ML20062G394
| ML20062G394 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 11/19/1990 |
| From: | Carroll R, Prevatte R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20062G388 | List: |
| References | |
| 50-324-90-41, 50-325-90-41, NUDOCS 9011290162 | |
| Download: ML20062G394 (17) | |
See also: IR 05000324/1990041
Text
{{#Wiki_filter:. . .. , , . [[pft Rip; ' REGION 11 UNITED STATES o NUCLEAR nEGULATORY COMMISslON 101 MARIETTA STREET, N.W. g t 'r ATLANT A, GEORGI A 30323
h i o
- .*
Report Nos.: 50-325/90-41 and 50-324/90-41 Licensee: Carolina Power and Light Company P. O. Box 1551 Raleigh, NC 27602 Docket Nos.: 50-325 and 50-324 License Nos. DPR-71 and DPR-62 Facility Name: Brunswick 1 and 2 Inspection Conducted: October 2 - November 4, 1990 Lead Inspector: . IdxM[M I ///O/fC R. L. Prevat C ~ g g Date Signed Other Inspectors: W. Levis D. J. Nelson /I[/$ !1o h= Approved Bh: . E. C/orroll, Acting Section Chief , Date 5figned Reactor Projects Branch 1 Division of Reactor Projects SUMMARY Scope: This routine safety inspection by the resident inspectors involved the areas of maintenance observation, surveillance observation, operational safety verifica - tion,-initial response to onsite events, onsite review committee, onsite , followup of. licensee event reports, review of 10 CFR Part 21 items, and action , on previous inspection findings. Results:- In the areas inspected, no programmatic weaknesses or significant safety matters were identified. -A non-cited violation for the failure to place a channel of the reactor protection' system scram discharge volume water level high trip system in the trip condition after exceeding the Technical Specification two hour time limit' for having this equipment in test was identified, paragraph 7.a. Two minor deficiencies involving the lack of attention to detail by the contrml room operators were noted.during routine plant tours, paragraph 4. ,,o 1 1 ;;,o i c. o0111L ' 4 p g p, ADO (X 0500Qp s O
- _ _ _ _ _ _ _ _ _ - . , 4 i. - . ; ,, ..- , .. 2 .
- A-review of the unit trip-and degraded voltage event that occurred on
September- 27,1990,' identified four items where the inspector did not have sufficient information and/or. the necessary resources to fully evaluate. These . " - items will be referred to NRR for further review, paragraph 7.c. i Unit 1 was in a refueling outage during the reporting period. Unit 2 experi- ' encedian automatic trip on October 12 as the result of a blown fuse in-the '- feedwater level' control' system, paragraph 5. The unit was restarted on 0ctober- 18,:1990. The~ licensee appeared very conservative-in their approach to unit restart and reduced- power twice during equipment malfunctions _ that may have placed the unit at risk. I A . -> < . < + I ' 1 -> b -! f E t ,
. -
,
__ -_
e ..- , , 0 . REPORT DETAILS , 1. Persons Contacted Licensee Employees
- K. Altman, Manager - Regulatory Compliance
F. Blackmon, Manager - Radwaste/ Fire Protection S. Callis, On-Site Licensing Engineer T. Cantebury, Manager -- Unit 1 Mechanical Maintenance
- G. Cheatham, Manager - Environmental-& Radiation Control
M. Ciemnicki, Security R. Creech, Manager - Unit 2 I&C Maintenance- J. Cribb, Manager - Quality Control (QC)
- W. Dorman, Manager - Quality Assurance (QA)/(QC)
- M.-Foss, Supervisor - Regulatory-Compliance
- V. Grouse, Employee Relations 'J. Harness, General' Manager - Brunswick Steam Electric Plant ' W ' Hatcher, Supervisor - Security R. Helme, Manager - Technical Support J.-Holder, Manager-OutageManagement& Modifications:(OM&M) '*M. Jones,. Manager - On-Site Nuclear Safety - BSEP ' R. Kitchen, Manager:- Unit 2 Mechanical' Maintenance t*B. Leonard, Manager - Training
- J. Leviner, Manager - Engineering Projects
J.E McKee, Manager' -QA ', (*J. Moyer, Technical. Assistant to Plant. General' Manager ,
- P. Musser, Manager - Maintenance Staff
n B.'Poteat,? Administrative: Assistant to Plant General Manager . R.-Poulk,-Manager -. License Training- ' *J. Simon, Manager L- Operations Unit 1 . '> ' W. Simpson,-Manager - Site Planning and Control- S. Smith; Manager -Unit 1 1&C Maintenance. . - ' R.L Starkey, Vice President -' Brunswick Nuclear Project R. Tart,-Manager ,0perations; Unit-2 . , J.'Titrington,. Manager ,0perations, Staff o
- R. Warden, Manager
Maintenance B. Wilson,: Manager - Nuclear Systems Engineering -l + L0ther licensee employees contacted included construction craftsmen, ' o ' h
- engineers, . technicians, operators,toffice personnel, and security force
+ members.-
- Attended:the' exit interview'
g ' Acronyms:and initialisms used in the report are listed in the last: paragraph. -
o 1 -l
