IR 05000413/1992027
| ML20128B180 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 11/23/1992 |
| From: | Belisle G, Hopkins P, William Orders, Zeiler J 63 NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20128B150 | List: |
| References | |
| 50-413-92-27, 50-414-92-27, NUDOCS 9302020418 | |
| Download: ML20128B180 (16) | |
Text
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f Report flos.: 50-413/92-27 and 50-414/92-27 Licensee: Duke Power Company 422 South Church Street Charlotte, f Docket flos.: 50-413 and 50-414 License flos.: NPf 35 and NPF-52 facility Name: Catawba Nuclear Station Units 1 and 2 Inspection Conducted: October 11, 1992 - November 7, 1992
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Projects Section 3A -
Division of Reector Projects SUMMARY Scope: This routine, resident inspection included but was not limited to a the following areas; plant operations review; review of Catawba design pertaining to problem documented in McGuire PIR 0-M92-0406 concerning Auxiliary Feedwater System; Standby Makeup Pump operability verification; review of non-conservative mass and energy release data for Main Steam Line Break Analysis; review of applicability of Technical Specification 3.0.2 to Diesel Generator removal from service; review Of Reactor Coolant Pump 2D seal degradation; review Of steam generator equipment hatch control; emergency drill evaluation; quality assurance organization review; surveillance observations; maintenance observations; review of defective weld on steam generator tube plug; review of Licensee Event Reports; and followup on previous inspection finding PDR ADOCK 05000413 G PDR
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Results: In the areas inspected, two non-cited violations were identifie One non cited violation involved a failure to follow procedures regarding the control of a steam generator enclosure hatch (paragraph 4.) and the second non cited violation involved an inadequate QA weld inspection on a steam generator tube plug (paragraph 8.c).
One Unresolved lien, was identified involving design engineering's evaluation of the operability of the Standby Makeup Pumps with under-pressurized pulsation dampeners. (paragraph 8.b(2)).
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REPORT DETAILS l
- Persons Contacted Licensee Employees S. Bradshaw, Shift Operations Manager
- J. Cox, Acting Regulatory Compliance Manager J. Forbes, Engineerly Manager S. Frye, Operations Support Manager 1 E. Geddie, Operations Superintendent
- 1. Harrall, Safety Assurance Manager M. Hazeltine, Compliance
- J. Lowery, Compliance '
- W. McCollum, Station Manager K. Seasely, Compliance M. Tuckman, Catawba Site Vice president Other licensee employees contacted included technicians, operators,.
nechanics, security force members, and office personne NRC Resident Inspectors
- W. Order P. Hopkins
- J. Zeiler
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- Attended exit intervie . Plant Status ]
Unit 1 Summary ,
Unit 1 began the report period in mode 5, completing preparations for- ,
unit restart from refueling outage E006. . ihe unit achieved criticality- '
on October 18, completed reactor physics testing-on October 20, and was placed on line the following day. The unit achieved 100 percent power on October 24 and operated at or near full-power for the' remainder of the report period with no major problems. The unit ='was slightl restricted in' power from October 28, through the end of the report ._
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period due to reactor coolant flow calculation uncertainties attributed
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to temperature gradient anomalies in the reactor coolantLcold legs.
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L Unit 2 Summary
l~ Unit 2 operated at or near full power for the entire report period with- s j' .no major problem ; Plant Operations Review (71707)
] The inspectors reviewed plant operations throughout the report 3 oeriod to verify conformance with regulatory requirements,
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Technical Specifications (TS) and administrative control' ;;
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Control Room logs, the Technical Specification Action Item Log, '
and the Removal and Restoration (R&R) log were routinely reviewe Shift turnovers were observed to verify that they were conducted
- in accordance with approved procedures. The complement of i; licensed personnel on each shift inspected, met or exceeded the requirements of Technical Specification Daily plant status meetings were routinely attende Plant tours were performed on a routine basi The areas toured included but were not limited to the following
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Turbine Buildings Auxiliary Building Units 1 and 2 Diesel Generator Rooms Units 1 and 2 Vital Switchgear Rooms '
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Units 1 and 2 Vital Battery Rooms Activities prescribed in the following operations procedures were reviewed in detail:
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OP/1/A/6250/01 Condensate and Feedwater System Pumps Start Up -
OP/1/B/6300/01 Turbine Generator Start Up OP/2/A/6450/17 Containment Air Release Addition System
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Operation During the plant tours, the inspectors verified by observation and interviews that measures taken to assure physical protection of _,
. the facility met current-requirementsc Areas inspected included-
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the security organization, the establishment and maintenance of gates, doors, and isolation zonas-in the proper conditions, and that access control badging ware proper and procedures followe ,
In addition, the areas toured e re observed for fire prevention .
