IR 05000280/1987013
| ML18150A184 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 06/16/1987 |
| From: | Cantrell F, Holland W, Larry Nicholson NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18150A183 | List: |
| References | |
| 50-280-87-13, 50-281-87-13, NUDOCS 8706260128 | |
| Download: ML18150A184 (10) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTA STREET, ATLANTA, GEORGIA 30323 Report Nos.: *50-280/87-13 and 50-281/87-13 Licensee:
Virginia Electric and Power Company Richmond, VA 23261 Docket Nos.:
50-280 and 50-281 License Nos.: DPR-32 and DPR-37 Facility Name:
su*rry 1 and 2 Inspection Conducted:
May 3 -
1987 Inspectors: ;-;-,,~~~~.;;<~;.:::**,z::;~-~-~~~~i:...::::~---=-:-~~-;-~----:-~~~~~~
Accompanying Inspector: S. G. Tingen Approved by:
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Date Sfgned SUMMARY Scope: This routine inspection was conducted in the areas of plant operations, plant maintenance, plant surveillance, followup on inspector identified items,and licensee event report revie Results: No violations or deviations were identified in this inspection report *
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- Persons Contacted Licensee Employees REPORT DETAILS R. F. Saunders, Station Manager
- 0. L. Benson, Assistant Station Manager
- H. L. Miller, Assistant Station Manager
- E. S. Grecheck, Assistant Station Manager
- J. A. Bailey, Superintendent of Operations D. J. Burke, Superintendent of Maintenance S. P. Sarver, Superintendent of Health Physics
- R. H. Blount, Acting Superintendent of Technical Services R. L. Johnson, Operations Supervisor
- J. A. Price, Site Quality Assurance Manager W. D. Craft, Licensing Coordinator J. B. Logan, Supervisor, Safety and Licensing
- Attended exit meeting.
Other licensee employees contacted included control room operators, shift technical advisors, shift supervisors and other plant personne The NRC Region II Section Chief, Floyd S. Cantrell, visited the station on May 6 and 28, 198.
Exit Interview The inspection scope and findings were summarized on June 9, 1987, with those individuals identified by an asterisk in paragraph The following new items were identified by the inspectors during this exi One unresolved item (paragraph 6) was identified for reviewing the licensee justification for backseating loop stop valves during normal operation (280; 281/87-13-01).
One inspector followup item (paragraph 5) was identified to inspect the licensee performance in removing decay heat during low reactor coolant water level operations (280; 281/87-13-02).
The licensee acknowledged the inspection findings with no desenting comment The 1 icensee * did not identify as proprietary any of the materials provided to or reviewed by the inspectors during this inspectio *
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2 Licensee Action on Previous Enforcement Matters (92702)
This subject was not addressed in the inspectio.
Unresolved Items*
Unresolved items are matters about which more information is required to determine weather they are acceptable or may involve violations or deviation One new unresolved item is identified in paragraph.
Plant Operations Operational Safety Verification (71707)
The inspector 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; review of control room operator logs, operating orders, plant deviation reports, tagout logs, jumper logs, and tags on components to verify compliance with approved procedure The inspector conducted weekly inspections in the following areas:
Verification of operability of selected ESF systems by valve alignment, breaker positions, condition of equipment or component(s), and operability of instrumentation and support stems essential to system actuation or performanc *
Plant tours which included observation of general plant/equipment conditions, fire protection and preventative measures, control of activities in progress, radiation protection controls, physical security controls, plant housekeeping conditions/cleanliness, and missile hazard The inspector 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 valve lineup(s); and verification that notices to workers are posted as required by 10 CFR 1 Certain tours were conducted* on backshifts or weekend Backs hi ft or weekend tours were conducted on May 9, 11, 16, 20, 26, 28, 29 & 3 Inspections included areas in the Units 1 and 2.cable vaults, vital battery rooms, steam safeguards areas, emergency switchgear rooms, diesel
- An Unresolved Item is a matter about which more information is required to determine whether it is acceptable or may involve a violation or deviation.
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generator rooms, control room, auxiliary building, cable penetration areas, independent spent fuel storage facility, low level intake structure, Unit 1 containment, and safeguards valve pit and pump pit areas. Reactor coolant system leak rates were reviewed to ensure that detected or suspected leakage from the system was recorded, investigated, and evaluated 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 l icensee 1 s physical 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 Unit 1 began the reporting period at powe The unit reduced power to approximately 30 % on May 9 to allow for containment entries in order to clean boric acid out of the control rod drive mechanism (CROM) cooler The cleaning was required due to buildup of boric acid in these coolers during past operation with 1 eakage from the reactor head vent va 1 ve tailpipe. This blockage was preventing proper operation of the cooling system and a 11 owing containment average temperature to increase where technical specification limits may have been approached during the Summer month This condition may have resulted in unit shutdown when peak power demand was required if not correcte The inspector questioned the licensee about the status of the boric acid leakage in conjunction with NRC Information Notice 86-108, Supplement 1 (Degradation of Reactor Coolant Pressure Boundary Resulting From Boric Acid Corrosion).
