ML20236M646
| ML20236M646 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 10/29/1987 |
| From: | Fredrickson P, Garner L, Ruland W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20236M591 | List: |
| References | |
| 50-324-87-35, 50-325-87-31, EA-82-106, IEB-79-07, IEB-79-14, IEB-79-7, IEIN-84-53, NUDOCS 8711130225 | |
| Download: ML20236M646 (24) | |
See also: IR 05000324/1987035
Text
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i f AT LANT A, oEoRGI A 30323 %,+- / ..s Report Nos. 50-325/87-31 and 50-324/87-35 . Licensee: Carolina Power and Light Company P. O. Box 1551 Raleigh, NC 27602 Docket Nos. 50-325 and 50-324 License Nos. OPR-71 and DPR-62 Facility Name: Brunswick 1 and 2 Inspection Con tct d: September 1-30, 1987 / /o/J-$ //9 Inspectors: ) < ~ - 4 W.'He-R and Fate 51gned w . / o l3>) f)>'7 f % L. W. Garner Da'te Sfgned Other Contributing Inspectors: L. S. Mellen R. M. Latta S. J. Via. . /o/2;3 /A7 Approved by: - ) me A. P.~E. Fredrickson, Section Chief 0#te Signed Division of Reactor Projects SUMMARY Scope: This routine safety inspection involved the areas of followup on previous enforcement matters, maintenance observation, surveillance conservation, operational safety. verification, onsite Licensee Event Reports (LER) review, followup on inspector identified and unresolved items, 10 CFR 21 followup activities, review of special reports, bulletin followup, ESF System walkdown, inservice inspection program hydrostatic tests and diesel generator thermostatic control valve. Results: One violation was identified - inadequate surveillance procedure, f l 8711130225 871030 PDR ADOCK 05000324 l ! G PDR l I ( ___.___1_ __ _ j
_ _ _ - . _ , . . - >. ., . - L L 4 , REPORT DETAILS 1. PersbnsContacted i q- l Licensee Employees ! -P. Howe, Vice President'- Brunswick Nuclear Project C. Dietz, General Manager - Brunswick Nuclear Project T. Wyllie, Manager - Engineering and Construction G. Oliver, Manager - Site Planning and Control J. Holder, Manager - Outages R. Eckstein, Manager - Technical Support E. Bishop, Manager - Operations- ~L. Jones, Director - Quality Assurance (QA)/ Quality Control (QC) ~ R.,Helme, Director - Onsite Nuclear Safety BSEP J. O'.Sullivan,-Manager - Maintenance 'G. Ch'eatham, Manager - Environmental & Radiation Control ..J. Smith, Manager --Administrative Support K, Enzor, Director - Regulatory Compliance R. Groover, Manager - Project Construction V.-Wagoner, Director - IPBS/Long Range Planning
- A. Hegler, Superintendent - Operations
W. Hogle, Engineering Supervisor B. Wilson, Engineering Supervisor 'B. Parks, Engineering Supervisor R. Creech,.I&C/ Electrical Maintenance Supervisor (Unit 2) , R .: Warden,-I&C/ Electrical-MaintenanceSupervisor(Unit 1) W.' Dorman, Supervisor - QA W. Hatcher,. Supervisor'- Security -R.' Kitchen Mechanical Maintenance Supervisor (Unit 2) ' .R. Poulk, Senior NRC Regulatory Specialist D. Novotny, Senior Regulatory Specialist Other licensee employees contacted included construction craftsmen, engineers, technicians, operators, office personnel, and security force members. 2. Exit Interview (30703) The inspection scope and findings were summarized on October 5, 1987, with the general manager. One violation - inadequate diesel(TCV) generator surveillance procedure: Thermostatic Control Valve not in restoration line-up of PT-12.2.b (paragraph 14), was discussed in detail. -The licensee acknowledged the findings without exception. The licensee did not identify as proprietary any of the materials provided to or -reviewed by the inspectors during the inspection. - - _ _ _ ___
p. L 3 u 2 ! 3. Followup on Previous Enforcement Matters (92702) i L (CLOSED) Violation 325/82-10-02 and 324/82-10-02, Failure to Implement Double Verification. Volume I, Book 1 of the Plant Operating Manual, Administrative Procedures, now explicitly specifies, in Section 11.7, the independent verification requirements for all plant procedures. The , ! operations subgroup implements the above requirements in 01-13, Rev.17, Valve and ' Electrical Line-up Verification, Valve Identification, and locked Valve Identification and Locking, Section 4.4. The inspector also reviewed 01-01, Rev. 20, Operating Principles and Philosophy, and 01-28, Rev. 7, Preparation and Review of Operations Procedures, and Selected System Operating Procedures, to verify that operations completely implemented the independent verification requirements. The inspector noted that 01-13, Section 4.4.1, stated that "both ' individuals must have performed or directly observed the verification of valve or breaker position as described previously......" This statement implies that it is permissible for an operator to watch another operator verify a valve position, with only one operator actually attempting to f verify valve position (by attempting to turn the valve handwheel in the shut direction). The licensee's response to item I.C.6, dated December 31, 1980, committed them to only having a second person verify i proper system alignment and does not describe the nature of the ! verification. The inspector discussed this issue with plant management during the exit interview. The licensee stated that they need to resolve what is meant by " return to service" as specified in I.C.6, as well as resolve any ambiguities in the operation group's practices. This is an 1 Inspector Follow-up Item: Operations to Resolve Independent Verification Policy Questions (325/87-31-03 and 324/87-35-03). The maintenance subgroup implements the independent verification requirements in their procedures through MP-52, Revision 1, Standards for Preparing and Maintaining Maintenance Procedures. Section 5.1.3.10 of MP-52 Independent Verification (IV), requires independent verification of restoration to service of all systems and components listed in the plant's Administrative Procedures as requiring IV. The inspectors verified through a review of selected maintenance procedures, that IV has been implemented in maintenance procedures. The inspectors also conducted interviews with selected maintenance workers and determined that the workers understanding of plant policy toward IV was adequate. (CLOSED) Violation 325/82-28-01 and 324/82-28-01, Surveillance Program Breakdown Resulting in Multiple Missed Surveillance. Inspection Report 82-28 contained six separate violations. On February 18, 1983, Proposed Civil Penalties, EA-82-106 was issued concerning the i tems . _ It consolidated the items into two violations: Failure to Provide Surveillance Procedures and Failure to Take Action to Preclude Recurrence. The former item was inspected and closed out in Inspection Report 82-39. Specifically, the inspectors verified that the specific missed _ _ _
__ , . l 1 ! ' 3 i surveillance were performed and other similar problems were identified and properly corrected. In addition, the inspectors have verified that ' recent surveillance requirement changes, specifically, 4.8.2.3.1, 4.1.5.C 3, action statement 3.8.1.1.b.1 & .2 and 4.8.1.1.2.a.4, were incorporated in procedures PT-12.6, PT-6.2.1, PT-12.8 and PT-12.2.a. b, c j & d, respectively. Review of the second item is contained in Inspection ) Report 84-13 with a subsequent followup in Inspection Report 84-27. j i The corrective actions associated with the management and program 1 weaknesses which allow the violations to occur were incorporated into the 1 Brunswick Improvement Program (BIP). On December 22, 1982, the items contained in the BIP became requirements when they were confirmed by Confirmatory Order EA-106. In January, 1984, a letter from the Region II 3 Regional Administrator to the Carolina Power and Light Co. Executive Vice J President, stated that the requirements imposed by the order had been i A satisfied. Status of items which had long completion dates, designated as ongoing items, were to be sent at six month intervals until completion. On May 30, 1986, having completed addressing all the ongoing items, the licensee submitted their final status report. Inspection of these items may be found in Inspection Reports 87-05 and paragraph 8 of this report. (CLOSED) Violation 324/83-17-01, Failure to Comply with a Technical Specification Action Statement Concerning Inoperability of Steam Jet Air Ejector Radiation Monitor. The inspector reviewed the licensee's response to the Notice of Violation, dated August 11, 1983. The following corrective actions were verified as currently implemented: Item 3.b Once per shift review and update the Lighted Annunciator Status Report as required by paragraph V 6 of 01-05, Abnormal Annunciator Status, Revision 9, dated April 7, 1987. Item 3.c Review and initial recorder charts as required once per shift per Section 4.1.8.c of Administrative Procedure (AP), Revision 106, dated August 10, 1987. l Item 3.d Review Annunciator Status Report with offgoing control operator and update the report before going off shift as required by the Control Operator Shift Checklist of 01-02, Shift Turnover Checklist, Revision 23, dated September 16, 1987, i Item 4.b Daily trending of critical parameters as required per 01-23, Trend Analysis, Revision 8, dated October 17, 1986. Changes have been made to AI-58, Equipment Clearance Procedure, to improve function and auditability, as stated in Item 4.a. l (CLOSED) Violation 325/84-27-01, Inadequate 10CFR50.59 Review. The inspector interviewed selected 10CFR50.59 preparers and information provided during 10CFR50.59 training. Additionally, the inspector reviewed selected reference materials provided to qualified 10CFR50.59 preparers ! < - - - - - - -
_ __ _ __ - - _ _ - _ _ _ _ _ _ - , l