. . . . , .* , 2 2. Maintenance Observation (62703) The inspectors observed maintenance activities, interviewed personnel, and reviewed records to verify that work was conducted in accordance with approved procedures, Technical Specifications, and applicable industry codes and standards. The inspectors also verified that: redundant components were operable; administrative controls were followed; tagouts were adequate; personnel were qualified; correct replacement parts were used; radiological controls were proper; fire protection was adequate; quality control hold points were adequate and observed; adequate post-maintenance testing was performed; and independent verification requirements were implemented. The inspectors independently verified that selected equipment was properly returned to service. , Outstanding work requests were reviewed to ensure that the licensee gave priority to safety-related maintenance. The inspectors observed / reviewed portions of the following maintenance activities: 90-ALUC1 2C RBCCW Pump Rebuild 90-ASYH1 MSL Rad Monitor B Power Supply Changeout 90-PME411 Route on 2-CAC-1262 90-PZI395- Sample Pump Replacement for 2-CAC-4409 90-WFI414 Rosemount ATTU Output Voltage Check Violations and deviations were not identified. 3. SurveillanceObservation(61726) The inspectors observed surveillance testing required by Technical Speci- fications. .Through observation, interviews, and record review, the .. inspectors verified that: tests conformed to Technical Specification requirements; administrative controls were followed; personnel were qualified; instrumentation was-calibrated; and data was accurate and complete. The inspectors independently verified selected test results and proper return to service of equipment. , The inspectors witnessed / reviewed portions of-the following test activities: 1-MST-IRM12W IRM Channels B, D, F, H Functional Test 1-MST-SRM22R SRM Channels A and C Channel Calibration PT-7.1.1.a Core Spray Injection Check Valve Operability Test PT-20.7.2 1-E11-F050B LLRT Violations and deviations were not identified. O
. . . .. ,, ..- 3 4. Operational Safety Verification (71707) -The inspectors verified that Unit 1 and Unit 2 were operated in compliance with Technical Specifications and other regulatory requirements by direct observations of activities, facility tours, discussions with personnel, reviewing of records and independent verification of safety system status. The inspectors verified that control room manning requirements of 10 CFR 50.54 and the Technical Specifications were met. Control operator, shift sup?rvisor, clearance, STA, daily and standing instructions, and jumper / bypass logs were reviewed to obtain information concerni .g operating trends and out of service safety systems to ensure that there were no conflicts with Technical Specification Limiting Conditions for Operations. Direct observations of control room panels and instrumentation and recorder traces important to safety were conducted to verify operability and that - operating parameters were within Technical Specification limits. The inspectors observed shif t turnovers to verify that system status continuity was maintained. The inspectors verified the status of selected control- room annunciators. Operability of a selected Engineered Safety Feature division was verified weekly by ensuring that: each accessible valve in the flow path was in its correct position; each power supply and breaker was closed for components that must activate upon initiation signal; the RHR subsystem cross-tie valve for each unit was closed with the power removed from the valve operator; there was no leakage of major components; there was proper lubrication and cooling-water.available; and conditions did not exist which could prevent fulfillment of the. system's functional requirements. Instrumentation essential to system actuation or performance was verified operable by observing on-scale indication and proper instrument valve lineup, if accessible. The inspectors verified that the licensee's health physics policies / procedures were'followed. This included observation of HP , practices and a review of area. surveys, radiation work permits, postings, and instrument calibration. The inspectors verified by general observations that: the secarity ' organization was. properly manned and security personnel were capable of performing- their assigned functions; persons and packages were checked prior to entry into the PA; vehicles were properly authorized, searched and escorted within the PA; persons within the PA displand photo identi- fication badges; personnel in vital areas were authorized; effective compensatory measures were employed when required; and security's response to threats or alarms was adequate.. I Ine inspectors also observed plant housekeering controls, verified position of certain containment isolation ' alves, checked clearances, ard , verified the operability of onsite and of' site emergency power sources. ..
