IR 05000483/1993017
| ML20059E261 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 12/29/1993 |
| From: | Tongue T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML19352B819 | List: |
| References | |
| 50-483-93-17, NUDOCS 9401110058 | |
| Download: ML20059E261 (27) | |
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V. S.-NUCLEAR REGULATORY COMMISSION REGION III :
Report No. 50-483/93017(DRP)
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Docket No. 50-483 License No. NPF-30
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Licensee: Union Electric Company ,
Post Office Box 149 - Mail Code 400 -
St. Louis, M0 63166 Facility Name: Callaway Plant, Unit 1
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Inspection at: Callaway Site, Steedman, M0 Inspection Conducted: September 19 through November 20, 1993
Inspectors: B. L. Bartlett D. R. Calhoun J. F. Schapker i T. M. Tongue .
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Accompanying Per onnel: S. S. Lee
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Approved By: T. M. Tongue, Acting ,ii , 1_gJj93 ~
Reactor Projects, Se .io 3A ate'
Inspection Summary Inspection from September 19 through November 20, 1993 (Report No. 50-483/93017(DRP)) .
Areas Inspected: Routine unannounced inspections of plant operations, 4 maintenance and surveillance, ESF actuation, followup on previously identified items, and in-service inspectio ;
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One cited violation was identified for a failure to report an ESF actuation signal in a t.imely manner (pa. Tgraph 4). Two non-cited violations were identified. One non-cited violation was for a failure to generate a 3 corrective action document (paragraph 3a) and the other non-cited violation '
was for a _ failure to get shift supervisor approval prior to starting work-on a !
modification packag (paragraph 3a)
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i The strengths noted included: gcod root cause analyses and effective corrective actions on several issues, improved housekeeping, and safe operation during the refueling outag ;
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The weaknesses noted included: The failure to promptly recognize an ESF '
actuation, failure to initiate a corrective action document, failure to obtain authorization to begin work, and two weak surveillance procedure i
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DETAILS 1. Management Interview (71707)
The inspectors met with licensee representatives - denoted in Paragraph 8 on December 3, 1993, to discuss the scope and findings of ;
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the inspection. In addition, the likely informational content of the inspection report with regard to documents or processes reviewed by the-inspectors during the inspection was also discussed. The licensee did ;
! not identify any such documents or processes as' proprietar Highlights of the exit interview are discussed below: j Strengths noted: i
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(1) Refueling activities were carried out in a methodical, safe, ,
and consistent manner (paragraph 2d). j (2) An additional . system for identifying individual fuel' *
assembly defects was installed and utilized during the ,'
refueling outage. This gave the licensee a redundant and diverse method for identifying leaking fuel assemblies l (paragraph 3b).
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(3) After questions were raised by the inspectors, good followup !
and independent assessment was performed concerning a weak l surveillance procedure (paragraph 2b). !
(4) After an inspector identified a loose hanger bolt, good root cause analysis and effective corrective actions were carried i out (paragraph 7a). .
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(5) Following a self revealing failure of a relief valve piping !
inlet, good root cause analysis and effective corrective .
actions were performed (paragraph-3a). l (6) Following an event where three maintenance mechanics ~!
inadvertently worked on a pump that had not been removed i from service, an event review team (ERT) was promptly [
formed. Discussions at the'ERT meeting were candid, j detailed, and thorough. Good root cause analysis and effective corrective actions were implemented I (paragraph 3a). j
.i (7) Plant housekeeping was good and better than previous - i refueling outages (paragraph 2c). j
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L Weaknesses noted: ;
b. (1) The operating crew failed to realize that an ESF actuation .i had occurred and performed inadequate followup (paragraph 4). .
(2) A corrective action document was not written for an equipment actuation until a concern was raised by the *
inspectors (paragraph 3a).
(3) There were several weaknesses observed during modification work: .
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e One modification was initiated without receiving permission to start by the shift supervisor (paragraph '
3a). ;
i e Workers were unclear if a cleanliness inspection was ;
required prior to installation of a valve. After ,
discussion with the inspectors a cleanliness inspection was performed and debris was identified in the pipe (paragraph 3a).
. Workers nearly used carbon steel wool on stainless steel pipe (paragraph 3a).
(4) Three maintenance mechanics were almost injured due to clearance order weaknesses (paragraph 3a).