l' 4 l l and selected 10CFR50.59 evaluations. The licensee's program appears adequate. This was previously inspected in Inspection Report 87-06. (CLOSED) Violation 325/84-30-02 and 324/84-30-02, Surveillance Test Fails to Verify All Actuated Devices Actuated on Simulated Signal. The l inspector reviewed the licensee's response dated November 30, 1984. The ! inspector had reviewed, on October 4, 1984, the special test procedures, ! SP-84-150 and 153, which had been performed to verify that valves 1-B32-F031A & B and 1-B32-F032A & B would close on a loss of Coolant ' Accident (LOCA) signal in conjunction with a low reactor pressure l permissive signal. The inspector also verified that PT-08.1.2, Low l Pressure Coolant Injection (LPCI) SimA ated Automatic Actuaticn and Logic l Functional Test, Revision 21, dated September 18, 1985, adequately tests these ulves in Section VII.VVV. Logic Functional Test, 1-MST-RHR41R, l Revision 9, dated June 17, 1987, and 2-MST-RHR41R, Revision 6, dated l- August 5,1987, entitled Residual Heat Removal (RHR), LPCI, and Primary Containment Isolation System (PCIS) Group 8 ISOL LOGIC SYS FUNC TEST, also test the subject valves in Sections 7.7 and 7.12. The inspector witnessed, on June 8,1987, successful performance of Section 7.7 of 1-MST-RHR41R. The inspector also reviewed Operating Experience Report 0ER-2-84-03, LPCI Flow Test Automatic Alignment Test, dated November 2, 1984, which documented the results and disposition of items identified during an engineering review for similar type problems on other systems. In addition, the inspector verified that Section 5.4.4.4, Logic System Functional Test, of Maintenance Prncedure MP-52, Standards for Preparing and Maintaining Maintenance Procedures, Revision 1, dated July 29, 1987, discusses testing of actuated devices and the inter-relationship betwun the automatic actuation portion of the system functional test and the logic functional test. This reflects the philosophy utilized during the surveillance procedure rewrite program completed in early 1986. (CLOSED) Violation 325/85-05-01, Failure to Properly Resume Average Power Range Monitor (APRM) Calibration Procedure PT-1.17PC Resulted in an Unanticipated Half Scram. The inspector reviewed the Notice of Violation response dated June 3,1985. The inspector verified that Maintenance Standard Operating Procedure, S0P-2.17, was issued on May 29, 1985, to provide guidelines on turnover of surveillance activities. The inspector verified via lesson plan and training roster that appropriate Periodic Test (PT) crew members received training as comitted in the licensee's response. (CLOSED) Violation 324/85-27-01, Bolts Replaced on Hydraulic Control Units with Type Other Than That Specified on Drawing. The inspector reviewed the Response to the Notice of Violation dated October 23, 1985. The inspector reviewed the completed work request, dated December 9, 1985, which replaced the subject bolts with cadmium plated bolts. Verification that training on proper use of documents when researching parts was via training roster. In addition, the inspector reviewed Operating Experience Report 0ER-2-85-30, Assembly Fasteners - Hydraulic Control Units. No additional violations or deviations were identified. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . __ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __
- . u . ! 5 i ' ! 4. ' Maintenance Observation (62703) The inspectors observed maintenance activities and reviewed records to ! verify that work was conducted in accordance with approved procedures, Technical Specifications, and applicable industry codes and standards. The l i 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' , l 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: 87-BEKJ1 Weld Repair of Reactor Core Isolation Cooling (RCIC) Valve 2-E51-F040. 87-BEBE1 Diesel Engine Analyzer Testing. MI-25-41 Welding Instruction WPS-P-M1-B-1 (Weld Repair of 2-E51-F040 Disc). No violations or deviations were identified. ! 5. Surveillance Observation (61726) i The inspectors observed . surveillance testing required by Technical Specifications. Through observation 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. l The inspectors witnessed / reviewed portions of the following test activities: l 1MST-RCIC15M RCIC Steam Leak Detection Channel Functional Test. I 2MST-RLE26Q Condensate Water Level Functional Test and Channel Calibration. . 1 PT-13.1 Reactor Recirculation Jet Pump Operability. l l No violations or deviations were identified. l l .
_ __ , . 6 6. Operational Safety Verification (71707) The inspectors verified conformance with regulatory requirements by direct observations of activities, facility tours, discussions with personnel, reviewing of records and independent verification of safety system status. m'> The inspectors verified that control room manning requirements of 10 CFR j 50.54 and the Technical Specifications were met. Control room, shift supervisor, clearance and jumper / bypass logs were reviewed to obtain i information concerning operating trends and out of service safety systems 'to ensure that there were no conflicts with Technical Specifications Limiting Conditions for Operations. Direct observations were conducted of control room panels, instrumentation and recorder traces important to i l safety to verify operability and that parameters were within Technical Specification limits. The inspectors observed shift turnovers to verify f that continuity of system status was maintained. Some inspections were
conducted during backshift hours. The inspectors verified the status of selected control room annunciators. Operability of a selected Engineerd Safety Feature (ESF) train was verified by insuring that: each a ssible valve in the flow path was in its correct position; each power supply and breaker, including control room fuses, were aligned for components-that must activate upon initiation signal; removal of power from those ESF motor-operated valves, so identified by Technical Specifications, was completed; there was no leakage of major components; there was proper lubrication and cooling water available; and a condition did not exist which might 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 a review of area surveys, < radiation work permits, posting, and instrument calibration. ! The inspectors verified that: the security organization was properly manned and security personnel were capable of performing their assigned functions; persons and packages were checked prior to entry into the protected area (PA); vehicles were properly authorized, searched and escorted within the PA; persons within the PA displayed photo identification badges; personnel in vital areas were authorized; and ' effective compensatory measures were employed when required. t The inspectors also observed plant housekeeping controls, verified position of certain containment isolation valves, and verified the operability of onsite and offsite emergency power sources. l On September 13, 1987, the inspector observed that the Unit 2 Electrical l Protection Assembly (EPA) breakers No. 5 and 6, were tripped. These breahrs are an alternate power source for one of the Reactor Protection ' _ - - - - - - _ - - _ _ _______---______a
,__ , ,ti '_.' fE 7 ISystem L(RPS) buses. Per attachment :1 of Operating.- Procedure OP-03., <
- Reactor' Protection ~ System, Revision 17, dated January 23,'1987, the:
breakers'are normally shut. This condition was reported to the control - room,for corrective' action. Late in- the report: period,. on September 29, 1987, the.1icensee informed the inspectors that they had found additional Environmental Qualification issues. . 'No operability concerns were raised;.the questionably qualified equipment' was replaced. 'This item will be fully documented in'a subse.- .quent inspector report. ' No violations or deviations were identified. 7. Onsite Review of Licensee Event Reports (92700) 'The listed Licensee Event Reports . (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, licensee conditions and connitments. (OPEN).- LER 2-85-02, Inoperability of High Pressure Coolant Injection (HPCI)' System. Since the event, the licensee has replaced the HPCI electronic speed controller and differential pressure switch . 2-E11-PDIS-N0218,'with different models. The inspector determined that the HPCI failure should also have been documented as requiring a report as indicated in'10CFR50.73(a)(2)(v). The failure of the F006 valve was a condition that alone.could have prevented the fulfillment of the safety ' function of high pressure injection. In the exit' interview, the licensee agreed to. submit a supplementary report addressing the HPCI failure. No notice of violation is being issued since the licensee has improved their , LER process after submission of the LER as documented in a review of the licensee's LERs from AE00, forwarded from Region II on August 4, 1987.. In i addition,.the licensee volunteered to submit additional information on the
- F006 failure in the corrected LER.