- . .. ,, ' ... 4 An inspector found two minor control room deficiencies on Unit 2 not discovered by operators. On October 19, 1990, at approximately 7:30 a.m., the control switch key was found inserted in the control switch for Core Spray Valve F001A, Su)pression Pool Suction Valve. Important component control switches on tie control board are equipped with locks that require keys for operation to preeent inadvertent manipulation. Dummy keys are normally inserted with the color coded real keys attached. The valve was in its correct position, but had been manipulated for the performance of a surveillance test on the previous shift. Following the test, the key was not switched back. Prior to the inspector's discovery, three control operators and one trainee had performed their board walkdown without detecting the error. The inspector informed the operators of the problem and the key was promptly switched. On October 24, 1990, the inspector observed that one of six APRM GAFs indicated 0.00 on the 7:00 a.m. process computer P-1 printout. Normal GAF values are 0.98 to 1.00. The printout also flagged APRM B as a failed sensor. The process computer calculates the GAF from a ratio of calculated power and APRM indicated power. The 0.00 value observed by the inspector indicates that a meaningful GAF could not be calculated due to a problem with calculated or indicated power. The inspector reviewed previous hourly P-1s, which also indicated GAFs of 0.00 back to 3:00 a.m., which had a 4.17 value. When questioned, the on-duty-operators were unaware of and could not explain the discrepancy. Et 'ntually the operators determined that a surveillance test prior to 3:00 a.m. for APRM Channel B caused the computer to cease its " scan" of APRM B due to inconsistent values calcu- lated when the APRM output was manipulated during the test. The APRM-was placed back in service following the surveillance test, but the computer " scan" was not reset. This did not affect the operability of the APRM, and no other indications in the control room were affected. GAF values are the most consistent indication of APRM reliability. Although the APRM trend chart recorders would also indicate an actual APRM problem, the recorders cannot trend all six APRM channels simultaneously. Therefore, it is important for operators to monitor the GAFs. These two examples, while not safety significant, indicate that continued diligence by operators-is needed in monitoring control room indications. Violations and deviations were not identified. 5. Initial Response to Onsite Events (93702) Unit 2-Scram Unit 2 was operating at 100 percent power on October 12, 1990, when a fuse blew in the FWLCS resulting in a loss of power to a . number of components in the control circuit. The loss of power gave the appearance of low reactor water level which caused an increased demand signal to the RFPs and a trip' signal to the reactor recirculation pump runback circuits. The l
' ' .. .o , 5 "A" RFP also locked up due to the loss of power resulting in the "B" RFP responding to the master controller demand for increased feed flow. Twenty-six seconds after the fuse blew an actual high water level condition was reached which caused a turbine trip and reactor scram. The lowest reactor water level reached was approximately 117 inches. Group 2, partial Group 3 and Group 6 isolations were received. Reactor recirculation pumps tripped due to the low level condition. RCIC auto started and injected and HPCI auto started but did not inject due to the short duration of the level transient. HPCI was subsequently used to raise water level to the normal band. During post trip recovery, the operators had difficulty in placing a RFP in service. Because of the loss of power to the FWLC circuit, the master controller was still demanding 100 percent output. When 2B RFP was placed back in service, in automatic, the pump increased speed to 5700 RPM and discharge pressure increased to 1700 psig. With the flowpath isolated due to feed pump trip recovery actions, the 4 and 5 feedwater heater relief valves lifted and began releasing steam into the feedwater heater room. When steam was reported as being released in these rooms, the feedwater heater inlet valves were shut securing the leak. The FWLC system was also placed in single element and selected to channel B, which then restored level feedback to the controller.
- The fuse that blew in the' FWLC circuit was a Gould Shawmut A25Z? fuse and
was similar to the fuse which blew in this circuit on August 16, 1990, that also resulted in a reactor scram. The licensee has subsequently replaced the Gould Shawmut fuses in the FWLC circuitry with Bussman MIN fuses which are the type installed in Unit 1. The Gould Shawmut fuses were installed as an Appendix R modification to provide separation between safe shutdown circuits and other associated circuits of concern. Subsequent review by the licensee documented in EER 90-0262, October 14, -1990, determined that separation of this circuit was not required. . The inspector will review the licensee's corrective actions taken with respect to.the feed pump operation and the fuse failure when the LER is . issued. 6. Onsite Review Committee-(40700) The inspectors attended selected Plant Nuclear Safety Conrittee meetings conducted during the period. The inspectors verified that the meetings were conducted in accordance with Technical Specification requirements regarding quorum membership, review process, frequency, and personnel qualifications. Meeting minutes-were reviewed to confirm that decisions /recomendations were reflected in the minutes and followup of- corrective actions was completed. Violations and deviations were not identified. ,