(5) A weak engineering surveillance procedure resulted in the '
loss of shutdown cooling to the reactor for about 10 minutes ;
(paragraph 3b). -
(6) Following the inspectors' identification of a loose hanger bolt, the licensee identified that hanger reinstallation '
procedures were weak and that additional training for installation and Quality Control (QC) personnel was necessary (paragraph 3a). ;
(7) Corrective action was not taken on a weak surveillance 7 precedure and technical specification until concerns were '
raised by the inspectors (paragraph 2b). ;
(8) The inspectors identified loose debris in the containment .;
after the containment had been declared " clean" by the - '
licensee (paragraph 2c). ,
(9) The inspectors identified loose debris in and around the !
ultimate heat. sink (paragraph 2c). :
q c. One information followup item was closed (paragraph 5).
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i Four team members from Gosatomnadzor (the Russian equivalent to-the NRC) were introduced and they gave.the licensee thei impressions after spending a week at the plant. They said that l they were impressed with the licensee and appreciated the opportunity to observe an NRC inspector perform work activities !
and to learn about the NR . Plant Operations (71707)
The objectives of this inspection were to ensure that the facility was !
being operated safely and in conformance with license and regulatory requirements and that the licensee's management control systems were effectively discharging the licensee's responsibilities for continued safe operation. The methods used to perform this inspection included direct observation of activities and equipment, tours of-the facility, interviews and discussions with licensee personnel, independent verification of safety system status and limiting conditions for i operation (LCOs), corrective actions, and review of facility record j Areas reviewed during this inspection included, but were not limited to, ;
control room activities, routine surveillances, engineered safety feature operability, radiation protection controls, fire protection, '
security, plant cleanliness, instrumentation and alarms, deficiency reports, and corrective action General The plant entered refueling outage 6 during this inspection period. The outage lasted 52 days and at the end of the report i period the plant was being returned to full powe > Control Room Tours During the review of a surveillance test the inspectors became concerned that a reactor operator's (RO) concerns had been ,
insufficiently addressed. The R0 was concerned about the method used to verify the diesel generator met.the requirement for a i speed greater than or equal to 514 revolutions'per minute (rpm). 'j During the followup of the R0's concerns, the inspectors identified a weak surveillance procedure land a poorly worded ,
technical specification (TS). j l
On November 8, 1993, shortly after surveillance procedure, OSP-NE- ;l 00002, " Standby Diesel Generator Periodic Tests", was' completed, )
the inspectors interviewed a R0 about the-test results.- Procedure; OSP-NE-00002 implements TS surveillance 4.8.1.1.-2.a.4 which jl requires, in part, " Verifying' the d%sel starts and accelerates to
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i at least 514 rpm in less than or equai to-12 seconds. The I generator voltage and frequency shall be 4000 plus or minus 1 320 volts and 60 plus or minus 1.2 Hz withic 12 seconds ...." {
This TS was poorly worded. It recognized that frequency would !
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L vary slightly around 60 Hz but failed to recognize that speed will also vary around 514 rpm. Since the diesel and the generator are ;
l' mechanically secured any variance in speed will be due to a ;
l variance in frequenc Test personnel stationed locally at the. diesel generator-were unable to use local indication to verify that the diesel generator achieved a speed af 514 rpm within the time limits. When the R0 i was informed of this condition, he monitored his frequency meter' >
and recorded 59.8 Hertz (Hz). At that time, additional testing; equipment was obtained to locally verify diesel generator spee A reading of 512.9 rpm was obtained. Since the~12 second time '
requirement had elapsed, the RO felt that the speed acceptance criterion had not been met. He designated the surveillance as being completed unsatisfactorily. The R0 informed the control ,
room operating supervisor (05) of the test results. The OS stated :
that this issue had come up a number of times in the past and ,
there was an engineering evaluation that would address this concern. The evaluation was approximately 41/2 years old. A ,
copy of the engineering evaluation was later attached to the .