This LER will remain open pending - -inspector review of the corrected LER. (OPEN) LER 1-85-59,- Reactor Scram Resulting from Containment Group 1 Isolation. This item was previously addressed in Inspection Report 87-06. The No. 2 Diesel Generator (DG) failed to start initially. The licensee 1 found that the diesel lube oil cooler temperature control valve affected j L 'the response time of the lube oil pressure switch, leading to the low lube 1 l . oil . pressure trip on the DG. The specified setpoint of the start time -delay relay, 2-DG4-STR, was 30 4.5 seconds. The "as found" condition of the relay was 26.81, seconds, within specification. When the STR relay cleared, the low pressure lube oil trip was activated, and the DG tripped and locked out. A problem was also found with a leak in the pressure switch line that was repaired. l 1 __ _ __ N
r; - - 1 . , f- I 8 The inspector reviewed a january 30, 1986 memo from the maintenance . manager to the Plant Nuclear Safety Committee (PNSC) that stated that DG 2 also had been operating with lower than desired lubricating oil s ' temperature, making the oil pressure respond slowly on DG start. While .all the other DGs operated satisfactorily during this event, this problem , ) could have affected other DGs. The inspector concluded that the DG TafTure exposed a problem that could have prevented the fulfillment of a safety function and thus was report'able per 10CFR50.73(a)(2)(v). The ] licensee agreed to submit a corrected report. As with LER 1-85-02, no notic of violation is being issued due to the extensive improvements the . licensee has in place in their LER development system. ] The inspector discussed the RCIC turbine trip with the cognizant maintenance engineer and observed the trip and re-' latch operation of V8 on the Unit 2 RCIC turbine during' surveillance testing. No' problems were noted. The inspector did find small cracks in the trip mechanism ball socket. The. cracks did not appear to affect the operability of the 1 i turbine at this time. Maintenance issued work request 87-BEZS1 to replace the ball socket during the upcoming Unit 2 outage. The licensee changed the setpoint of relay 2-DG4-STR in Maintenance
Instruction M1-3-12I to 45 seconds from 30 seconds to further insure-that ! the pressure switch would see the increase in lube oil pressure, further . i reducing the possibility of an inadvertent DG trip. Plant Modification PM-85-123 permanently changed the setpoint. The inspector also verified j that instrument calibration frequency of PS-6534, which had been drifting, was 'ir, creased from every '18 months to every 6 months. Through testing, the licensee found no indication of misplacement, air entrapment or line j sizing problems with the pressure switch. The licensee found that the ! thermostats for the jacket water and lube oil heaters used during diesel i' shutdown were not on any calibration program. The inspcctors verified that thermostats MUD-TS-6549 and LO-TS-6575 are now on a yearly i calibration schedule using MI-0k3H. The licensee decided that relocation { of the low lube' oil pressure switch was not required, based on changes 1 described above, solving the problem. The licensee we informed by the Thermostatic Control Va ve (TCV) j manufacturer (htertshaw) that the TCVs should be operated only if the j valve fails to wtomatically control temperature. The licensee modified > OP-39, Diesel Generator Operating Procedure, accordingly to require that 3 the TCVs be placed in the thermostatic position during performance of a ! diesel generator start.-up checklist. During a walkdown of the system, the I inspector noted a TCV cut of the thermostatic position. See paragraph 14 ! for details. l , This LER remains open pending licensee submittal of the corrected LER and I further inspector followup. _ _ - _
' x . .. 9 (OPEN). . LER 2-82-83, .Drywell to Torus Vacuum Breakers - X18A,' C and E > Problems. This item was previously addressed in Inspection Report 324/86-30, and left open pending establishment of. a suitable inspection method. The licensee plans to incorporate additional inspection i ' requirements in PT-20.6, Drywell to Torus Leak Rate' Test,' by October 30, 1987. (CLOSED) LER 2-83-01, Spurious Actuation of the Reactor Core Cooling, Turbine Exhaust Diaphragm " Pressure High" Annunciator Alarm Associated with Pressure Switch 2-E51-PS-N012A. Two separate instances of actuation of the subject annunciator alarm were identified in this LER. The first was attributed to instrument drift of PS-N012A which was corrected by recalibration of the pressure switch. The second event occurred 3 days later and was attributed to moisture accumulation in the instrument switch housing of PS-N012A. The immediate corrective-action was to remove the moisture accumulation and -seal the switch housing cover. Onsite investigation by the inspectors revealed that the 2-E51-PS-N012A-D
pressure switches had been replaced with instrumentation that was less i susceptible to instrument drift and moisture accumulation in accordance with plant modifications 2-83-173 and 1-84-184. Since the replacement of these switches, there have been no further instances of instrument drift or moisture accumulation ~. Reference Inspection Report 86-30. (OPEN) LER 2-83-33, Main Steam Line Radiation Monitors A and D Out of Calibration. This item was previously addressed in Inspection Report 324/86-25. The item was left open to track the licensee's evaluation and installation of applicable new electronic packages (GE NUMAC) for these monitors. The NUMAC drawers have been installed on Unit 1. Some of the < units purchased for Unit 2 have experienced failure during pre-installation checkout. If replacement units can be obtained, the new , monitors will be installed during the upcoming refueling outage, scheduled to begin January 2, 1988. The item remains open pending final resolution of the failure and installation on Unit 2. One violation and no deviations were identified. 8. Followup on Inspector Identified and Unresolved Items (92701) (CLOSED) Inspector Followup Item 325/82-25-01 and 324/82-25-01, Review Activities to Establish and Implement a Valve and Breaker Label j Replacement Program. This item was previously inspected in Inspection ' Report 325/86-32 and 324/86-33. That report indicated that standby gas treatment valves V8 and V9 were confusingly labeled on the Unit 2 Motor Control Center (MCC). This condition was again pointed out to the licensee during this inspection period. The licensee has issued work L request 87-BFIR1 and 87-BFITI to correct the deficiency. The report also i addressed 120 volt distribution panel labeling. The licensee intends to , l provide new labeling for the 120 volt distribution breakers associated j with safety-related MCC. Resources have been budgeted in 1988 and 1989 l i i ! ! I $ I
_ . _ _ _ _ _ _ _ _ _ _ . _
,- 10 . for this effort. The inspector verified via random sample that valves are properly tagged. In addition, 01-13, Valve and Electrical Line-up Verification, Valve Identification, and Locked Valve Identification and Locking, Revision 17, dated July 30, 1987,. requires that, if valve and/or breaker tags or labels are found to be missing, it is the responsibility of the shift foreman to replace them as needed. Valves and breakers with no tags may only be operated in an emergency to protect equipment or personnel. (CLOSED) Inspector Followup Item 325/84-SC-02 and 324/84-SC-02, BIP-B5.3.2.B - Consider Organi7ational Structure to Provide More e Centralized Control of Engineering. The inspector reviewed CP&L memo dated October 1. 1984 (CQAD 84-2231), and CP&L letter to NRC dated May 30,
1986(Serial: NLS-86-166). Both of these documents indicate that CP&L has