and protection activities and radiological control.practic'es' The .-
inspectors also reviewed Problem Investigation Reports (PIRs) .t'o determine if _ the licensee was appropriately documenting problems and implementing corrective' action Review of Catawba Design Pertaining to Problem Documented In McGuire PIR 0-M92-0406 Concerning' Auxiliary Feedwater (CA) System '
. During this report period, the inspectors evaluated'a problem ,
identified at the McGuire Nuclear Station during a design basis '
study concerning CA; System Pump t,perability following a hypothetical seismic event that severed the normal suction lin '
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As part of their< evaluation, the inspectors reviewed a drawing; '
which provided an overview of the relative elevations for all' the- -
suction sources to,the CA Pumps, including the " assured source"-
- which is supplied by the Nuclear Service Water (RN) syste .
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Pressure switches and associated controls are designed to detect a- ,
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loss of source (either pipe break or source depletion) and transfer pump suction to RN to maintain-continued operabilit The residents also reviewed an engineering analysis which postulated accident scenarios occurring under both seismic and
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non-seismic conditions as well as a related security scenari The licensee's analysis indicate: . 'iat Cat awba's Auxiliary.-
Feedwater System was capable of sustaining a wide variety of suction source failures, was capable of maintaining adequate suction flows to the various required motor driven and turbine driven pumps to assure continued system operation consistent with their design basis requirements and did not appear to be vulnerable to the problem identified at McGuir Standby Makcup Pump Operability Verification
During the report period, the inspectors were alerted to a problem .
l discovered at the McGuire Nuclear Station involving one of their Standby Makeup Pumps (SMPs) which was found to rotate in the wrong direction due to the incorrect wiring of the moto These !
positive displacement pumps are part of the Standby Shutdown System (SSS) and function to provide reactor coolant pump seal-injection during.an SSS event. The inspectors questioned the i licensee to determine if they had verified that Catawba's -SMPs --
were operating correctly. Catawba personnel indicated that their counterparts at McGuire had contacted them when the problem was identified and the Unit 1 SMP had been tested and confirmed to b operating correctly- but the Unit 2 pump had not been tested. Due to heightened interest over. the issue, the licensee decided to run 1 both Units' SMPs. On October 23, both pumps were verified to be ;
operating ccrrectl ;
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As part of their inspection, the inspectors reviewed NRC Information Notice 91-27, Incorrect Rotation of Positive Displacement Pump, which was issued April 10, 1991. The Notice discussed an incident ~ in which a positive displacement pump was discovered to be rotating.in the reverse direction due to the-
. incorrect wiring of the motor. The concern focused.on the lack of adequate pump lubrication if the pump rotates backwards, t
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The inspectors reviewed the licensee's )rocessing of Information Notice 91-27. It was determined that tie licensee had discounted -
the possibility of such an event-at Catawba since maintenance procedures require-independent verification of the lifting'and re- }
termination of motor leads. The licensee failed to consider the
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possibility that the pumps may have been mis-wired during initial'
installation, as was the case at McGuire. As an added safeguard to ensure that similar problems are not encountered, the licensee- ;
. plans to revise the quarterly-pump operation procedures to -;
incorporate pump rotation' check ,
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. Review of Non-Conservative Mass and Energy Release Data for Main Steam Line Break Analysis During the report period, the inspectors evaluated a problem discovered at Tennessee Valley Authority's Watts Bar Nuclear Plant involving the use of non-conservative mass and energy release data for the Main Steam Line Break (MSLB) accident analysis. As a result of this non-conservatism, the pressures produced from a MSLB could exceed the structural design limits in the main steam valve vaults (MSVVs) challenging their structural integrit Failure of these structures could potentially damage equipment or piping housed in the vaults associated with the feedwater and auxiliary feedwater system The Catawba design does not employ an enclosed MSVV concep Rather, the similar structures are open to the environment at the top, minimizing the possibility of structure pressurizatio Thus, this problem is not applicable to Catawb Review of Applicability of Technical Specification 3.0.2 to Diesel Generator Removal From Service On October 19, members of the licensee's compliance staff and the inspectors discussed instances in which the licensee was applying TS 3.0.2 to the action statement of TS 3.8.1.1. when an emergency diesel generator (D/G) was removed from service for planned maintenance. The Limiting Condition for Operation (LCO) for TS 3.8.1.1 requires that two D/Gs be operable in Modes 1-4. TS 3.8.1.1, Action Statement c states that if a D/G is inoperable, the licensee is to demonstrate the operability of the AC offsite power sources within I hour and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter. TS 3.0.2 states that if the applicable equipment is _