The licensee explained that during the timeframe that leakage was occurring from the reactor head vent valve, all of.the boric acid which precipitated out of solution was being sucked into the Control Rod Drive Mechanism ventilation system. This condition resulted in the buildup of boric acid on the cooler The licensee also evaluated the boric acid buildup condition based on the information provided in NRC Notice 86-108 and coricluded that no degradation of Reactor Coolant Pressure Boundary Components had occurre This conclusion was based on inspection during the reduced power entrie This conclusion was reviewed by the station safety committee on May 11, 1987, and approved *by the committe The inspector reviewed the approved minutes with regards to the boric acid issu The unit returned to full power operati6n on May 1 Unit 1 operated at full power until May 16, when at 0824 hours0.00954 days <br />0.229 hours <br />0.00136 weeks <br />3.13532e-4 months <br /> a reactor trip from full power (low flow - A loop) occurre The trip was a result of partial loss of flow on the A reactor coolant loop due to partial closing of the loop A hot leg isolation valve. All systems performed as designed during and after the trip. The unit was cooled down and reached
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- cold shutdown at 2017 hours0.0233 days <br />0.56 hours <br />0.00333 weeks <br />7.674685e-4 months <br /> on May 17, 198 Repairs were accomplished on the A hot leg isolation valve (MOV-1590 - see paragraph 6) and other minor work was performe The licensee also conducted a more thorough inspection for boric acid on reactor coolant system pressure boundary components during this outage due to NRC information notice 86-106, Supplement This inspection involved removal of insulation panels from the vessel head in 9rder to inspect for boric acid accumulation. Approximately 80 pounds of boric acid was found on the head in the vicinity of the pad eye support lifting collar at a point directly below the head vent line. The licensee removed the boric acid residue and determined that no degradation of the head, hold down studs, or nuts had occurre The inspector also reviewed the licensee 1s pictures taken before and after boric acid cleanup in addition to making containment entries to evaluate the conditio The condition was also discussed with region management on May 26, 198 The unit commenced heatup above 200 degrees Fon May 28 and was critical on May 29, 198 The unit recommenced power operation on May 30 and operated at power for the remainder of the inspection perio The inspectors reviewed the licencee 1 s evaluation of problems encountered with the residual heat removal pumps (RHR) during the May forced outage of unit The RHR system was placed in service and pump 1-RH-P-lA was started on May 17,198 Periodic Test Procedure 1-PT-30.1, 11 RHR System Operability 11, was performed on May 21 and determined that the developed head, Delta P, for pump 1-RH-P-lA was 108.6 psig. This value corresponds to an alert condition as defined by the above procedure. The lA pump was subsequently secured and pump 1-RH-P-18 started. The 1B pump was then tested using the same procedure and determined to be inoperable due to low developed head. An Engineering Work Request (EWR 87-225) was written and evaluated to de*c1 are 1-RH-P-lA fully operable and 1-RH-P-18 operable but in an alert condition. The 18 pump was retested on May 23 and a revision to the EWR was issued declaring this pump fully operabl The licensee concluded that 'instrument error and back leakage through check valves resulted in the above conditio The inspectors discussed with station management the importance of having a dependable method of decay heat removal in conjunction with an accurate means of measuring reactor vessel water level as detailed in NRC Information Notice 87-23:
11 Loss of Decay Heat Removal Ouri.ng Low Reactor Coolant Level Operation 11 *
The licensee assessment of this notice and performance in this area is identified as an inspector followup item (280; 281/87-13-02).
Unit 2 began the reporting period at powe The unit operated at power*
for the duration of the inspection perio Engineered Safety Feature.System Walkdown (71710)
The inspector performed a walkdown of* the accessible areas of t:he containment vacuum leakage and monitoring system for both units to verify its operabilit This verification included the following: confirmation that the licensee'.s system lineup procedure matches plant drawings and
actual plant configuration; hangers and supports are operable;. house-keeping is adequate; valves and/or breakers in the system are installed correctly and appear to be operable; fire protection/prevention is adequate; major system components are properly labeled and appear to be operable; instrumentation is properly installed, calibrated and functioning; and valves and/or breakers are in correct position as required by plant procedure and unit statu Within the areas inspected, no violations or deviations were identifie.
Maintenance Inspections (62703)
During the reporting period, the inspectors reviewed maintenance activities to assure compliance with the appropriate procedure Inspections areas included the following:
Repair to Loop A Hot Leg Isolation Valve (MDV - 1590)
On May 16, 1987, the Unit 1 reactor tripped due to MDV -
1590 inadvertent closur The unit was cooled down and corrective maintenance was performed on the MD The work was accomplished using Mechanical Corrective Maintenance Procedure MMP-C-RC-105 (30
Darling Loop Stop Valves Disassembly, Repair, Reassembly "Safety Related").