considered the appropriate organizational changes. The Corporate Design Organization is being phased in and will provide centralized control of engineering. Reference Inspection Report 87-05. (CLOSED) Inspector Followup Item 325/84-04-01 and 324/84-04-01, Licensee to Identify and Repair Cable Tray Raceway Z Clamp Problems. This item was previously inspected and left open in Inspection Report 87-03. The inspection report referenced a Notice of Deficiency (N00), associated with Non-Conformance Report (NCR) E-86-002, issued on February 27, 1987; i however, response to the NCR had not addressed poor work control practices j that had led to the inadequate Z clamp restoration. The inspector reviewed Wyllie to Jones memorandum dated March IC,1987, which detailed the corrective action taken by the Brunswick Construction Unit in response to the N00. The inspector verified via training roster that the supervisory training was performed as stated. The inspector also reviewed WP-217, Cable Pulling Procedure, and Specification 048-001, Section 2.2.11.5, Installation of Cable Tray Raceway Covers. The inspector reviewed Deitz to Jones memorandums dated March 10, 1987 and April 23, 1987, which addressed the plant's response to ' the N00. The inspector verified via training records that Maintenance Electrical Craft personnel were provided training on Z clamp installation and the need to restore tray covers to their design configuration. Based on the above, action to prevent future problems appears adequate. Therefore, this item is closed. However, additional existing problems with cable trays have been discovered by the licensee and a second N0D was issued on July 14, 1987, concerning these additional findings. Per Howe to Jones memorandum of 1 ' July 24, 1987, a target date of September 30, 1987, has been set to provide a schedule when the identified scope and final corrective action { completion date can be established. None of these new findings renders l any cable trays inoperable. In addition, Jones to Howe memorandum of September 11, 1987, requested that the work controls be addressed to l l lC __
-q ,. -{ 11 ensure loosening of fasteners is properly evaluated for impact on systems,
components or structures. Response to the second N00 and the Jones to -l Howe memorandum ; 1s an Inspector Followup Item: Work Controls for Fasteners Loosened on Operable Equipment (325/87-31-02 and 324/87-35-02).
J (OPEN) Inspector Followup Item 325/84-07-02 and 324/84-07-03, Periodic Test PT-46.4 Being Reviewed and Procedure Revision to Incorporate Required Acceptance Criteria. At the time of this inspection, the subject i procedure had not been modified to reflect test acceptance criteria for the control room positive differential pressure. The inspectors ., determined that this matter is currently being evaluated by NRR cnd that ! the resolution date had not been established. -This item remains open l pending establishment of an acceptable value for control room positive differential pressure and corresponding procedural implementation. Item , previously inspected in Inspection Report 87-06. 1 (OPEN) Inspector Followup Item 325/84-13-04 and 324/84-13-04, Completion of Plant Modification PM-82-030 and Resolution of Task Assistance Request (TAR)' B84-025 to Help Prevent Spurious Actuations of Emergency Core Cooling Systems. The inspector reviewed the , turnover package and , acceptance test of PM-82-030, 125 VDC Battery Charger Overvoltage l Protection. Full operability of PM-82-030 was declared on October 4, i 1985. TAR B84-025 has recommended modifications PM-85-020 (Unit 2) and I PM-85-021 (Unit 1) be installed. These modifications would provide separate battery sources ~for each of the two inverter / power supplies that , supply power to each division of emergency core cooling system actuation ' logic cabinets. The licensee anticipates that PM-85-020 and PM-85-021 i will be installed and operational by the end of the first quarter of 1988 i and the bnd of the third quarter of 1988, respectively. Due to changes in administrative procedures, TAR B84-025 was closed on August 16, 1985, and the item transferred to project control numbers 84809B and C. This item , will remain open pending completion of modifications 85-020 and 85-021. ! I (OPEN) Inspector Followup Item 325/84-31-01 and 324/84-31-01, Licensee to Develop and Submit a Technical Specification (TS) Change Request for Rod l Sequence Control System (RSCS) Testing. Reference Inspection Report j 87-11. The licensee submitted the TS change request on July 1, 1985. In j that submittal, the licensee also added an ACTION statement that allowed I bypassing inoperable control rods if: (1) the position and bypassing was ! I verified by a second licensed operator, (.2) there may not be more than three inoperable control rods in any RSCS group. This change was consistent with the GE/BWR-4 Standard TS. On February 14, 1986, General
Electric (GE) informed the licensee that NRC issued a generic Safety l
Evaluation Report (SER) which allowed the use of Banked Position Withdrawal Sequence (BPWS) for the first fifty percent of withdrawal for group notch control reactors. GE also stated that an inoperable control rod from 100% to 50% control rod density is not allowed to be inserted during group notch mode unless an analysis has been performed to demonstrate acceptable consequences. GE further stated that no formal l _ _ _ _
, - - , - - - , _ _ .- - - - - - . - _ - - - - - - - - - - -_ _ _ _ - , n < Q: 12 + l analysis h'ad'been performed.. This prompted the licensee to withdraw the. proposed TS change request by letter dated April 10, 1986. The surveillance-'. requirement addition, although not affected by . the above questions.,was withdrawn once'the TS amendment request was withdrawn. The licensee. plans.to request elimination of the RSCS system when GE's new [ Rod Worth Minimizer (RWM) is accepted by the NRC.. This item will remain { open until the licensee has an approved TS change to eliminate RSCS or the- '{ surveillance ~ requirements are placed in TS as originally intended by the licensee. . The licensee stated that the request to upgrade the RWM and eliminate RSCS should be submitted -in approximately 1 18 months. The . inspector. re-verified that 'the licensee had kept the surveillance requirement in GP-5, Unit Shutdown, Revision 24. Section 5.1.27. -(CLOSED) Inspector' Followup Item-325/85-16-02 and 324/85-16-02, Remote Shutdown Panel Wiring ' Deficiencies. The inspector. reviewed site ! memorandum EPB-2635 dated October 29, 1985, which documented- the "as i built" and drawing discrepancies of-the remote shutdown circuitry. The inspector verified that all major . components. . pumps, valves ; and controllers were reviewed and ~ plant . drawing correction packages PDC- .! ' 86-0002 and; PDC 86-0003, were issued for required' drawing changes. The inspector verified, . via site memorandum BEM-13362, that drawings. were revised to clarify orientation of HFA relays for the two different relay configurations. Special' Procedures, SP-84-017, SP-85-003, SP-85-023, SP-85-027 and SP-85-039 were verified to contain functional testing of all 1 major components: pumps, valves, dampers and fans which are required to i be operated by A0P-32.0, Plant Shutdown from Outside Control Room. These , tests ' included operation of RCIC from the . remote shutdown panels, . i operation of valves and pumps via their local controls, verification that ! valve and pump interlocks functioned correctly when controls are in their local control position and verification that the Group 5 isolation was- ! disabled ~when controls are in their loca1 ' control position. ! ! The licensee is in the process of preparing procedures which would require functional testing of some of these components, such as RCIC, at a yet to l J be determined periodicity. Development of these test procedures is an Inspector Followup Item: Develop Functional Test Procedures to Periodically Verify Operability of Remote Shutdown Equipment (325/87-31-04 and 324/87-35-04). (CLOSED) Inspector Followup Item 325/86-08-01 and 324/86-09-01, Revision of program documents to reflect organization changes. The inspector reviewed PSE&C Organization Manual (KB/5-5-86/04), Conduct of Nuclear Operations (Manual KB/3-5-86/04), Corporate Organization Manual (X-E-1), q and Corporate Quality Assurance Program (Revision 10). The inspector reviewed selected organizations for structure, functional responsibilities, levels of authority and lines of internal and external interfaces. For the areas selected, the upper tier and lower tier documents reflected the appropriate organizational changes. This item was previously addressed in Inspection Report 87-05. s 1 L- L i-_ _ - . .