restored prior to the expiration of the specified time intervals of the LCO Action statement, completion of the Action requirements is not require The particular instances discussed-in which the licensee has applied TS 3.0.2 involved the removal of a D/G from service in order to perform short duration planned maintenance, e.g.,
periodic air rolls, to ensure that water had not accumulated in the cylinders. During these activities, the D/G was out of service for approximately 5 minutes. Therefore, if one were to apply TS 3.0.2, the AC offsite power surveillance would not be performed since the D/G would be returned to service within I hou The inspectors discussed this position with personnel in the NRR Technical Specification Branch. Based on those discussions, it was concluded that TS 3.0.2 is applicable for these circumstance l l
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i 5 Reactor Coolant Pump 20 Seal Degradation The inspectors have been monitoring the gradual degradation of the Number 1 seal on Reactor Coolant Pump (RCP) 2D since April lhe leakof f flow rate has been gradually increasing over the period but the seal has exhibited no erratic or unpredictable perturbations which would indicate a serious instabilit Discussions between the licensee and Westinghouse have been on-going since the trend was identified. Based on the evidence which indicated that serious instability of the seals was not occurring, on October 1, Westinghouse concurred that it was reasonable to extend the operational limits for the No. I seal leakoff flow rate from 6.0 to 7.0 gpm. This additional margin is applicable only to RCP 2D until the next refueling outage scheduled for January 199 In the event that leakof f reaches this new limit, the unit will be shutdown. By the end of the report period, RCP 2D No. I seal leakoff had reached 6.0 gp No violations or deviations were identifie . Steam Generator Equipment Hatch Control (71707,62703)
On October 15, 1992, Unit I was in Mode 3, with the reactor coolant system at full temperature and pressure. During the heatup to Mode 3, a small leak was identified in the gasket area of t. manway cover on Steam Generator (S/G) 10. The licensee determined that the leak could be repaired at the existing plant conditions by injecting the gasket area of the cover with a sealant compound. Two Work Orders (W0s) were initiated to perform the maintenance. The first, WO 92078419-01, was for the removal of the Steam Generator 10 enclosure equipment hatc The second, WO 92078420-01, was for the actual leak repair activit The equipment hatch was to be removed to allow personnel and equipment access to the manway cover and was to remain off during the activity to assure a safe work environmen Operations approval to commence work for WO 92078419-01 (enclosure equipment hatch) was obtained on October 15, at approximately 8:15 The responsible operations personnel; however, failed to recognize that the containment divider barrier TS (3.6.5.5) action statement was impacted by the W This resulted in the activity not being entered in the Technical Specification Action !. tem Log (TSAIL) which is used to track and control the entry of TS action statements. On the following morning, at 12:50 a.m., maintenance personnel removed the equipment hatch in preparation for performing the leak repair. The hatch remained open, without the knowledge of the operations staff, until the following day at approximately 11:00 a.m., when an Operations Shift Manager detected the error, and instructed maintenance personnel to replace the hatc Technical Specification 3.6.5.5, Divider Barrier Personnel Access Doors And Equipment Hatches, Action Statement a), requires that in Modes 1-4, the S/G enclosure equipment hatch be closed within one hour after
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opening or the plant must be in Mode 3 in the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and then Mode 5, Cold Shutdown, in the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. When the hatch is i, open, it creates a direct flowpath from lower to upper containment, bypassing the ice condenser.The equipment hatch was open for approximately 10.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. Since this did not exceed the TS action limit, a viola, ion of TS 3.6.5.5 did not occu ,
The inspectors reviewed Station Directive 3.3.7, Work P.equest Origination, which provides the guidance for. initiating and ap) roving station work orders. Section 4.10 describes how work orders witch affect TSs must be handled. Item 4.10.1 requires that a work request that places the plant in a TS action statement must be entered in the TSAIL Logboo The Senior Reactor Operator in the Work Control Center is responsible for reviewing W0s to determine their impact on TSs and for ensuring that the WO is sent to the Control Room SR0 for the appropriate TSAIL entr This event is considered to be a violation of the requirements of-TS !