completed surveillance. The R0 who performed the test stated that '
although his onshift management was going to accept the surveillance, he felt the surveillance was unsatisfactory because not all acceptance criteria data had been obtained by methods required by the procedur i The inspectors informed the plant manager of the R0's concerns on j the above TS. The inspectors had compared the surveillance test l against the TS and were of the opinion that the TS requirement of _;
greater than or equal to 514 rpm was not being met. This :
technical non-compliance with a poorly worded TS did not result in ;
any safety hazar j The licensee promptly initiated a review of the surveillance procedure and test results. Independently, the plant manager 1 recognized the poorly worded TS and increased the scope of the review. Subsequently, it was identified that the 514 rpm speed criterion had not been met and the diesel was declared inoperabl The next day, the licensee conducted another diesel generator test-and verified that a speed of 514 rpm had been achieved. The '
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reason for the initial failure to meet the 514 rpm test ~ criterion was traced to an inadequate procedure which resulted in'an inconsistent method of measuring the time at which the speed ,
should be recorde j Conclusions The licensee reacted promptly once;this' issue was brought to their attention by the inspectors. The issue was thoroughly evaluated - 3 and effective corrective actions taken. Additional guidance has'
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been added to the surveillance procedure and a TS change will be submitted to the NRC. The licensee recognized that the delay in
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correcting the TS and in including guidance in the surveillance procedure was not appropriat c. Plant Tours
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The inspectors performed routine tours of the licensee's facilities. During the tours the inspectors observed that the
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areas were, in general, clean and properly maintained. The new
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outage maintenance facility allowed the welders, pipe fitters, electricians, carpenters, and other temporary outage personnel to make use of permanent, well laid out facilities. The use of temporary trailers and other temporary structures were minimized due to the use of the new outage maintenance facilit ,
On October 4,1993, and again several days later, the inspectors identified loose debris around the ultimate heat sink (UHS). ihe UHS is a small, open pond that the licensee utilizes to cool the reactor if the normal cooling systems are unavailable.
i While none of the identified debris would have prevented the UHS or essential service water (ESW) pumps from performing their intended functions, accumulated debris could eventually result in partial blockage of the ESW pumps intake area. Licensee management informed plant personnel of the need to pick up debris, l directed that the areas near the UHS be cleaned up, and periodically toured the UHS to ensure its continued cleanlines It appeared that the debris was blown into the UHS area by high
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winds.
l Containment During the refueling outage many items and material were brought into containment. Near the end of the outage the licensee removed this material and inspected containment to ensure all debris and equipment was removed. Following the licensee's completion of the inspection, the inspectors performed their own walkdown of containment. The inspectors identified and removed approximately twenty pounds of debris from the containment. Even though this debris should not have been present it would not have resulted in the blockage of the screens to the emergency sump The licensee was informed that while containment cleanliness was l much better than previous outages, additional improvement was .,
needed and their efforts to improve the cleanup following outages should continu d. Refuelinq *
l The inspectors observed the licensee's activities during refueling and reduced inventory conditions. The licensee was observed to 7 ;
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perform these activities safely and methodicall Fuel movement was deliberate and well planned. When problems were encountered fuel movement was stopped until safety issues were appropriately addresse No violations or deviations were identifie . Maintenance / Surveillance (62703) (61725)
Selected portions of the plant surveillance, test, and maintenance activities.on safety-related systems and components were observed or reviewed to ascertain that the activities were performed in accordance with approved procedures, regulatory guides, industry codes and i standards, and the Technical Specifications. The following items were considered during these inspections: the limiting conditions for operation were met while components or systems were removed from service; approvals were obtained prior to initiating the' work; activities were accomplished using approved procedures and were inspected as applicable; functional testing and/or calibration was performed prior to returning the components or systems to service; parts and materials that were used were properly certified; and appropriate i fire prevention, radiological, and housekeeping conditions were
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maintaine Maintenance I l
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Work Reauest N Activity W540279 Inspection of lugs on component cooling water valve EG HV-012 W514109 Inspection of lugs on component cooling water valve EG HV-012 W514106 Replacement of lugs on component cooling water valve EG HV-007 W541454 Adjustment of limit switches on valve BN HV-881 P497300 Cleaned heat exchanger tubes of the "B" centrifugal charging pump _ room coole P409167 Replacement of_ the rotating assembly in the "A" >
centrifugal charging pump.
l C528881 Weld in new replacement pipe and replace ESW
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containment isolation valve EF HV-003 L .
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C528884 Weld in new replacement pipe and replace ESW !
containment isolation valve EF HV-004 .
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W154928 Replace snubber with strut on support EP06R508242 ,
C532810 Replaced piping to class IE air conditioner, '
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C532815 Replaced piping to class IE air conditioner, I
SGK05 C525735 Installed new "B" feedwater flow ventur C525736 Installed new "C" feedwater flow ventur ,
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W161369 Replaced ESW piping to auxiliary feedwater pump Mechanics Work On Pump Without Clearance Order
On November 9, 1993, a maintenance foreman authorized three i mechanics to commence disassembly of the "A" auxiliary feedwater
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pump. The foreman believed he was signed on the necessary i workman's protection assurance (WPA or occasionally referred to as ;
a clearance order) but in fact was not. In addition,- the control room operators were unaware that this maintenance activity was-being performed.- Immediately after the mechanics had-disassembled ;
the coupling between the pump and the motor,' the R0' started the ;
pump for routine activities. Fortunately, no one was hurt and no l equipment was damage l The licensee immediately reverified all WPAs to ensure other near-misses would not occur and instituted an event. review team (ERT). !