. . ' 13 i (CLOSED) Unresolved Item -325/80-11-04 and 324/80-10-04, . Storage of '
Training and Surveillance Inspection / Test Records. This matter was I previously . discussed in Inspection Report 82-37. The NCR which tracked this. item was closed on September 23, 1986. Records Management Procedure RMP-00'2, Rev. 29, Records, Receipt and Storage, Section 6.2, states that temporary storage of completed Quality Assurance (QA) records may be maintained in one-hour UL-rated fire-proof file cabinets for up to 90 days prior to transmittal to the Records Storage Facility (vault). Exceptions to this requirement must be approved in writing by the Vice-President-BNP. The inspector verified that training records are now' stored in the vault and that temporary storage (less than 90 days).of the records is in a UL approved one-hour rated file cabinet. (CLOSED) Unresolved Item 325/81-22-01 and 324/81-22-01, Control of Plant ! Modification Drawings. The inspector verified that Revision 2 of Plant Modification 80-255 included the missing sketches. The Brunswick .
Construction Unit. currently controls drawings using procedure WP-115, j Rev. 6, Drawing Control and Inspection. WP-115 requires that Field copy 4' drawings must be' Quality Control (QC) verified prior to QC final inspection and that the original drawings are to be included in the plant modification package. - ENP-03, Revision 35, Section 5.4.6, requires drawings, in sufficient detail to accomplish the plant modification, be included in the modification package. (CLOSED) Unresolved Item 325/83-26-01 and 324/83-26-01, Incorporation of - Setpoint Verification Into Channel Functional Test. The philosophy of incorporating setpoint verification which ~ was utilized during the q procedure upgrade program is provided in MP-52, Standards for Preparing and Maintaining Maintenance Procedures, Revision 1, dated July 29, 1987. Section 5.4.4.2, page 107, states the following: ......where " practicable, the channel test shall verify the setpoint of the instrument channel-is within specifications. If an unreasonable impact on continual stable plant operation, personnel safety or' radiation exposure could result from verifying the t_ rip setpoint of the channel, then setpoint verification need not be included in the channel functional tests. Where setpoint verification is not part of ^he channel functional test, the signal used to trip the channel should be approximately at the setpoint yet within the Technical Specification allowable limits." (CLOSED) Unresolved Item 325/85-22-05, Failure to Follow Modification Procedure Results in Core Spray Pump Overfilling Spent Fuel Pool. The licensee issued a supplement to LER 1-85-39 on September 30, 1985, to ! describe corrective actions. This LER was inspected and closed in ] Inspection Report 325/87-06. l ' (CLOSED) Unresolved Item 325/85-28-01 and 324/85-28-01, Evaluation of Installed Process Instrument. The inspector reviewed the September 30, ! 7 ' 1985 letter from ANS Chairman, W. T. Ullrich, and the October 28, 1985 ' letter from S. R. Zimmerman on the Clarification of Measuring and Test ) l l l l ' __ _--
_ _ _ _ _ _ __ ' 1 .q. -e 4 ,
- . ;
14 ,m j 1 i Equipment' (M&TE)' versus Installed Process Instrument Calibration. These ] specifically. . delineate ANS- and CP&L's interpretation of . J ANS-3.2/N18.-7-1986, Section 5.'2.1.6.- CP&L's implementation of- ANS-3.2/N18.7-1986, Section - 5.2.1.6, conforms with this interpretation. Currently installed instrumentation contains;a calibration sticker or when the instrumentation.is.used the procedure requires verification that the instrumentation is correctly calibrated. .The inspector verified this
- through. discussions:of work practices with M&TE supervision.
No violations or deviations were identified.
9. 10CFR'21FollowupActivities(92701) (CLOSED) P2184-01. 10 CFR 21 Notification 84-01, Report on Steam Leak
- Detection.
By letter on April- 30, 1982, the licensee informed Region II
of a : deficiency in the Reactor Building steam leak-detection system. A .l postulated pipe ~ crack in the High Pressure Coolant Injection (HPCI)- j system, Reactor Core Isolation Cooling (RCIC) system, Reactor Water Clean
' Up (RWCU) system, and' main steam.line drains'outside the pipe tunnel on .these connections to the main feedwater lines will not be detected or terminated by an existing system. Due to a lack of humidity or- temperature detection in the area .of these pipes' a postulated critical , crack would' have allowed .the Reactor Building to reach a steady j temperature-of.212 degrees F at 100% relative humidity. That environment ' would have. been 'in excess of environmental qualification of equipment' important-to safety in the Reactor Building, j The licensee corrected the problem by placing isolation. valves for HPCI 1 and the steam line drains and a check valve for RCIC/ Reactor Water Cleanup (RWCU) in the main steam valve tunnel. Thus, any' leak occurring upstream of the- valves would be detected by the existing steam line break instrumentation. . The inspectors verified, through record review and previous system walkdowns, that the following modifications were complete: Date Full Final Plant Mod Title Operable 83-213 Unit 1 Main Steam Line Drains 12/16/85 83-214 Unit 2 Main Steam Line Drains 06/18/86 84-380 Unit 2 HPCI Return Isolation 05/16/86 I 84-381 Unit 1 HPCI Return Isolation 11/08/85 83-222 Unit 2 RCIC/RWCU Discharge Isolation 04/23/86 j ' 83-223 Unit 1 RCIC/RWCU Discharge Isolation 08/04/85 j ' l: . i (CLOSED) P2184-02. 10 CFR 21 Notification 84-02 concerned the use of ' Loctite 242 and other anaerobic adhesive sealants. In particular, as . delineated in IE Information - Notice No. 84-53, Philadelphia Electric
Company reported that, on November 17, 1983, two control rods had exceeded L
_i_____________m____._____z_
m, 1 - , l i l 1 15 i 'their allowed scram times at Peach Bottom Unit 3. An investigation determined the cause of the failure to be a foreign substance that. bonded ! the solenoid core plunger to the solenoid base assembly of.the scram pilot solenoid valve. The foreign material was determined to be Loctite 242. j Apparently, excess Loctite 242 had not been wiped off when it was used on the acorn nut of the solenoid housing. The material on the nut had appeared to be a solid and was thought to be cured'. Upon.the return of .j the system to service, the Loctite 242 exposed to air liquified and j migrated to the position where it eventually ca'used a . failure of pilot solenoid valves. , 1 In response to the subject Part 21 notification and IEN, the licensee, as stated in a memorandum from Tinney to Treubel, dated November 12, 1984, j provided a copy of the pertinent Loctite information to all maintenance ) foremen and provided specific training of both the Unit 1 and Unit 2 mechanical maintenance personnel in this area. ! .The inspectors also determined that . Maintenance Instruction MI-'10-40,
Rev.14, Control Rod Drive (CRD) Solenoid-0perated Scram Pilot Valves, ! Core, Diaphram and Gasket Replacement, does not specify the use of any i anaerobic adhesive sealant. l 1 (CLOSED) P2185-01. The inspectors evaluated the licensee's response l to 10 CFR 21 Notification 85-01. Notification was provided to CP&L l by a letter from Mr. A. F. Kaiser, Brown-Boveri, Inc. (BBC) to Mr. R. C. De Young, Office of Inspection and Enforcement, dated March 19, 1985. -The condition originally reported at the Seabrook Station indicated that the control wire insulation on the eight pole K-1600 or K-2000 auxiliary switch may be cut by the top edge of the dust shield when the circuit breaker is racked out to the full disconnect position with the 4 compartment door closed. This same condition could occur on the smaller size electrically operated K-225, K-600 and K-800 circuit breakers on both j the right and left side auxiliary switches. The inspectors determined through a review of the applicable auxiliary one i line diagrams (480 V system) emergency key one line diagrams, and key one line diagrams (480 V), that the licensee does not employ any mechanically operated K-1600 or K-2000 Brown Boveri circuit breakers in the 480 V E-Buses. The inspectors did note that Brown Boveri K-1600 electrically l operated circuit breakers are utilized in the unit 480 V tie buses. However, as stated by BBC in their 10CFR21 notification to j Mr. J. M. Taylor, Director, I&E, dated June 30, 1986, concerning a similar 1 problem on K-600 and K-800 circuit breakers, the K-1600 electrically l operated circuit breakers, because of their larger size, have adequate l clearance for their associated control wiring. ' (CLOSED) P2185-02. 10 CFR 21 Notification 85-02 relates to a f notification by NDT International, Inc. of an anomaly discovered during
testing of a high temperature accelerometer model No. NDT-838-1 and ITT i Cannon Connector CE9444-1002. This accelerometer is of the type utilized j i l l l l .