_ - - . - - . . _ _ _ _ _ . . . . . . . . . . . . . . . _ . . . . . .. ,, . . - 6 7. Onsite Followup of Licensee Eveat Reports (92700) The below listed LERs were reviewed to verify that the information provided met NRC reporting requirements. The verification included adequacy of event description and corrective action taken or planned, existence of potential generic problems, and the relative safety significance of the event. Onsite inspections were performed and concluded that necessary corrective actions have been taken in accordance with existing requirements, license conditions, and commitments, unless otherwise stated, a. (Closed) LER 1-90-16, Operation Prohibited by Plant Technical Specifications During SDV Maintenance and Surveillance Activities. On September 17, 1990, the Unit 1 SF authorized the performance of IMST-RPS27R, RPS Scram Discharge Volume High Water level Channel Functional Test and Channel Calibration. In conjunction with the test, the SF also authorized work to be performed in accordance with the instructions of WR/JO 90-AMCT1, which replaced the electronic printed circuit board for level switch 1-C11-LSH-4516C. The SF entered a tracking LC0 for this work believing that only one input to channel A2 would be disabled. Subsequent review of this work on September 20, 1990, by a different shift foreman, revealed that the MST disables the trip function of channel A2 during a portion of the test. A two hour time limit is allowed by Technical Specification 3.3.1 for a channel to be disabled during surveillance testing provided that the other channel in the same trip system is operable. After the two hour period, the channel must be returned to operable status or placed in the tripped condition. Because of the maintenance actions performed on 1-C11-LSH-4516C, the two hour time limit was exceeded during performance of the MST on September 17, 1990. .The SF did not place the channel in the tripped condition, as required by Technical Specification 3.3.1, because he did not realize that the A2 channel was disabled during the testing. The channel was inope aole for 2 hours and 20 minutes, in their investigation..the licensee noted several contributing conditions. First, it is not appropriate to perform corrective maintenance while performing surveillance tests. The two hour time period allowed by Technical Specifications for ' inoperable channels applies. to surveillance testing and not corrective maintenance. A Standing Instruction was put in place until permanent procedure revisions are made that prohibit corrective maintenance to be performed during surveillance testing under the two hour time constraint. In addition, the MSTs will be revised to enhance the specific instruments that will be made inoperable by jumper, and training for the appropriate people will be performed. The failure to place channel A2 of the RPS SDV water level high trip scram signal in the tripped condition is a violation of TS 3.3.1:- Failure to Place Channel- A2 In the Tripped Condition, (325/90-41-01). . . _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___--
....A.. - . , ' , ; , , . . . , , , , , ' ' '
- .
, , 7 This licensee identified violation is not being cited because i criteria specified in Section V.G.1 of the NRC Enforcement Policy were satisfied. b. (Closed)LER 1-90-017, Unit 1 High Pressure Scram During Performance of Turbine Control /Stop Valve Tightness Test. A unit 1 scram, from high pressure, during the ierformance of Turbine Control /Stop Valve Tightness Test, occurred oi September 27, 1990, and was discussed in m inspection report 90-37. At the time of that report, the licensee was still investigating tne event and preparing a LER. The investi- gation was subsequently completed and LFR 90-017 was issued on October 26, 1990. The LER provided a description ci the events surrounding this-occurrence and contained recommendations to prevent recurrence. This event occurred with the unit at approximately c2 percent po;:er while in the process of shutting down for a refueling outage. The licensee had started the planned Periodic Test (PT) 40-2-10, Turbine Control /Stop Valves (TCV/TSV) Leak Tightness Testing. The event occurred due to erroneous procedural guidance provided by the vendor, General Electric, in GEK-25406A and defective switches on the TSVs, which allowed the TCVs to open when the TSVs were closing. The turbine BPV's open demand signal was limited by the maximum combined flow circuitry of the turbine control system. The closure of the TSVs without the BPVs being able to open caused reactor pressure to increase to the SCRAM setpoint. A detailed licensee review of this event has determined that this proceaure contained weaknesses which are applicable to both nuclear and fossil plants which have used this procedure / guideline to develop their plant tests. The licensee had made this information available to other utilities through " Network" and have indicated that GE may issue an information letter on this item, c. (Closed) LER 2-90-15, Unit 2 Reactor Scram-Due to Loss of Excitation on Main Generator. A' scram occurred on Unit 2 on September 27, as a . result of the loss of excitation on the main generator. This item was also discussed in inspection report 90-37, and the licensee was investigating this event at the close.of the previous inspection period. The licensee subsequently completed their investigation into the event and provided the details in LER 90-015, dated October 26, 1990. The inspectors have reviewed the LER and other licensee docu- mentation associated with the event and discussed this matter.with NRR. Based on this review and discussion with the licensee and NRR, the inspectors are unable to determine the following: -(1) was the capacity and. stability of the offsite power system prior to and immediately after the Unit 2 trip adequate to provide acceptable voltages to handle a design basis event in Unit' 2 and a safe shutdown on Unit 1; (2) does the licensee's loss of voltage protection system (relays). which only sheds the safety-related equipment off the "E" buses, provide- adequate protection for system voltages less1 than the setpoint of the degraded voltage relays; (3) if the offsite power voltage drops to a level that is inadequate to start safety-related loads, will the 10.5 second degraded voltage relay time delay result ' in a delaying safety bus transfer to the diesel generator such that _