The inspector reviewed the completed work order (Job Number 3800053467) and also visited the job site in containment while the work was being accomplishe No discrepancies were identifie The inspector did note that the valve procedure required several deviations (changes) in order to remove the internals due to the stem break at the backsea After the failure mechanism of the valve stem was identified, the inspector was informed that a similar failure occurred on one of the Unit 1 loop stop isolation valves in 197 The failure mechanism was fully evaluated by Westinghouse, and a failure report "Surry Unit No. 1 Reactor Coolant Loop Isolation Valve Stem Failure Report" dated March 7, 1974, was prepare In that report the failure mechanism was identified as a high strain -
low cycle failure resulting from a severe notch at the steam colla The report further recommended that the subject valves should not be electrically backseated on torque; and that should backseating become necessary dur{ng maintenance, it should be accomplished manually with minimum applied load and without exceeding the springback deflection specified in the revised instruction manua The inspector then reviewed the technical manual "Instruction Manual Motor Operated Reactor Coolant 30 11 Loop Stop Valves for Reactor Coolant System Westinghouse WNES 546-CAK-70497B Darling Valve S. 0. E-5004 11 *
The manual stated in.a caution that manual backseating is permissible only to the extent that the open deflector indicator reading does not exceed 1/16
maximum, and that manual backseating may be used only when the packing needs replacemen The inspector then reviewed the Surry Power Operating Procedure 1-0P-18,
"Containment Checklist".
In that procedure the inspector noted that step 22 torqued the subject valves on their
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backseats to 1/16
deflection while the unit was in cold shutdown prior to startu The same procedure reverified torque of the subject va 1 ves to 1/16 11 deflection on their backseats when the unit reached hot shutdow The inspector questioned the licensee as to why the valves were backseated during startup and was informed that the backseat operation minimized stem leakage during operatio The inspector then requested that the licensee provide ~dditional information to justify backseating of the loop stop valves during normal operatio This item is unresolved pending review of the licensee 1s reply (280; 281/87-13-01).
Within the areas inspected, no violations or deviations were identifie.
Surveillance Inspections (61726, 61700)
During the reporting period, the inspectors reviewed various surveillance activities to assure compliance* with the appropriate procedures as fo 11 ows:
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 followin~:
Emergency Diesel Generator Operability On May 5, 1987, the inspector witnessed surveillance testing of the No. 3 Emergency Diesel Generator per test procedure 1-PT-22.3C, 11 Diesel Generator No. 3 Test 11 * This monthly test demonstrates that the emergency diesel generator wil 1 respond promptly and properly to a manual start, synchronization, and assumption of load as required by technical specifi-cation 4.6. The licensee also performed an overspeed trip test per procedure EE-EDG-M/Al, 11 Emergency D/G Engine One year Service &
Inspection 11 *
No discrepancies were note Electrical Penetration Leakage Test On May 7, 1987, the inspector witnessed portions of* surveillance test 2-PT-34, 11 Electrical Penetration Leakage Test 11 * This test records the as found pressure in the e 1 ectri ca 1 penetrations and recharges those penetrations as required. A five minute drop test is required for*
penetrations requiring repressurization. No discrepancies were noted.