y. q . ! l ! ! i 16 , by the Brunswick Steam Electric Plant in the Safety Relief Valve (SRV)- k position indication system (acoustic monitoring). They are installed on l the discharge side of the piping on each SRV and are used to detect flow l through the SRV by generating a signal which is amplified and trar,smitted ( to the flow detector located in the control room. ~ j The operational anomaly identified by the accelerometer manufacturer involved a sensor / connector assembly which did not utilize a shrink sleeve environmental seal as recommended by the Raychem Corporation. As a result of the absence of the shrink sleeving on the connector, the accelerometer y l failed it's environmental qualification test. The inspectors examined the licensees Qualification Data Package QDP-58 ) for the NDT Hi-Temperature Accelerometer. This QDP indicated that , qualification testing was performed on an accelerometer assembly including a shrink tubing installation over the connector by Wyle Laboratories.. .i Qualification testing was successfully completed with no inconsistencies j identified. (CLOSED) P2185-04 The inspectors reviewed the licensee's corrective measures concerning 10 CFR 21 Notification 85-04. This event involved the reporting of excessive seat leakage, discovered during bench testing, of Rockwell 3/4" and 1" check valves. The subject check valves were purchased as part of a plant modification (PM) 84-297 involving accumulator check valve replacement for the Automatic Depressurization System and Main Steam Isolation Valve. In order to determine the acceptability of the corrective actions taken by the licensee concerning 10 CFR 21 Notification 85-04, the inspectors ~ reviewed the following documentation: o CP&L Letter Serial BSEP/85-1850, from Mr. C. R. Dietz, to Dr. J. N. Grace, dated October 18, 1985, Subject: Notification of a 10 CFR 21 Reportable Event. o 10 CFR 21 Evaluation / Notification Number 85-04, Attachment 1, Appendix A, C, and D. o CP&L Memorandum from Mr. C. Seubert to Mr. Dan Saccone, dated July 5, 1985, Subject: Potential 10 CFR 21 Reportable Details. , 4 I o Plant Modification PM-84-297, Automatic Depressurization System and Main Steam Isolation Valve Accumulator Check Valve Repacement, Unit 1. The inspectors determined that the check valves were designed to prevent I accumulator air supply loss in the event of a supply header failure. The l valves were designed to have a maximum leak rate of 0.4 cubic feet per l hour for the one-inch valves and 0.3 cubic feet per hour for the 3/4 inch l valves. These valves were apparently tested by the vendor and determined ! not to exceed this leak rate criteria as specified by the procurement l l l l I _ - - - _ . _ _ - b
, . e. : 3 (' ) 17- ' 1 jdocument.. Several valves :(bothJ1 inch and 3/4 inch category) when locally bench tested,Jfailed to meet the leak rate requirements. Specifically, a total'of twelve of the sixteen one-inch valves failed to passLa 0.4 cubic feet per hour leak rate test. . This 'high failure rate on . one-inch valves provided reason to suspect that the 3/4 inch valves might .'not: pss a < similar leak" test Two of' twenty two 3/4 inch valves were . testerf and both failed. ' 1 . On' June 27, 1985,Lfour valves'- two one-inch and two 3/4 inch.- that had , failed seat leakage testing were returned-to the vendor for examination. Leak testing'was performed by the vendor and it was determined that these- - valves could'not meet the stated leakage criteria. ' Further examination by 'the - vendor indicated a manufacturing defect which was corrected ' by_ ' 1 ' reworking the valves. [ ! . The ' inspectors also' reviewed ' the past maintenance acceptance. testing ! performed on these valves and determined that they successfully passed the l seat leak criteria as stipulated in the purchase specification. j 1 Based onLpersonnel interviews with members of the licensee's staff and a ! l review of'the above-listed-documents,-the inspecter concluded that CP&L ' took appropriate and timely action 'in providing the required 10 CFR 21 notification and that the followup. action was acceptable'. (CLOSED) P2186-01. The inspectors examined the licensee's response to.10 ! ? CFR 21 Notification 86-01. This- event involved the use. of bolting materials which were. supplied by. a non-qualified source for use in- ASME Section' III applications, . .Specifically, CP&L was notified by Valtek, Inc. on March 12, 1986, in'.a . letter from Mr. L. L. Schroeppel to Mr. L. H. Dishman, that Cardinal Industries . Products Corporation, a
non-qualified ' source, had provided studs 'and nuts which did not comply with ASME Section III, NCA 3800 requirements. ! ! The inspectors reviewed a subsequent letter from Valtek, Inc. dated April 4,~ 1986, from Mr. L. L. Schroeppel to Mr, L. H. Dishman,-where the j supplier stated that they had received additional .information from , Cardinal in which they had performed testing on each piece of starting bar .j material. .The test results as endorsed by Valtek indicate that the ) subject material meets the requirements of the material specification and
support the. originally submitted certifications for the material. Based j ' on .the above information and a review of the licensee's 10 CFR 21 j Evaluation No. 86-04 dated April 4,1986, the inspectors determined that f the bolting material, SB/B 164 Monet 400, appeared to meet the physical J and chemical requirements of the material specification as shown in the I ' vendor's certified material test report, which complied with the system specification requirements. ' (CLOSED) f2186-02. The inspectors examined the licensee's actions relating to 10 CFR 21 Notification 86-02. This event was identified to ] the licensee on July 29, 1986, by SOR, Inc. and concerned a potential
l i I i I L.. . . . l
- ___ _ . - _ - , 4 - 'f j ( 18 ] q change in allowable operating repeatability ranges of. gauge pressure switches with ' designators beginning with No. 9 8, or 1 (1. e.,- 9-N6-B45-NX-01A-JJTTX6,- 8-N6 . . . . . . , etc.). - The potential problem 1 - initially" identified at Commonwealth Edison Company /La Salle Station was >{ isolated ltoJIE qualified. pressure switches sold after January 1983. The i phenomena resulted in a repeatability of approximately 5% of' full. scale. l The original design criteria and. performance expectations were for 1% of f . full. scale repeatability. ' The inspectors reviewed the licensee's evaluation and close-out ' action d documentation.and determined that the subject switches were purchased as.