Discussions at the ERT meeting were candid, detailed, and ?
thorough. A number of short and long term corrective actions were' ;
identified for this event and were being implemented on a timely !'
basi i The licensee has experienced a number of WPA problems over. the !
last six months. This event was the most serious of these events and effective corrective action following this event was-being take !
Improperly Installed Safety Related Hanaers l During the inspector's observations of a hydrostatic pressure test, a loose jam nut on a safety' related hanger.was identifie One hanger, EF06-0027/125, out of approximately eight hangers that .
had been removed and reinstalled for a ESW modification, was i identified by the inspectors as having a loose jam nut. The- i licensee was informed and they initiated an investigatio j 9 ;
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The licensee decided to perform a walkdown of all the hangers .
associated with the "A" train ESW modification as well as those :
hangers associated with the replacement of ESW piping to the "A" train class 1E air conditioner. The walkdown, performed by QC and
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engineering personnel, revealed a number of additional discrepancies. Work requests were generated to correct all the- ,
deficiencies identifie l L The following day the hanger engineer went to verify that all {
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rework had been properly completed. During this second inspection :
the engineer identified loose nuts on a hanger that had previously ,
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been verified to be tight. As a result of this second loose nut, management decided to convene a meeting to discuss the issues, ;
identify root causes, and determine additional corrective action '
Representatives from Quality Assurance, QC, Engineering, as_ well '
as the plant manager and the manager of work control, attended.the meetin The meeting consisted of a frank discussion of all :
potential possibilities of how the nuts became loose (inadequat !
QC verification inspections, inappropriate installation _
techniques, etc.). The actions decided upon at the conclusion of ,
the meeting were: QC management would provide clear expectations to the QC inspectors on their inspection responsibility during hanger reinstallation verification; hangers that had been signed-off as complete would be reinspected; additional spot-checks of other hangers in the general area would be performed during the performance of any upcoming hanger reinstallation work; and engineering was tasked with identifying all outage work which involved the removal and reinstallation of hangers, as well as providing additional training to the welders / pipe fitters on i proper hanger reinstallation technique Subsequent inspections of several hanger reinstallations, based on ,
work identified by engineering, identified some additional weak work performance practices. Work requests were generated t ,
repair these minor deficiencie l The inspectors reviewed the initial hanger list generated by i engineering and performed periodic inspections of hangers to i l ensure the licensee's corrective actions were effectively i implemented. The inspectors did not find any additional !
deficiencie ;
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The root causes of the hanger installation deficiencies were traced to insufficient training in the installation of the hangers !
and to inadequate QC verification of the hanger installatio !
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.f Inadvertent Ground During maintenance activities, two instrumentation and control (I&C) technicians inadvertently caused a ground which resulted in' l I
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normal supply to the emergency _ supply _(refueling water storage
, tank). The R0 immediately swapped the valve lineup back to normal l after verifying that the swapover was not required. A Suggestion,
Occurrence and Solution (a corrective action document referred to as an SOS) report was not generated by the I&C personnel involved nor by the onshift operations' crew as required by procedure.-
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After the inspectors expressed a concern about the lack of an SOS, one was written by the I&C department. However, the SOS failed to address the possible reportable occurrence and only addressed the '
inadvertent ground. The licensee subsequently concluded that the l event was not required to be reported to the NR J
Interviews with plant personnel revealed that most people quickly I recognized that this event was not reportable. This may have led l to the erroneous conclusion that an SOS was not required. This j i
appeared to be an isolated example of a failure to generate an l
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l_ A violation was not cited for the licensee's failure to generate l an SOS because the criteria specified in 10 CFR Part 2, Appendix i C,Section VII.B were satisfie j Fatique Crack at a Residual Heat Removal Relief Valve On October 12, 1993, the licensee ider,tified a leak from piping i near a residual heat removal (RHR) relief valve socket weld on the
"B" train. The "A" train outage had recently commenced when the leak was noted. Actions were immediately taken to restore the "A" ,
train to an operable status while repair efforts were initiated to i repair the leak on the "B" train. Non-destructive testing l revealed a 3/4 inch circumferential flaw near the socket wel .;
The crack began at the toe of the weld but then proceeded through l the base metal. Timely repairs were made.with minimal impact on ;
the outage schedul j q
The licensee instituted an event review team and had'a contractor l come to the site to help. perform a root cause analysis. The j preliminary evaluation indicated the failure was due to fatigue cycling. The discharge piping of the relief valve had a j relatively long moment arm resulting in the fatigue cycling. The i licensee added additional support to the long discharge piping.