- . . . . . . . . . . . . . . . . . . . .. . . . , . . - 8 the plant may be outside the boundary of this safety analysis; and (4) does CP&L provide adequate direction and guidance to system load dispatchers to ensure that adequate voltage is maintained at all nuclear units? These items will be referred to NRR for review. Pending the outcome of the above, these items will be tracked as an inspector followup item: Adequacy of Offsite Power, (325,324/90-41-02). One e-cited violation was identified. 8. 10CFRPart21 Items (36100) (OPEN) 325,324/P2188-01 - Worn Shaft Gear Failures in Size 2 Limitorque Actuators and Also in Fisher Supplied H3BC Actuators. The inspector discussed this item with the licensee. A review of licensee records could not determine that this report had ever been received. They contacted Limitorque who verified that the information had been provided, so it was apparently lost. The licensee has entered this item into FACTS and has assigned responsibility for investigation and resolution. This item will be reviewed further as information becomes available. (CLOSED) 325,324/P2188-04 - Reinstalling Foxboro Controller Circuit Cards May Cause 100 Percent Output and Subsequent Transient to Occur. This controller has been identified in BWR Recirculation Flow Control System -DCS-88080301. This Part 21 was sent by the Foxboro Company to the Perry Nuclear Power Plant, it related to SPEC 200, Model 2AC-D+44 controller card with its associated 2AX+RM removable manual card. This notice was not sent to Brunswick since they do not use these units in their recirculation flow control system. A review of parts and installation by the licensee has verified that these units are not installed or used in spare parts at Brunswick. (CLOSED) 325,324/P2189-01 - Brown Boveri K-Line, K-225 Through K-2000 Circuit Breakers Delivered Prior to 1974 Had Rebound Spring Added to Slow Close In. This item identified that testing had discovered that the above breakers may fail to function properly in that persistent sine dwell vibration could occasionally cause the sicw close bar-to move into-a position such that the breaker, when called on to close, could slow close rather than closing normally. Adding a rebound spring to the slow close ' lever will prevent the slow close bar from vibrating to the undesired position. Brunswick was identified as a plant that had received shipment of these breakers. A survey by the licensee found that 10 electrically operated breakers of this type were installed in the plant. Substations 3L and 4L located in the hot machine shop have K-1600 incoming main breakers. These units were manufactured in 1972 and do not contain rebound springs but are used in non-safety applications to provide power .to the warehouse, maintenance shops and office buildings. Emergency substations ES, E6, E7, and E8 were found to have K-3000 breakers and the Part 21 does not apply to the breakers. The four remaining breakers are installed as crosstie breakers on substations ES, E6, E7, and E8. These .are K-1600 breakers manufactured before 1972 and do not contain rebound _ ..
- _ _ _ _ . . _ _ _ . . . . . ,,
. 9 spri ngs ~. However, these breakers are administratively controlled in the racked out position and verified racked out as part of normal system surveillance testing. The breakers may be closed at the discretion of the shift supervisors when both units are in mode 4 or 5 and other TS requirements are met. No credit is taken in the accident analysis for use of these tie breakers. Based upon the above, the licensee discovered that the use of these components at Brunswick does not pose a substantial safety hazard and, therefore, is not reportable under 10 CFR Part 21. Since this defect would only occur under seismic event conditions and the components are not used for safety applications, the licensee does not have any current plans to install recommended springs in the substations used for the warehouse, maintenance shop, and office buildings. The licensee plans to purchase and install the springs in the crosstie breakers for substations ES, E6, E7, and E8 at the next scheduled maintenance period after receipt of the springs. (CLOSED) 325,324/P2189-05 - PT-21/ Germane to Safety from GE: Susceptibility of Weld Between Core Spray Line and Thermal Sleeve to IGSCC. GE recommends that welds be included in IE Bulletin 80-13 surveillance. This item was evaluated by GE and the licensee and determined to be germane to safety, but not reportable. The licensee has implemented the vendor recommendations and included this item in their surveillance program to be tested each refueling outage under Periodic Test, Core Spray /Feedwater Visual Examination, PT-90.1, and reported under IEB 80-13. The iaspector verified that these reports had been submitted as required. (CLOSED) 325,324/P2189-06 - PT-21/ Germane to Safety from GE: Concerns with Core Neutron Flux Monitoring and Reactor Protection During Refueling. This item was the result of NRC questioning the conservatism cf center-spiral reloading because the SRMs were not on scale and, therefore, not monitoring neutron flux changes during a refueling at the Brown's Ferry Plant. GE performed an evaluation of this event and concluded that this event did not-constitute a substantial safety hazard and was not reportable under the context of 10 CFR Part 21. However, they oid conclude that this issue was germane to safety. Based on the above, RICSIL No. 039 was issued by_GE on February 10,-1989, to alert BWR owners that interim recommendations would be provided. These-interim recommendations were: (1) that during refueling, the neutron monitoring system should be capable of continuously monitoring changes in neutron flux in the region of the core'where fuel is being loaded or control rods are being removed to provide operators with indications of an approach to criticality; (2)'that the RPS should be capable at all times of reliably. initiating)a reactor scram based on inputs from the neutron flux detectors; and (3 that the recommendations of SIL 372 and SIL 68 should be followed when refueling interlocks are bypassed. The inspector reviewed the licensee's Engineering Evaluation Report (EER) 89-0022, dated January 16, 1989, that evaluated this event and the vendor recommendations, the licensee's Refueling Fuel Handling Procedure FH-11, Volume IX, Revision 41, and Engineering Procedure Guidelines for Prepara+1on of Core Component Sequence Sheets, ENP-24.12, . i ! .. .