6.8.1, for failure to follow Station Directive 3.3.7. After review of the circumstances relative to the issue, the inspectors determined that the criteria specified in Section VII B(1) of the NRC Enforcement Policy'
were satisfied, in that, the violation was not wilful, had low safety significance, and, appropriate corrective action was initiated prior to
- the end of the report period. For those reasons, this issue is-
-documented as Non-Cited Violation (NCV) 413/92-27-01: failure to follow
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Procedures Regarding Control of Steam Generator Equipment Hatch Openin One NCV was identifie . Emergency Drill (82701) .
The licensee conducted an emergency drill on November 4,1992. The resident inspectors participated in the drill and observed the activation, staffing.and operation of the emergency organization-in the Control Room (plant simulator) and Technical Support Center. Even though a large percentage of upper level management was not on site, the TSC was staffed and activated in an expeditious manner using the alternate emergency response organization staf It should be noted that the licensee routinely incorporates their on-site simulator in their emergency drills to add realism to the. dril This is considered a positive programmatic attribute.- ,
No violations or deviations were identified.- Quality Assurance Organization Review (40500)
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During the report period, the inspectors performed a limited review of
- the licensee's Quality Assurance (QA) organization to determine the effectiveness of the QA program. ,-m-e ei4- - . = = a v- - - *-r-T-- w -
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ine QA organization is divided into three distinct areas. The offsite or corporate organization, called the Quality Verification Department, is responsible for the Nuclear Safety Review Board (NSRB), routine plant audits, and, the qualification of Vendors and Suppliers. The manager of this organization reports directly to the Executive Vice Prasident (VP)
of Power Generatio The NSRB is composed of 10 members, with a variety of expertise, and functions to provide an independent review and audit of important plant activities. The Quality Verification Manager of Audits is responsible for performing planned and periodic internal audits for all phases of station activities. This group is composed of approximately 10 member The Quality Verification Manager of Vendors is composed of approximately 10 members and is responsible for evaluating and approving vendors which supply equipment and service __
Another corporate level QA organization is the Nuclear Services Safety Assurance group, the manager of which reports directly to the Senior VP of Nuclear Generation. This group performs reviews every 6 months for each Duke station to assess thei performance with regard to Nuclear Safety. The results of these ".ategrated Safety Assessments" are presented to upper management for their revie The onsite QA organization, called the Catawba Safety Review Group (CSRG) is responsible for monitoring the daily and overall performance of the plant through the investigation of significant plant occurrence They also perform in-plant reviews used to evaluate effectiveness, identify problems, and develop recommendations for improvement, in addition, the CSRG verifies that the station QC program requirements are properly implemented. The manager of the CSRG reports directly to the onsite Safety Assurance Manager. The CSRG is composed of 14 member The onsite Quality Control (QC) program is composed of inspections that ensure Inservice Inspection (ISI) requirements are met, verify compliance with cleanliness criteria, verify compliance with instrument -
and maintenance procedures, and, verify conformance of materials, parts, and components received at the station. Under the 1991 re-organization, QC inspectors report directly to the manager of each craft area. There are 24 inspectors assigned to the mechanical maintenance, welding, and civil groups. There are seven QC inspectors assigned to the Instrumentation and Electrical department. The personnel performing these inspections are examined and certified in their particular fiel An offsite organization, the Technical Services Group is responsible for furnishing QC inspectors for performing and documenting Non-Destructive Examinations (NDE) at each statio The QA crganization was completely re-structured early this year. There has been a good transition fron the old QA organization to the presen A brief review of the audits which have been performed under the new organization, revealed that they were thorough and performance base There appears to be better trending and a higher visibility of problems identified at the site under the new organization due in part to the fact that the Manager of the Safety Assurance group reports directly to the site Vice President. Also, under the new program, site self assessments are sent to corporate Safety Assurance where they are
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8 j packaged with those of McGuire.and Oconee and sent to NSRB for evaluation and necessary actio No violations or deviations were identifie l
' Surveillance Observation (61726) ! General i During the inspection _ period, tiie inspectors verified plant operations were in compliance with various TS requirement Typical of these requirements were confirmation of compliance with the TS for ceattivity control systems, reactor coolant systems, safety injection systems, emergency safeguards systems, emergency
- power systems, containment, and other important plant support
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system The inspectors verified thatt surveillance testing was '
performed in accordance with approved written procedures, tes instrumentation was calibrated,_ limiting conditions.for' operation .