l The licensee also performed 12 additional penetration tests (pts) -!
on various systems at locations which had similar configurations'. 'l'
All results were satisfactor Observations of Modifications The inspectors observed an extensive amount of work on several modifications, which included: ESW to auxiliary feedwater (AFW)-
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pumps' piping replacement with carbon steel; replacement of vent valves (carbon steel) on ESW to AFW pumps piping with stainless steel valves; replacement of ESW piping (carbon steel) to and from the Class IE air conditioners (A/C) with stainless steel piping; :
replacement of vent valves (carbon steel) on ESW.to Class.1E A/C with stainless steel valves; replacement of ESW inlet and outlet valves (carbon steel) to the containment coolers with stainless steel valves; and support modifications on the ESW piping to !
containment cooler The inspectors reviewed work packages in the field to verify approval had been given to start work; verified workers were on the appropriate equipment; ascertained that hot work permits were signed and posted in the general area or located in the work package; determined correct parts were issued as indicated by the work instruction requirements; sampled the qualifications o radiographers and welders to ensure workers were qualified to perform tasks; evaluated a selected sample of weld radiographs; observed hydrostatic test of ESW replacement piping to verify proper installation and acceptance of work; and reviewed procedures to verify acceptability and technical conten In reviewing the above modifications, a number of minor problems were note The ESW to AFW pumps piping replacement was performed under a work request and had been authorized to start. The ,
stainless steel vent valve replacement work, although simultaneously performed, was worked under a Callaway Modification Package (CMP) and had not been authorized to start. After the inspectors discussed this issue with the responsible foreman, the control room shift supervisor was notified of the in-progress work, and the package was ;
authorized to star ;
The licensee's failure to obtain shift supervisor's approval prior to starting work on the CMP is a violation of TS *
6.8.1.a. However, this violation was not cited because the criteria specified in 10 CFR Part 2, Appendix C, Section- i VII.B.1 were satisfied.
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During the ESW valve replacement activities, valve EF HV- ;
0045 had been set-up in preparation for welding without a .j cleanliness inspection having been conducted by QC. The
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valve set-up had been performed by the day crew and th night shift crew was scheduled to perform the weld. In- l discussions with the night shift foreman, it was not -l apparent if the work package required a' cleanliness ;
inspection. The inspectors discussed these issues with the construction projects coordinator, the contractor foremen responsible for the work and Quality Control personnel There appeared to have been some confusion and i
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miscommunication among the various departments involved with respect to the ASME Section XI plan requirement Subsequently, plan requirements were clearly communicated to i
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the foremen and a cleanliness inspection was performed as required. A substantial' amount of material was removed from the interior of the pipe during this inspectio '
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A dye penetration test (PT). had to be performed on a weld a number of. times before the test was successful. After each PT failed, additional grinding was performed to smooth out- ,
scratches which had prevented successful completion of the tes While a PT was underway, the work group (welders and ,
pipe fitters) concluded that the grinding tool was not small enough to adequately remove the scratches. The workers-decided to use steel wool, to remove scratches, if the present PT failed. This decision presented the. potential to use carbon steel wool on stainless steel piping. A QC inspector was. informed of this planned action. .The ,
inspectors and the QC inspector discussed this issue with '
the welders and pipe fitters. The workers stated that they were aware of this issue and that they had intended to use stainless steel wool. However, at the time, one of the pipe fitters returned to the jobsite with carbon steel wool. The foreman was informed of this issue and verified that the workers fully understood material use requirements, j Surveillance 'i The reviewed surveillances included:
Procedure N Activity EDP-ZZ-01005 Post maintenance hydrostatic test program for ASME Section XI, Class 1, 2, and OSP-BB-VLOO6- Section XI reactor coolant system (RCS) pressure - '
isolation valves leak rate tes .