replacement material but'were never installed in either unit. Subsequent ? to the .Part 21 notification by SOR Inc., the switches were d.isposed of as , ' scrap. j ' > . ..
o - (CLOSE0) P2186-03. This. item concerns 10 CFR '21' Notification 86-03 j issued by Northeast Utilities, Millstone Nuclear. Power Station, Unit 3 to i the NRC Region I Office, May 9, 1986. The report identified a condition j in the type K600S circuit breakers supplied by BBC Brown Boveri, wherein a 4 control wire. harness containing eight wires which run from various l components in the circuit breaker to the auxiliary switch on a support j thelf on the top left front side of the breaker came into' direct contact with a_ racking gear'inside the breaker. Although the harness was wrapped with a plastic -spiral wrap, the gear teeth had worn through this material and severed the wire to the shunt trip coil. The. wire grounded on the gear which opened the fuses. It is noted that the above-described failure would'have prevented the y! associated circuit breaker-from tripping due to an over-current or faulted condition. i The inspectm reviewed the following ' licensee documentation on the I subject Part 21 notification to determine if adequate followup action had
been accomplished: ,! o CP&L Memorandum, File Q9X-XXX-AA-A651, from Mr. S. .McManus to Mr. P. W. Howe, dated July 11, 1986, Subject: Brow'n Boveri K-600/K-800 Circuit Breakers.
.o Brown Boveri letter from Mr. E. W. Rhodes to Mr. A. B. Cutter, dated , June 30, 1986, Subject: BBC Brown Boveri K-600/K-800 Circuit i Breakers, Possible Cut Wires in Wire Harness (10 CFR 21 Report). l 4 o Brown Boveri letter from Mr. A. F. Kaiser to Mr. J. M. Taylor, dated June 30, 1986, Subject: BBC Brown Boveri K-600/K-800 Circuit Breakers, Possible cut Wire Harness (10 CFR 21 Report). j The inspectors also reviewed a representative sample of the licensee's 480 volt . system control wiring and cable diagrams and auxiliary one line jl diagrams. Based on a review of the above-stated documents, the inspectors determined that Brunswick . Steam Electric Plant (BSEP) does not utilize j ' > 1 . I b- I
77., , - - - - - - - - e o l ' ?
, l -19 j , a o . ' electricallyfoperated K-600/K-800 BBC Brown Boveri circuit breakers in' l ' their 480. 'i E-Buses. The inspectors-did. determine that BSEP does utilize-
- !
four K-1600 and four K-3000 electrically operated circuit breakers on- 1 their E-Buses. . However, these circuit breakers, as. stated by BBC in their. ! Part 21 notification, the K-1600 series 'and larger electrically' operated I '. circuit 1 breakers, because of'their larger size, have adequate clearance , l' and do not fequire inspection. ! l No violations or deviations were identified.. .] q 10. Review of Special Report (90713) - ' (CLOSED). Special Report'325/NRE82-02, Automatic Start and Injection of 1A O ,A and 18 Core Spray Systems While Performing Ground Tests. This item is i described in paragraph 4.b of Inspection Report' 81-31. The inspector j ~ . verified that current procedures involving DC ground hunting requires ] verification that the subject power supplies and inverters are on after I being re-energized. Specifically, Unit 1 procedure OP-51, DC Electrical System Operating Procedure, Revision 7, dated. March 9, 1987, requires, in . s teps 8.1.B. 6.c. 3, 8.1.B. 6. d. 4,- 8.1.B. 7.c. 3, 8.1.B. 7. d. 4, 8.2.B. 20.c.3, ( 8.2.B.20.d.4, ' 8.2.B.21.c.3 :and 8.2.B.21.d.4, that the Power Available- Indicating Lights' are "0N" for the Emergency ' Core Cooling System -(ECCS) j power supplies and inverters after the applicable circuits are y re-energized. Steps 8.1.B.7.c.4, 8.1.B.7.d.5, 8.2.B.21.c.4 and 8.2.B.21,d.5, perform a similar check .for the Reactor Protection System (RPS) logics. Revision 15 of OP-51 for Unit 2 was verified to have similar requirements. 4 No violations or deviations were identified. l 11. Bulletin Followup (92703, 25529) (CLOSED) 325/79-BU-07 and 324/79-80-07,, Seismic Stress Analysis . of , Safety-Related Piping. .The-inspector reviewed the following documents to I determine whether Bulletin 79-07 requirements have been adequately j .' add:essed and implemented: ] I Parameter Report IE-141, dated January 1986, stating that Bulletin o 79-07 remains open for Units 1 and 2. The Parameter' Report stated that utility personnel responded many times between April 24, 1979 1 and November 26, 1984, indicating that a plan of action and schedule q a4 - for reevaluation of affected piping had been prepared and that
i interim-seismic capability had been evaluated. The early SER of i June 7, 1979, indicated that verification of as-builts and codes was incomplete. l l o NRC Memo from E. L. Jordan, dated February 6,1986, "IE Bulletin 79-07 Closecut." 'The following letters from Carolina Power and Light Co. to the NRC for responses to the bulletin were reviewed. 3 1 1 _ J
-_ __ _ __ _-_ . _ _ _ - . _ _ . _ _ _ s - 20 o Dated April 24, 1979,- original response. o Dated May 15, 1979, further amplification of previous letter, o Dated May 21, 1979, supplemental response. o Dated May 22, 1979, supplemental response, o Dated May 29, 1979, supplemental response. o Dated June 4, 1979, supplemental response. The following letter from the NRC to Carolina Power and Light Co. for the bulletin were reviewed: Dated June 7,1979, stating acceptable seismic response combination o techniques and acceptable results were obtained and concluded that the requirements set forth in Bulletin 79-07 will be adequately satisfied to allow resumption of power operation. However, as-built verification of all lines and supports and completion of code verifications will be required. The following project procedures for verification and implementation of the bulletin were reviewed: o No. CPL-PP-01, Pipe Stress Re-analysis Under IE Bulletins 79-02, 79-07, and 79-14, dated May 17, 1985, Revision 0. o No. PP-9527-062-001, Procedure No. 062, Pipe Stress Re-Analysis, l dated January 19, 1987, Revision 1. l Also reviewed was the licensee's Facility Automated Commitn.ent Tracking System (FACTS) for status of all items for Bulletins 79-07 and 79-14, , which showed that all but one item remains to be completed to close out Bulletins 79-07 and 79-14 for Unit 2. This action should be completed at the end of the next outage scheduled for January, 1988. Unit I has been completed. The open item is being tracked under the Bulletin 79-14 program. The inspector reviewed the above documents and determined that the requested actions of the bulletin have been adequately addressed. This bulletin is considered closed for the purposes of the Regional Inspection Programs. The remaining item for this bulletin is being tracked and will be completed under the Bulletin 79-14 program. Bulletin 79-14 will be reviewed after all modifications have been completed end the final response by the utility has been reviewed during a la+er inspection. (CLOSED) 325/84-BU-02 and 324/84-BU-02, Failures of General Electric Type HFA Relays in Use in Class IE Safety Systems. The inspector reviewed the licensee's response dated July 13, 1984, and the schedule submitted on February 14, 1985. The inspector verified that procedures were written to L--_-_---___-_______ __ _ _ _ _ __ _ _ _ . _ _ _ ____
4 p- " ' - - - -7- j , ,
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, , ,
- performiinitial jsu, ,cillance and- inspections ~as- required. IThe licensee 1
.." , 'y acomittedsto replace the safety'-related^ Unit 1- HFA. relays during refuel 4 ~ ' scheduledLfor March,11985 and Unit 2 relays during refue12 6 scheduled for~ , s Z ' April,1986 For the identification of normally energized relays,1Sargent ~ ' gm - a:LundyEtheL 11_censee's contractor. assumed,-in document SL-4295, that all- ', 't .; relays 0 that are ' energized: withL the. plant at full power operation and, " ' ' .. Econnected ..to . the agrid Lare' classified as "normally energized.": The , ~
t1 .. inspectors reviewed SL-HFA-2, Revision 1, BSEP Unit 2 HFA Relay Replacement 1 , Methodology l Document, signed as . completed ;o'n March 12, 1986,7 which . ' ,4 ,i
- documented completion of the rep)lacementfofL the;old ' relays with _the new
~ m;
- Century ' series (green nameplate . - 'Likewise, "SL-HFA-1, Revision.1 for !