were met, appropriate removal and reatoration of the affected I equipment was accomplished, test resalts. met acceptance criteria and were reviewed by personnel other than the individual directing the test, and any deficiencies i_dentified during the testing were i
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properly reviewed and resolved by appropriate management personne Surveillance Activities Reviewed 1) The inspectors witnessed or reviewed the following' ;
surveillances: ;
PT/1/A/4150/01A Reactor Coolant Leak Test PT/1/A/4150/13B Calorimetric Reactor Coolant flow Measurement PT/1/A/4200/01N Reactor Coolant' System Pressure Boundary Valve _ Leak Rate Test . -
PT/1/A/4250/03E CA System Discharge Control Valve *
Throttling Procedure Flow Balance PT/1/A/4600/01 Rod Cluster Control Assembly' (RCCA)
l Movement; Test'.
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PT/1/A/4600/02A Mode 1 Periodic Surveillance Items '
i PT/1/A/4600/020 Periodic Surveil.1ance item; PT/1/A/4600/16 -- Surveillance for Unit 1 Start U .
PT/1/A/4600/19C- -Premode_3 Periodic-Surveillance-Tests-
.lP/2/A/3222/00B Analog Channel Operational Test PT/2/A/4200/06C Containment Spray Valve Lineup -
Verification . , .. .
t PT/2/A/4350/02A-_ Diesel Generator 2A Operobility Test PT/2/A/4600/02A Mode 1 Periodic Surveillance Items PT/2/B/4250/05 Generator Core Monitor Monthly Test -,
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2) On October 24, Unit I reached 100 percent power following the completion of the End-of-Cycle 6 (EOC6) refueling outage. On October 27, testing to measure the Reactor Coolant (NC) total flow was initiated using PT/1/A/4150/13 NC Flow is calculated by a precision calorimetric heat balanc The results of this test indicated that the total NC flow rate was less than that required for 100 percent power operation. Based on these results, on Octoacr 28, power was reduced below 98 percent in order to comply with TS 3.2.5 which requires a minimum NC total flow rate of 385,000 gpm above 98 percent power. Over the next several days, a second test was performed which confirmed the first test's result Based on the licensee's investigation of the heat balance data, it is believed that the effects from NC system hot and celd leg streaming is causing the low flow indication problem. The licensea discussed with Westinghouse Corporation the possibility of allowing the use of the spare NC system cold leg RTDs for the NC flow calculation. By averaging the active and spare cold leg RTDs, the calculated value of NC flow would increase above the TS limit. The inspectors will continue to monitor the licensee's progress in resolving this proble No violations or deviations were identifie . Maintenance Observations (62703) General Station maintenance activities of selected systems and components were observed /re"iewed to ensure that they were conducted in accordance with the applicable requirements. The inspectors verified licensee conformance to the requirements in the following areas of inspection: activities were accomplished using approved procedures, and functional testing and/or calibrations were performed prior to returning components or systems to service; quality control records were maintained; activities performed were accomplished by qualified personnel; cnd materials used were properly certified. Work requests were reviewed to determine the status of outstanding jobs and to assure that priority was assigned to safety-related equipment maintenance which may affect system performanc : )
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b. Maintenance Activities Reviewed 4 (1) The inspectors witnessed or reviewed the maintenance activities associated with the following Work Orders (W0s): ;
WO 92079577-01 Verify Charge on Unit 1 Standby ,
Makeu) Pump (SMP) Dampers (Octo)er 23)
WO 92083266-01 Verify Charge on Unit i Standby I Makeup Pump (SMP) Dampen -
(November 5)
(2) Technical Specifications require that operability of the SMPs be verified through quarterly pump operation check Prior to running the pumps, pressures in the pump's gas ,
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filled suction and discharge pulsation dampeners, are .