MSM-KJ-QK001 Standby "A" diesel generator disassembly and a reassembl MSM-KJ-QK001 Standby "B" diesel generator-disassembly and reassembl ETP-ZZ-ST002 Initial criticalit '
ETP-ZZ-ST004 Boron endpoint measuremen ETP-ZZ-ST003 Determination of low power physics testing :
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ETP-ZZ-00012 Inverse count rate ratio monitoring approach to criticalit MSE-NK-QB004 Eighteen / snth surveillance on large stationary batterie MSE-NK-QB005 125 Volt battery performance discharge tes MSE-ZZ-QS002 480 Volt circuit breaker preventative maintenance and inspection OSP-EJ-V0002 Section XI residual heat removal and reactar coolant system check valve operabilit TG-ZZ-0002 Reactor startu MPM-ZZ-QA301 Limitorque operator inspection and lubrication on valve operators BN HV-8813, EG HV-0071, EG HV-0126, and EG HV-012 '
MPM-BB-QR009 Reactor vessel head remova MPM-BB-QR010 Reactor vessel upper internals remova MPM-BB-QR014 Reactor vessel head replacemen '
MPM-BB-QR014 Reactor vessel upper internals t ' installatio ,
Loss of All Residual Heat Removal (RHR) Flow During an engineering surveillance test of the reactor coolant system (RCS) check valves, the RO realized that by following th procedure he had caused all RHR flow and cooling to be briefly terminated to the reactor core. The test was temporarily >
suspended and cooling was restore '
On October 9, 1993, during Mode 6 (refueling) with the reactor vessel (RV) head off and greater than 23 feet of water above the .
RV flange, the licensee was conducting surveillance tests. During a surveillance test to ensure that the RCS check valves would properly open, certain valve manipulations were performed. The-valve manipulations were performed to maximize flow through the check valves and then to briefly stop that flow. In order to maximize flow, the bypass line around the;only. operating RHR heat'
exchanger was opened. This reduced differential- pressure and increased the flow rate. It also resulted in the RHR flow that -
was going to the reactor core from receiving any significant- 1 cooling. This caused core exit temperatures to start increasin Several minutes later the remote manual discharge valves were '
briefly closed to stop flow through the check valves. The remote :
valves were reopened and cooling reestablished to the RHR heat l
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exchange Immediately following the completion of these activities, the RO realized the consequences of the surveillance test, suspended performance of the test _and wrote an S0S. Core exit temperatures were not available, but the temperatures at the discharge of the RHR pump increased about 10 degrees (89 to 99 F) during the 10 minute even The ins.g. '. ors interviewed personnel concerning this event and reviewed the surveillance tests. As part of the licensee's corrective action the surveillance test was modified prior to ;
continuing the testing activitie The inspectors-reviewed the :
revised survaillance test and were concerned with the new :
procedure. The old procedure had required that RHR flow and .
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cooling be briefly suspended and the revised procedure corrected this error; however, the new procedure did not specifically '
pmvent the loss of RHR cooling or flow and under certain valve operations, the loss of RHR flow could have been repeated. The-licensee agreed to revise the new RCS check valv.' testing procedure prior to the next performance of the surveillance test to ensure that a loss of RHR flow or cooling would not occu ;
Technical specification (TS) allows the temporary suspension of RHR flow and cooling to core for specified time limits. However, this should only be done with the knowledge and involvement of the operators. The engineering test procedure did not represent a safety significant issue or a TS violation, but it did represent a *
weak and poorly written surveillance tes ,
Results Of Fuel Assembly Surveillance Activities Prior to refueling outage 6, the licensee'had identified that dose ,
equivalent iodine was trending up which indicated fuel boundary degradation. During the full core offload, the licensee performed sipping and ultrasonic testing (UT) of all fuel assemblies. To ensure better understanding of the fuel assemblies the licensee installed an in-mast sipping system. The sipping system enabled t the licensee to check each fuel assembly for radioactive gases .;
that would indicate fuel degradation. Sipping can only identify which fuel assemblies had fuel leakers. Only UT could identify f which fuel pins within the assembly were. leakin Sipping identified two leaking assemblies and UT identified an individual pin in another assembly as leaking. Ultrasonic testing did not identify a leaking pin in one of the assemblies that sipping identified as degraded. A total of two pins were replaced '
using bottom nozzle reconstitution. One rod had a thumbnail size -:
hole and the other rod broke'in two pieces while being remove Neither assembly will be utilized in future fuel cycles. At the present time, the root causes of the fuel failures have not been definitively determined. The final root cause analyses has not
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been completed. The assembly which-UT could not identify a failed rod was scheduled to be reinserted in the Cycle 7 reload. Plant management later decided not to reload the assembly and new j replacement assemblies were procured from the fuel vendo .
No deviations were identified, however, two non-cited violations were identifie ,
4. Engineered Safety Features (ESF) Actuation (93702)
On November 14, 1993, while conducting a surveillance, a feedwater '
isolation signal (FWIS) occurred. The control . room (CR) staff failed to ,
recognize that the FWIS had been received. Troubleshooting on the following morning revealed the FWIS and the event was reported to the ,
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Operations surveillance procedure, OSP-AL-V0002, "AFW to Steam Generators Flow Path Verification," was in progress to verify the auxiliary feedwater flow control valve limits the flow to each steam generator from the motor driven pump . Also, instrumentation and !
control (I & C) activities were being performed concurrently to support ,
the operations surveillance procedure. I & C technicians were adjusting the valve settings as the Reactor Operator (RO) provided required flow .
through tha valves in accordance with the procedure. This evolution resulted in feeding the "A" and "D" steam generators which were at level indications of approximately 65% and 40%, respectivel .!