. . . ,, ' . 10 Volume XX, Revision 8, to verify that the vendor recommendations had been implemented. This review indicates that Procedure FH-11 provides Administrative Control and direction over core reloads, that all control rods will be fully inserted during fuel movement, that the source range monitors will be operable and providing on scale indications, that refueling interlocks will .be operable and not bypassed or jumpered out, and that the core loading will progress in a sequence which ensures that the SRMs have accurate indication of changes in neutron levels. These procedures, therefore, implement the vendor recommendations and appear to be satisfactory. EPRI is currently conducting a study on this issue. It is anticipated that this study will be completed in late 1990. The licensee has indicated that they will review and implement the recommendations of that study as appropriate to the Brunswick Plant. (CLOSED) 325,324/P2189-18 - SMB Actuators Found to Have Melamine Torque Switches That Undergo Post Mold Shrinkage and Cause Cam Binding. Melamine Torque Switches Found to be Not Qualified. The licensee reviewed this item and determined it to be applicable to BSEP. A decision was made to replace the applicable torque switches on PCIS valves during the 1989-90 refueling outage for Unit 2 and on Unit I during its refueling outage in 1990-91.- The remaining torque switches were scheduled for replacement when routine maintenance is performed on the remaining valve actuators with all replacement work completed by July 3,1992. The inspector reviewed the completed work request for PCIS replacement accomplished on Unit 2 during the past outage, and the work scheduled for Unit 1 during the current outage. The listing of work scheduled under the routine maintenance program for the remaining torque switches was also reviewed. It appears that the licensee has determined which torque switches require replacement and have the program needed to complete these activities underway with an established completion date. (CLOSED) 325,324/P2189-12 - PRE-1981 SMB-000 and PRE-1976 SMB-00 Cam Type Torque Switches Can Fail as a Result of Stationary Contact Screws Loosening on Side of Torque Switches That Had Fiber Spacers. Two failure modes were reported. The first type failure resulted when one of the screws in the contact bridge came loose resulting in premature tripping of the torque switch. The second failure resulted when both stationary contact screws loosened and the contact bridge raised with the contact fingers maintaining continuity on the torque switch. The licensee's evaluation concluded that no events of this nature have occurred at BSEP and, due to the low number of failures reported, concluded that the_ safety significance was low. A licensee's review of safety-related SMB-00 and SMB-000 valves at BSEP determined that these switches, which were made of melamine or ahenolic materials, had been or were:in the process of being replaced wit 1 new fiberite torque switches as a result of 10 CFR 21.89-18. The remaining SMB-000 and SMB-00 torque switches-that were made of fiberite, were: verified as having been replaced since 1985; with the exception of one switch,1-E21-F001A-M0, that is planned to be worked under Work Request 89-AZIA1. The new torque switches and any SMB-000 purchased since 1980 and SMB-00 switches purchased since 1976, do not include fiber spacers and do not have this problem. Based on the above,
- - . _ _ _ . _ . . . . _ . . .. .. .. ,, ' . 11 it appears that the licensee has identified all equipment affected by this [ item and has either completed or established documentation to inspect and L replace the remaining items. (CLOSED) 325,324/P-2190-04 - Rosemount Resistance Bridges Can Exhibit - Premature Long Term Degradation Under Certain Combinations of Humidity. - Power and Temperature. Two of the units referenced in the above, serial numbers 0067897 and 0067898, were provided to BSEP. Neither of these units have been installed. They have been placed under administrative hold in stock under CP&L Part Number 731-758-12. They will be returned to the vendor for replacemen+ when the new units on order are received. This " item is closed. Violations and deviations were not identified, f 9. Action on Previous Inspection Findings (92701) (92702) -- (CLOSED) Violation 325,324/88-18 05, HPCI/RCIC High Steam Line Flow Instruments Inoperable. The inspector reviewed the licensee's response to the Notice of Violation and Civil Penalty dated January 27, 1989. The specific corrective actions taken to resolve the instrument's setpoints was detailed in LER'l-88-14, which was inspected and closed in Report No. g 90-37. The violation was issued because of the inadequate corrective actions taken to resolve the issue when it was first identified. In their response, the licensee coninitted to revising their corrective action program along with establishing a program to effectively implement the BSEP system engineering concept. Weaknesses in the licensee's corrective action and system engineering programs were also noted in the Diagnostic Evaluation Team report. As a result, the licensee included.these specific areas into their IAP program which resulted in further NRC inspections to followup on the program's implementation. The licensee's new corrective action program was inspected in Report No. 90-31. The report stated that the actions committed to in the IAP were completed. The effectiveness of the new program was not assessed and will be evaluated in future inspections. 