? . - ' .. Unit I was completed and signed'out'on' August 20,'1985. .
- .
, The cresident' inspectors' had previously. observed HFA~ relay- replacement 1 ' 1 -througho'ut the ~last 2 years , including Ja'.' review of procedure MI-16-0350,; - - Replacement of Ge~neral Electric HFA Relays. Selected walkdownscof the- . control room ;backpanels during this . report" period .and after . relayj ,J " - replacement showed no .normally energized. black nameplate or- old style HFA - d i L relays' ! , , Further, as part of- the licensee's project plan for Equipment Data Basei xSystem-Development,. Revision _4, dated January 21, 1985, a re-review of.the m ' HFA relay: list'will be performed as data is: entered into the. data base. J T' licensee expects to complete this re . review by December,1989. s. 'No. violations or deviations were identified, d . .12.. Engineered Safety Features (ESF) System Walkdown (71710)
'On September 29, 1987., an inspection was performed of the Control Buildi.ng l 1 Emergency?Ai.r; Filtration- (CBEAF) . System.; The inspection . included a verification:of: . proper valve and damper position, power availability-to ~ ~ u' the emergency recirculation . fans, instrument air availability to dampers, 7 chlorine and radiation detectors in' service, integrity of duct work and ~ , proper physical appearance of major components. The ,"B" . train filter : .i housing support steel base was observed to have three anchor bolts in. a
- l
row out of six on one- side with' the nut not fastened down. There was ' approximately one inch of thread between the nut and its' mating surface. The foundation. bolting on the other side was observed to have one nut not { tightened down and three others with less than full ' thread engagement.
i LThe "A" train foundation steel was found' to have one . nut not tightened L" down, two anchors bent and two anchors installed crooked. The Emergency -Recirculation Fans (ERF) were observed to have their vibration dampers not ' installed in accordance with drawing FP-4196, sheet 8. The vibration 3 dampers are required to' be adjusted such that a spring force retards vertical movement. - Of the four vibration dampers on ERF "A", two were 4 essentially rigged supports and one was bottomed out. Similar types of problems were observed on ERF "B". ' . . . ) . , t '!' '
n; - y - -- -- -- - - - 4 - - ' - - - - - , v- 93 g h , , , a 5 22 l ' ' j s n , . . .. ~ . ] . s
- The licensee evaluated.the~ conditions and concluded the' operability of.the
, , -systemLwas not' effected.. Work request 87-BFGG1"and'87-BFGJ1.were issued i to? correct:the conditions on ERF'"A" and "B"',' respectively. Engineering- is evaluating!the' priority for. correcting the. foundation bolting.
- The system engineerfindicated that to her; knowledge neither condition h'ad
" been previously evaluated; The' foundation bolts appeared to be triginal construction deficiencies.- The vib' ration damper adjustments appear to be ] ~ ' attributable to one'or more'of the following: : vibration, maintenance or
original' construction ' . At least threelof the' observed conditions cannot 3 ~ ? be : attributed to vibration. : Review of work requests completed in the last- q ' year- revealed. no, work which -should- have caused the conditions. These 1 . types of. items have recently'been addressed'in. Inspection ~ Report 87-17 and l -t by the- licensee's response, dated ' August 27,1987. 'Since corrective' . action for. as previous' violation has not yet .been completed, no notice of violation is'being' issued. g i No: violations or deviations were identified. e 13. Inservice Inspection Program. Hydrostatic Tests (73755) . The glicensee Tcompleted . a' review..of .. their ASME Code, Section XI,
- Hydrostatic Test Program and found ' additional problems.
Region II" issued Inspection Report 325/87-30 and 324/87-30:on September 30, 1987, listing- j three.. violations .relatedito hydrostatic testing. per Section XI. After a completion of, the inspection 'on September:4,: 1987, the licensee began a . re-review 'of their- hydrostatic test. program. . The review,' completed on - September 22,41987, ' identified four additional sections of piping systems .that' were not, adequately ' tested. The problems as reported' to the - ,, - inspector were: a. Portions off thel 2B RHR Service Water (SW) loop had been pressurized under plant modification 85-063 but VT-2 inspections were documented on the welds only.. b.- Unit 2' SW ' lube water supply piping had not been . pressurize'd or inspected. c. Unit 2 fuel pool cooling skimmer surge tank and associated piping, an unpressurized system, had not been inspected. I 1 d. Control Rod Drive (CRD) insert and. withdraw lines from the scram 1 valves inte the drywell, but not under vessel, were not inspected per VT-2 requirements. The licensee developed Justification for Continued Operation (JC0) for all four cases. respectively: i a. Inspectors of piping welds would have seen leaks and no current evidence of leakage exists. ! !
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( .. > e / l l t 23 l b. Adjacent piping of the.same material (90% Cu - 10% Ni) has no leaks and piping was walked 'down during normal system operation. .! c. Unpressurized system - no evidence of leakage from pool. . ~ d. No -leakage was identified during VT-2 inspection of adjacent piping.
A relief request was submitted to NRR on this item, a Class I system, l 1 on September 28, 1987. The licensee agreed, during the exit, that they would include their actions.on the above problems in their response to the violations listed in report 87-30. No violations or deviations were identified. 14. Diesel Generator Thermostatic Control Valve ' While inspecting the current valve line-up for the DG lube oil cooling and jacket water systems, the inspector found, on September 29, 1987, that the jacket water cooler outlet Thermostatic Control Valve (TCV), MUD-TCV-2155, " was not in the thermostatic position. The TCV is ' designed to automatically cor. trol jacket water temperature to the combustion air intercooler to 110*F during engine operations; and thus is maintained in the thermostatic position. The licensee confirmed that TCV-2155 was four turns toward the bypass pcsition. At the time of the inspection, the No. 2 diesel jacket water and lube nil' temperatures were adequate to support operation of the DG. However, by closing TCV-2155 four turns toward bypass, the valve's motion would be partially limited when going toward the cooler position during diesel operation. ,i Valve TCV-2155 was taken out of the thermostatic position during performan::e of PT-12.2.b, No. 2 Diesel Generator Monthly Load Test, Revision 29. Step 7.1.22.1 requires the operator to adjust TCV-2155 to
maintain a combustion air intercooler temperature between 105 degrees F and 115 degrees F. However, under the Section 7.4.3, System Restoration, the TCV position was not listed as requiring a check. Temporary Revision
87-480, initiated on September 30, 1987, added the TCVs to the restoration ' line-up. The procedures for the other DGs were also corrected. Failure to include the TCVs in the restoration line-up is a failure to establish an adequate surveillance procedure which is contrary to Technical Specification 6.8.1.c. This is a Violation: Inadequate DG Surveillance Procedure - TCV Not in Restoration Line-up of PT-12.2.b (325/87-31-01 and 324/87-35-01). l . _ _ }}