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Containment Isolation Trip Valve Test On May 11, 1987, the inspector witnessed the quarterly testing of miscellaneous containment trip valves per periodic test procedure 1-PT-18.68. This test cycles the containment trip valves that are not operated as part of other tests to their required positions for an accident and records the closing time. The inspector verified that pro bl ems encountered during this test were adequately documented and evaluated. No discrepancies were note Low Head Safety Injection Pump Tes The inspectors performed an extensive review of periodic test procedure 1-PJ-18.1, 11 LO Head SI Test & Flushing Of Sensitized Stainless Steel Pip.ing 11 * This inspection included a review of all test results and station deviations generated since 1985 regarding the low head safety injection pumps SI-P-lA & 8 for both units. On May 11, 1987, the inspector witnessed the performance of the above test on both unit 1 pumps. Although no specific discrepancies were noted the inspector commented that increased management attention was needed to improve the housekeeping in the safeguards pump roo *
Turbine Driven auxiliary Feedwater (AFW) Pump On June 2,1987 the inspector witnessed testing of the turbine driven AFW pump 1-FW-P-2 per deviated periodic test procedure 1-PT-15.lC. This special test was mandated as a result of the station safety committee concern with water in the steam lines to the AFW turbine that resulted in three consecutive overspeed trips prior to the two successful runs that were used to qeclare the pump operable for unit 1 restart on May 30, 198 Excessive water was previously noted in the main steam lines and documented via station deviation report Sl-87-446. The inspector noted the following observations to station management: While preparing for the pump start, the operator drained approximately one ga 11 on of water from the steam 1 i ne downstream of the steam admission valves. While this practice may be recommended for equipment protection, it may also mask the suspected problem of overspeed trip from steamline moisture. Although the procedure did not call for draining the steamline, the operator indicated it to be a general practice used prior to all turbine driven AFW pump run The three overspeed trips mentioned above* all occurred using the train 118 11 steam admission valve SOV-MS-1028. The retest on June 2 used the train 11A 11 valve SOV-MS-102A. The inspector noted that from a steaml i ne moisture concern, the use of the 11 811 train va 1 ve would constitute worst case system configuratio :.. :*;.***>;.***r*
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8 The inspector noted a general lack of understanding among the operators involved in the test with regards to the purpose of the test. Proper briefings with all parties involved could possibly have precluded the above problem Station management noted the above comments and subsequently repeated the test on June 3, 1987. The pump ran successfully using the 118 11 train valve with no overspeed trip. The licensee also instituted periodic draining and quantifying of water from the steamline and is further evaluating possible steam drain modifications. The inspectors will continue to monitor this item during subsequent inspection Inside Recirculation Spray Pumps The inspector reviewed results of Periodic Test 1-PT-17.2, 11 Containment Inside Recirculation Spray Pumps 11 *
Technical Specification 4.5 requires all inside recirculation spray pumps to be dry tested at least once per month. The test is considered satisfactory if the motor and pump shaft rotates, starts on signal, and exhibits the correct ammeter readings. The inspector noted that the Apri 1 test for pump 1-RS-P-lA was somewhat inconclusive in that the shaft rotation light on the main control board did not illumi~ate as expected. Although motor amperage should indicate shaft rotation, no other pump performance indication is available. This item was discussed with the licensee and was subsequently corrected during the forced outage for unit Within the areas inspected, no violations or deviations were identifie.
Followup on Inspector Identified Items (92701)
(Closed) Inspector Followup Item (IFI) 280; 281-T2500/16, IE Information Notice No. 85-45 informed the licensee of a potentially generic problem involving seismic interactions within the movable flux mapping system at Westinghouse Plants. The licensee was requested to review the information for applicability and consider actions, if appropriate, to preclude a similar problem from occurring at their facilit Inspection of Unit 1 identified the Thimble Support Frame as the only area of concer The Frame itself was found to be sufficiently rigid; however, the connecting studs at the wa 11 attachments were retorqued to a 11fi nger-t i ght 11 condition in order to ensure even loading of the floor supporting channel assemblie Inspection of Unit 2 identified the Thimble Storage Frame and overhead power cable trays and conduit as the only areas of concer The cable tray and conduit was found to be seismically adequate; however, a lateral brace was installed on the Thimble Storage Frame. The inspector has reviewed the applicable documentation and considers that the licensee has taken appropriate action to assure that the seal table and flux mapping system would not be endangered by falling equipment/structures during a seismic even This item is closed.
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(Closed)
IFI 280; 281/86-40-02, Followup on feed and steam flow difference The issue involved thermal power calculations utilizing the computer program TPDWR2 which is described in NUREG-116 This program calculates thermal power higher that the licensee's thermal power calculation The major difference between these calculations was that the calculation done by the inspector using TPDWR2 used feedwater flow as the mass input; whereas, the licensee used steam flow as the mass inpu Subsequent review of the licensee's program as outlined in the Surry Power Station Secondary Plant Performance Evaluation dated August, 1984, revealed that the Surry Units 1 and 2 feedwater flow is up to 2 % greater than steam flo The resulting study conclusion was that steam flow should be used for thermal power calculation Based on the inspectors review of this study, this issue is close.
Licensee Event Report (LER) Review-(92700)
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The inspector reviewed the LERs 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 280/86-22, Engineered Safety Feature Relay Failure The issue involved failure of a reactor protection relay on two separate occasions resulting in a partial Train B Engineered Safety Feature actuatio The cause of both failures was determined to be a failure of the relay coil, apparently from overheatin Corrective action included replacement of the failed coils and subsequent testing of the new component The licensee is also conducting additional studies to minimize this type of failur The inspector reviewed corrective action and is also tracking additional corrective actions as an open inspector followup ite This item is close (Closed) LER 280/86-23, Containment Sump Trip Valv The issue involved testing of the inside containment isolation valve which identified excessive leakag Immediate corrective action included manual isolation of the valv Additional corrective action included identification that the leakage flowpath was the valve packing and subsequent tightning of the packing follower corrected this conditio In addition, the system had a check valve and the control logic modified to minimize cycling of the trip valve during the feedwater piping outage. The inspector reviewed the LER and verified that the modification was installed during the outag This item is close *.... "*<. ~::' ;-*--:::
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