7 1 System engineering program improvements were inspected in Report No. 90-16. The report stated that the appropriate procedures and programs were in place to correct deficiencies in this area. Based on the inspections. completed to date, which address the required corrective actions of the violations and the additional inspections planned to assess the effectiveness of the licensee's corrective action, this item is _ closed. (CLOSED) Violation 325,324/88-24-03, Silicon Bronze Bolts Corrective . Actions. .The inspector reviewed the licensee's response to the Notice of Violation and Civil Penalty dated January 27, 1989. This violation, along with 325,324/88-18-05, were included in EA-149 and constituted the 2 _ examples of inadequate corrective action. The specific actions taken with regard to the silicon bronze bolt issue is detailed in LER 1-88-06, which _ . _ .-'
s p ' > ,, , 12 was inspected and closed out in Report No. 90-37. As discussed in the closcout of Violation 88-15-05, the licensee's actions to resolve corrective action program deficiencies have been and are continuing to be inspected as part of the followup to the licensee's IAP. Based on these inspections, this item is closed. (CLOSED) Violation 325,324/88-34-01, Inadequate Design Control Related to RCIC Steam Exhaust Check Valve. This violation concerned the Unit 1 and 2 RCIC Steam Exhaust Check Valve, E51-F040. Plant Modifications 81-274 and 81-275, replaced the Unit 1 and 2 valves which included discs with design pressures of 25 psig. Based on the calculated peak containment pressure - during a DBA of 49 psig, and containment design pressure of 62 psig, the - discs were under rated. The inspector reviewed the licensee's response to the violation and supporting documentation.. tER-88-0461 was written to evaluate the design pressure discrensocy and concluded that the discs were sufficient. This was based on successful local leak rate testing at 49 psid and documentation from the valve manufacturer stating that the discs would " withstand a pressure of 62 psig at 248 degrees F for a sustained period of time". The EER concluded that these values are greater than the containment system requirements for the DBA and higher than credible turbine exhaust operating pressures. The cause of the design error was that the containment accident pressure and LLRT pressure were not evaluated; only operational exhaust pressures were considered during the design process. Revisions to applicable administrative and engineering procedures have added formalized checklists and training requirements for safety reviewers. (CLOSED) Violation 325,324/88-38-01, Failure to Control Combustibles in Restricted Plant Areas. The inspector reviewed the licensee's response to the Notice of Violation and supporting dxumentation. The licensee stated that the reason for the violation was t' n personnel not f amiliar with fire-retardant wood requirements obtaired wood from outside the protected area for use as forms to install a plant modification. A contributing factor was that the modification instructions did not specify the type of material to be used for the forms. The-licensee enhanced the existing controls to allow only fireproof material and/or fire retardant wood to be used within the plant protected area. An improvement in identification markings for fire retardant wood to ensure this type of wood, including cut up pieces, is readily identifiable for use in the protected area. The training lesson plan for the plant construction support group annual training was revised to address this topic. , (CLOSED) Unresolved Item 325,324/88-05-01, Service Water System Operating Mode Concerns. Other inspections of this issue are documented in Inspec- tion Reports 89-09, 89-12 and 89-14. -As a result of these inspections and others perfonned on the service water system, a Notice of Violation and. I Imposition of Civil Penalty was issued on January 26, 1990, for failure to take adequate corrective actions for identified service water deficiencies. Example A of- the violation describes the inadequate evaluation performed when determining system operability with the single failure concerns
. .. . , 13 associated with the SW-V106 valve. Further inspection of this issue will be performed in the closeout of Violation 325,324/89-34-47. (CLOSED) Unresolved item 325/88-34-02, Reactor Vessel Water Level Wide Range Indication Anomolies. This item was also discussed in inspection o report 325,324/90-02. Currently the licensee is replacing Rosemount transmitters in Unit 1 and is delaying any further action on the wide range indicators pending completion of the replacements. The licensee expects some indication difference with the new transmitters. The inspector determined that no regulatory issues existed. Therefore, this item is closed. (CLOSED) Unresolved item 325,324/88-38-02, Failure to include All LPCI and Suppression Pool Cooling Flow Path Boundary Valves in Surveillance Program. This item was also discussed in inspection report 89-05 and was expanded to include core spray system valves. This item concerned the pts that meet the monthly TS surveillance requirements of ECCS systems for verification that each valve.in the flow path that is not locked, sealed, or otherwise secured in position is in its correct position. The licensee originally disagreed with the inspector on