checked as part of preventive maintenance. Both unit's SMPs were run on October 23, in order to verify that they were rotating in the correct direction (see paragraph 3.c for ,
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details on reverse rotation problem). Prior to running the-pumps, maintenance personnel checked the dampener pressures under WO 92079577-0 Pressure in three of the four daupeners was lo The inspectors. questioned how frequently pressure in the dam)eners had been found low in the past and if evaluations.had acen performed to determine the minimum pressure necessary for. pump operability. Following the :
review of dam)ener work history data, both the inspectors >
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and licensee 3ecame concerned when it was discovered that-L low suction dampener pressures were found during the last four tests on the Unit 1-SMP.= Based on these concerns, on.
l November 5, the Unit 1 SMP dam)eners were checked ~under W 92083266-01. The results of t11s check verified that the suction-and dircharge dampener pressures were within acceptable limit ,
The licensee initiated Problem Investigation Report No. 0-
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92 0812 to address the problem of decreasing dampener ;
pressure and to determine the minimum dampener pressure !
necessary fcr pump operability. . In the: interim,4the dampeners will be checked on an increased frequency to ensure that pressures are maintained.within acceptable .
-limits.
L This issue' will be carried as-an Unresolved-Item (URI)-
p pendingicompletion of the:llcensee's analysis of _ dampener; . 1 L pressure operability requirements. This: item is documented- :
n' as URI 413, 414/92-27-02:: . Review Licensee Analysis:of SMP- ,
L Dampener: Pressures.- ,
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I 11 c. Defective Weld On Steam Generator Tube Plug Catawba Unit i entered the E0CG refueling outage with routine steam generator testing and maintenance scheduled. Early in the outage, eddy current test results revealed that the S/Gs had experienced significant outside diameter, stress corrosion cracking (ODSCC) degradation. The scose of the S/G work quickly expanded. The flaws detected during t ais expanded effort would have translated into an extended re) air outage if the licensee were to have applied the 40% throug1 wall repair limit methodology delineated at that time in TS 3/4. However, the licensee was granted a TS change which basically modified Catawba Unit 1 S/G repair methodology to one which relies on an eddy current voltage signal strength as opposed to the traditional percent through wall ,
criteri Even with the TS change, the licensee sleeved 108 tubes and plugged another 254 collectively between the four steam generator To support the technical basis of the revised repair methodology, sections of three tubes were removed from the C and D steam generators in order to extract additional qualifying in onation pertaining to the ODSCC phenomeno A plug was welded 4.no each ,
of these three tube location On October 3, with the unit in Mode S preparing to restart, chemistry samples of the secondary system indicated that there was a primary to secondary leak. Subsequent investigation revealed that the tube plug located in position 9-76 in the C steam generator was leakin It was later determined that the weld on the plug was faulty in that there was lack of fusion on the tube sheet side of the weld. Ultimately, the defective plug was removed, a new plug was welded in place, and all three plugs were -
visually inspected by Q In analyzing the circumstances associated with this ev.!nt, the inspectors noted that the initial post weld visual inspection required by ASME Section XI, was performed remotely using a video camera on the end of a hand held extension device. The inspection was inadequate in that the weld on plug 9-76 in the C S/G clearly exhibited lack of fusion which was detectable with the naked ey This was proven when a QA inspector performed a S/G entry on October 9, and easily identified the proble Although remote visual inspection is allowed (ASME Section XI, IWA-2121 (c)), it is required that the remote inspection be at least equivalent to that attainable by direct visual examinatio CFR 50 Appendix B, Criterion X, inspection, requires in part that a program for inspection of activities affecting quality be established and executed by or for the organization performing the activity to verify ccnformance with the documented instructions, procedures, and drawings for accomplishing the activity, l
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Examinations, measurements, or tests of material or products j processed are required to be performed for each work operation i where necessary to assure quality. Impilcit in those requirements !