The operations procedure did not specify an allowed level band for the SGs nor was the R0 instructed by onshift management to maintain SG level within a specified limit. Without any guidance, the R0 decided to limit SG level to 70% .in order to ensure that the level did not reach the hi-hi SG alarm of 78%. The R0 was using a recorder to monitor for SG level. The recorder was selected to transmitter channel AE LT-0551, which was subsequently determined to be reading lower than all.other 1 SG level indication During valve setting activities, the 70% limit in the "A" SG was reached, but the I&C technicians had not completed their valve setting The I&C technicians requested additional time from the RO stating that- .
the valve setting was near completion. The R0 decided to set a new SG limit of 75% in order to assist the technicians. Upon reaching the_self_- >
imposed 75% limit, the R0 began closing the_ supply valve to. isolate-fl ow. At that time, the hi-hi SG level / turbine trip annunciator came-i Concurrent with the annunciation, the solid state protection system l (SSPS) sent signals to both trains to actuate a FWIS. The R0 informed '
onshift management of the alarmed annunciator. Both the shift j supervisor (SS) and the control room operating supervisor (0S) '
investigated the issue. Based on conflicting and limited information, onshift management concluded that a FWIS did not occur. Difficulties encountered included: unavailability of the partial trip panel (SB069),- j
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other feedwater isolation valves on the engineered safety features actuation (ESFA) panel showing that they were locked in, the failure of the chemical injection valves (which also receive a FWIS) to show as l actuated on the ESF staNs panel (this was later traced to a failed
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relay), "B" train main steam and feedwater isolation system (MSFIS) '
cabinet showed that it had received an actuation signal while "A" train '
MSFIS cabinet did not, and the previously mentioned differences in all four "A" SG level indicator '
Other significant contributing factors to the shift supervisor's decision was the misunderstanding that the annunciation of the alarm window and actuation of SSPS occurred at different SG levels. In addition, the SS felt that since feedwater isolation was not required in Mode 4, that if an ESF actuation had occurred it would not be *
reportabl After the "A" SG level was reduced below the alarm limit, the FWIS was still present on the "B" train but not on the "A" train as previously mentioned. Troubleshooting the following morning confirmed that a.FWIS had indeed occurred. The discrepancy between the two trains was due to a faulty cell on the "A" reactor trip breaker which allowed. the FWIS to clear on the "A" train. A bad relay was the root cause of the chemical injection valves not illuminatin Onshift management did not meet plant management's expectations in .the investigation / resolution of this incident. As a result of misdiagnosis of the event, onshift management failed to recognize the occurrence.of a ,
FWIS at 6:13 p.m. CST on November 14, 1993. The licensee was required >
to report this event to the NRC by 10:13 p.m. CST to meet the four-hour reportability requirements. The NRC was not notified of the event until 2:30 p.m. CST on November 15, 199 ,
The licensee's failure to report the automatic actuation of the ESF within the four hour reporting limit of 10 CFR 50.72 is a violation, <
483/93017-01(DRP).
No deviations were identified, however, one violation was identifie ]
5. Followup on previously Identified Items (93702)
(Closed) Inspection Followup-Item No. 483/93006-01(DRS): The licensee was requested to respond to a concern regarding reactor plant control -)
manipulations not being performed by all members of a crew. The- ;
licensee appeared to be committed to individual performance of the plant 1 control manipulations by their training procedure and in practice gave '
credit to all crew members for plant manipulations performed by the j cre o The licensee's reply stated that their requalification training program )
was based on a systematic approach to training-(SAT). 10 CFR 55.59 (c)
allows a SAT program to define how the required plant manipulations are to be accomplished. The licensee stated that their intent was to give i
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e credit to all crew members for plant manipulations since crew l performance is evaluated .nd critiqued at each training session's conclusion. The licenser stated that each individual is held accountable for the crew s performance on all plant manipulation The j licensee revised the training department procedure (TDP-ZZ-00022) to ;
reflect this position far giving credit to each individual for the ;
crew's. control manipulhtions. -This item is close l
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No violations or deviations were identifie . In-Service Inspection , Program Review (73051)
Personnel from Conam Nuclear, Inc., and Nuclear Energy Services ,
(NES) performed the in-service inspection (ISI) in accordance with _i'
the licensee's program and ASME Section XI,-1980 Edition, Winter 1981 Addenda. The licensee did not make a request for relief from the ASME code for this outage. Organizational staffing for the ,
ISI program was found to be acceptable and the services of an Authorized Nuclear Inservice Inspector (ANII) were procured from the Hartford Steam Boiler Inspection and Insurance Compan Procedure Review (73052) ,
All applicable ISI procedures were approved by the ANII and were'
reviewed by the NRC inspectors. The ISI-procedures were found to '
be acceptable and in accordance with ASME Section V,1980 Edition, Winter 1981 Addenda, Data Review (73755) 1
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General The examination data was found to be in accordance with the applicable ISI procedures and ASME Code requirements. The NRC i inspectors reviewed documents related to nondestructive examinations, equipment, data, and evaluation Eddy Current Examination (ET)
During this outage,100 percent of all accessible tubes in-steam !
generators "A" and "D" were examined full length. The examinations were conducted utilizing the Aetec MIX-18 l multifrequency digital test equipment with associated acquisition !
software and remote positioning devices. -The Zetec Digital-Analysis System was used for the data evaluatio {
Motorized rotating pancake coil (MRPC) examinations were utilized ;
to supplement the bobbin coil examinations. The MRPC-was used to ;
further characterize manufacturing burnish marks, undefined i indications, distorted indications, and a sample of 600 top of the ;
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tube sheet' examinations subsequent to shot peening performed at
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The following tubes were plugged as a result of this examination: !
Steam Tubes Tubes. Plugged l Generator Pluaaed Previously A 19 17
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D 18 .30 ;
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All of the new steam generator tube plugs were-Inconel 690 allo No tubes were sleeved this outag ;
Observation of-Work Activities (73753) j
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The NRC inspectors observed work activities and;had discussions- q with personnel during the ISI activitie These observations- j included the following: 1
1) NES personnel performing ultrasonic and liquid penetrant 1 examinations on pipe welds No 2AE-050C525-737-FW01; 2AE-05- i
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C525-737-FWO2; 2AE-04-C525-735-FW01 and 2AE-04-C525-735-FW0 These welds were in the main feedwater syste ) MQS personnel performance of radiography examinations on pipe welds No. A525-734-FWO2; FWO2Rl; A525-735-FWO2; FWO2Rl; A525-734-FW01; and FW0lRI. These welds were in the main !
feedwater syste ;
The NRC inspector reviewed the qualifications and. certifications = .l of all inspection personnel performing ISI to ensure conformance (
with SNT-TC-1 ,
No violations or deviations were identifie l Systematic Assessment of Licensee Performance (SALP) g
On November 2,1993, a public meeting was held with:the licensee. .The 1 meeting was. chaired by Mr. Hubert Miller, Deputy Regional . Administrator, i Region III and was held in.the licensee's emergency. operations facility 1 (EOF). The latest SALP report (50-483/93001)-was discussed and the-licensee. presented their views and responses to the' report. A copy _of the overhead slides that were used during the meeting are included as an- .
attachment to this repor !
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' Violations -for Which a " Notice of Violation" Will- Not be issued ~
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The NRC uses the Notice of Violation as a standard method for formalizing the existence of a violation of a legally binding- ='
requirement. However, because the NRC wants to encourage and support ;
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the NRC will not generally issue a Notice of Violation for a violation i u that meets the tests of 10 CFR 2, Appendix C, Section V.A. These tests-l are: (1) the violation was identified by the licensee; (2)-the violation would be categorized as Severity Level IV or V; (3) the i violation was reported to the NRC, if required; (4) the violation will !
be corrected, including measures to prevent recurrence, within a-l- reasonable time period; and (5) it was not a violation that could- !
reasonably be expected to have been prevented by the licensee's :
corrective action for a previous violatio j i
Two violations for which no Notice of Violation will be issued areL i discussed in paragraph . Persons Contacted D. F. Schnell, Senior Vice President, Nuclear
- L. Randolph, Vice President, Nuclear Operations-
- J. D. Blosser, Manager, Callaway Plant C. D. Naslund, Manager, Nuclear Engineering J. V. Laux, Manager, Quality Assurance J. R. Peevy, Manager, Operations Support M. E. Taylor, Assistant Manager, Work Control
- D. E. Young, Superintendent, Operations M. S. Evans, Superintendent, Health Physics
- S. S. Sampson, Supervising Engineer, Site Licensing G. J. Czeschin,- Superintendent, Planning and Scheduling G. R. Pendegraff, Superintendent, Security
.C. E. Slizewski, Supervisor, . