude and time duratio It is expected that the effect will limit the D/G output to a level below the 30 minute rating of 2950 K After the exhaust air warms up the D/G output will increase; and the D/G will be tripped by the operator at 3000 K The 11Transient 11 test will demonstrate the D/G's ability to handle the proposed load sequence schem The criteria for this test is that the motors wi 11 accelerate within the safe start times and that motor starters, or other relays, do not drop out as a result of the voltage di A transient analyzer recorder will be* utilized to acc;ept and process variables to be monitore Key output variables will be profiles of generator output voltage and curren Also monitored will be frequency, KW, KVAR, and exciter termi na 1 voltag Obviously, motor con troll er cente~ contactor status must be observe The inspectors observed the 11 before modification 11 testing of the No. 2 and No. 3 emergency diesel generator These test observations are discussed in paragraph 7 of this repor In conclusion, the licensee has accurately defined the D/G loading problem, and has taken a complete and appropriate approach to its solutio The test procedure is valid, and should achieve the objectiv If the test results are as predicted, the test will validate the proposed loading sequence scheme to be installed during the present outage The inspector's comments as described herein were relayed to the licensee's management in an exit interview conducted on October 14, 198 Within the areas inspected, no violations or deviations were identifie.
Exit Interview The inspection scope and findings were summarized on November 9, 1988, with those individuals identified by an asterisk in paragraph The follo_wing new items were identified by the inspectors during' this exi One apparent violation (280,281/88-41-0l) with four examples for failure to take appropriate corrective actions for identified deficiencies was noted as follows:
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Fai1ure to promptly identify a deviation to the shift supervisor and prepare a deviation report on August 29, 1988, that potential gas binding may adversely effect the operability of the high head safety injection pumps (paragraph 5).
Failure to adequately evaluate the adverse condition documented in station deviation Sl-87-946 from November 20, 1987, to April 11, 1988, with regard to control room chiller capacity (paragraph 3).
Failure to identify the control room envelope ventilation potential problem at the time information was available to question the capability of the system (paragraph 7).
Failure to take appropriate corrective actions for a NRC identified violation with regard to inventory of special nuclear material which was discussed in inspection report 280,281/87-10 (paragraph 3).
The apparent example violations listed above indicates a weakness in past implementation of the licensee's corrective action -program at the Surry Power Statio One inspector followup item was identified in paragraph 5 for followup on licensee evaluation of technical issues identified during review of loss of PRT water event (280,281/88-41-02).
The 1 i censee acknowledged the inspection findings with no dissenting comment The licensee did not identify as proprietary any of the materials provided to or reviewed by the inspectors during this inspection.