is the requisite that the inspections be adequate to ensure ;
qualit t Contrary to the above, the remote visual QA inspection performed on September 12, 1992, on the weld of tube plug 9 76 in the Unit 1 ,
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C S/G was inadequate to assure quality in that the weld was defective but was not detected. 1he defective weld led to a ;
primary to secondary leak when the systems were re-pressurized in .;
the return to servic In as much as the NRC wants to (ncourage~and support licensee'
initiative for self-identification and correction of problems 'and in consideration of the fact that the proposed violatica qunlifies under the provisions of 10 CFR 2, Apaendix C, Section Vil B, in-.- .
that, the violation was not wilful, lad low safety significance,
and, appropriate corrective action was initiated prior to the end of the report period, this issue is documented as NCV 413/92-27 ,
03: Inadequate Post Weld Inspectio One URI and one NCV was identifie . keview of Licensee Event Reports (92700) I
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The Licensee Event Reports (LERs) listed below were reviewed to- >
determine if the information provided met NRC requiremer,ts. The determination included: adequacy of description, verification-of .
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compliance with Technical Specifications and regulatory requirements, corrective action taken, existence of potential generic problems, :
report'ing requirements satisfied, and the relative. safety-significance' :
of each even ; (Closed) LER 413/90-14: Technical Specification 3.0.3 Entry'Due '
to Ventilation Heater Controller. Inaccuracie '
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The inspector roted.that several of the planned corrective action 1
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items which involved-axtensive ventilation system design' reviews
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had been incorporated into the licensee's Design Basis ,
Reconstitution (DBD)_ Program. To date, the only ventilatio'n '
system for which the analysis has:been completed is the Annulus :
l' Ventilation System. Since it is expected that this-particular- -
- issue will be adequately evaluated during DBD ' analysis and' those '
!? action-items are being tracked by the. licensee's commitment _ list =,.
! this LER is being closed.
l 1 (Closed) LER414/90-02:; Technical Specification. Violation for an l Inoperable Manual Containment-Isolation Valve Due to Management =
- - Deficiency -
No violations or deviations'were identified.
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1 Followup on Previous inspection findings (92701 and 92702) ,
, (Closed)-Ifl 413,414/90-32-02: Review Licensee Resolution of EMF-34 Problem The inspectors reviewed the documentation associated with tho'
design and 50-59 review of a modification which will move the :
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automatic functions currently associated with radiation effluent :
monitor EMF-34, $ team Generator Water Sample Monitor to EMF 33, Condenser Air Ejector Exhaust Monitor. The licensee is in the i process of installing the modification and drafting the necessary Technical Specification change '
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, (Closed) Violation 414/90-03-01: Failure To Take Prompt And-Timely Corrective Action On Inoperable Unit 2 ABFX Syste The licensee responded to this violation by letter dated March 30,-
1990.
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The inspectors have reviewed the licensees corrective. actions in order to determine their adequacy to prevent recurrence. . The corrective actions appeared adequate and for the most part have been implemented. A committed long term corrective' action, a -
Design Basis evaluation (DBD) of the Auxiliary Building Filtered-Exhaust (ABFX) System, is to be performed to assess the possibility of foreign material-intrusion into the system and its impact. This is scheduled to be completed by April 1, 1993, (Closed) Violation 414/90-19 05:- Failure To follow instrumentation Procedures, ,
. Tha licensee responded to this violation by letter dated Septamber .
27, 199 The inspectors reviewed the licensee's corrective actions associated with this event to determine their adequacy to prevent
, recurrenca. Corrective actions included, but were not limited to improvements to instrumentation procedures controlling work on safety related ventilation systems including documentation o leads lifted to perform calibration t
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No violations or deviations were identified,
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11. -- Exit Interview The inspection scope.and findings were summarized on November 10, 1992, .
with those persons indicated-in paragraph:1. The i_nspector described :
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the areas -inspected and discussed in' detail the inspection findings .!
> listed below. No dissenting comments were received from the' license The licensee did not identify as proprietary any of the materials-provided to or reviewed by the inspectors during this inspectio *
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Item Number Description and Reference l
NCV 413/92-27-01 Failure to follow Procedures Regarding Control !
of Steam Generator Equipment flatch Opening (paragraph 4). !
URI 413, 414/92-27 02 Review Licensee Analysis of SMP Dampener Pressure Requirements (paragraph 8.b).
NCV 413/92-27-03 Inadequate Post-Weld Inspection (paragraph 8.c). ;
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