what was a " flow path" valve. - Specifically, minimum flow valves, vent isolation valves, valves normally outofpositioninstandbylineup(i.e.,E11-F027A(B),RHRSuppression Pool Spray Isolation), or flow path boundary valves were not included in the surveill'ances. The licensee has subsequently revised the pts to include these valves. No occurrences are known where these valves have ' been found out of position due to their omission in the pts. This item is -closed. (CLOSED). Unresolved item 325,324/90-17-02, Potential Inoperability of CBEAF System. Based on inspector's quertions regarding the differential pressure measurement technique used to determine the CBEAF system operability, the licensee changed the test procedure to more accurately measure the differential pressure (see inspection report 90-02). Subsequent tests verified that previous test results may have been s inaccurate in cietermining that a positive pressure existed in the control . building relative to the outside atmosphere. Based on the licensee's previous analysis and documentation from 1985,.the erroneous differential pressure measurement technique has minimum safety significance, therefore, this item is closed. (CLOSED) IFI 324/88-15-05, Normal Position for SW-V117, Nuclear Header to Vital Header Isolation Valve. As a result of extensive review and analysis of the service water system design, the normal position of the SW-V117 was changed from closed to open. The valve position was changed to allow a.RHR room cooler to be placed in service affording the service water pumps minimum flow protection under worst case single failure , scenarios. The Service Water System Operating Procedures, 1-0P-43, ' Revision 31, and 2-0P-43, Revision 68, require that the valve be open. The licensee plans to keep the valves open until the service water pump thrust bearing modifications are completed. This work is currently l scheduled to begin in 1992. ,
___ . . . - . _ _ - . . . . _, . .. , 14 (CLOSED) IFl 325,324/88-38-05, Licensee Activities Related to Correcting Keepfill System Discrepancies. The licensee has determined that the primary cause of keepfill system problems are bent stems on pressure control valves due to over-pressurization downstream. Several causes of over-pressurization have been identified. Currently three of twelve keepfill systems are inoperable. Modifications are scheduled for the current Unit 1 outage and the next Unit 2 outage to replace pressure control valves. This item is closed based on the licensee's identification of the causes of the keepfill system discrepancies and planned corrective actions. Violations and deviations were not identified. 10. ExitInterview(30703) The inspection scope and findings were summarized on November 5, 1990, with those persons indicated in paragraph 1. The inspectors described the areas inspected and discussed in detail the inspection findings listed below. Dissenting comments were not received from the licensee. Proprietary information is not contained in this report. Item Number Description / Reference Paragraph 325/90-41-01 h0N-CITED VIOLATION - Failure to Place Channel A2 in the Tripped Position (paragraph 7.a). 325,324/90-41-02- IFl - Adequacy of Offcite Power (paragraph 7.c). 11. Acronyms and Initialisms A0- . Auxiliary Operator APRM Average' Power Range Monitor ATTU Analog Transmitter Trip Unit BPV Bypass Valve BSEP Brunswick Steam Electric Plant -BWR Boiling Water Reactor CBEAF Control Building Emergency Air Filtration CP&L Carolina Power & Light Company DBA Design Basis Accident ECCS Emergency Core Cooling System EER Engineering Evaluation Report ENP Engineering Procedure EPRI Electric Power Research Institute- ESF Engineered Safety Feature F Decrees Fahrenheit FACT" Facili y Automated Conmitment Tracking System FWLCS Feedwater Level Control System GAF Gain Adjustment Factor GE General Electric HP Health Physics . . . . . . , . . . . . . . . . . . . . . . . . . . , . . . . . . = _
_ , _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ __ - _ _ __ - _ _ _ __ ______ __-__ __ _ _ . . , . . ' .: 15 i 3 HPCI High; Pressure Coolant Injection IAP- Integrated Action Plan I&C- Instrumentation and Control ~1E Inspection- and ' Enforcement IEB- . Inspection and Enforcement Bulletin IFl -Inspector Followup Item IGSCC' Intergranular Stress Corrosion Cracking
IPBS Integrated Planning, Budget'ng and Scheduling IRM Intermediate Range Monitor LC0~ Limiting Condition for Opetation ' -LER Licensee Event-Report' LLRT -Local Leak Rate Test LPCI Low Pressure Coolant Injection MSL Main Steamline MST Maintenance Surveillance Test - NRC Nuclear Regulatory Commission i ' NRR Nuclear Reactor Regulation PA Protected Area .PCIS Primary Containment Isolation System ! PNSC- ' Plant Nuclear Safety Committee PSID Pounds per Square Inch Differential i PSIG, Pounds per-Square Inch Gauge PT ' Periodic Test QA Quality Assurance QC- Quality Control. - t RBCCW Reactor Building Closed' Cooling Water RCIC- Reactor-Core Isolation Cooling , RFP Reactor Feed Pump RHR = Residual Heat Removal . ,RICSIL Rapid Information: Communication Service Information Letter i , RPM Revolutions Per Minute 'RPS: ' Reactor Protection System .SDVt Scram Discharge Volume'-
, ' ~SF Shift Foreman iSRM Source Range Monitor i STA Shift Technical-Advisor- - t TCV/TSV Turbine Control-Valve / Turbine Stop Valve .; , TS >-Technical Specification ' URI? Unresolved. Item - > WR/J0: Work. Request / Job Order ( - !, . ' y ' . }}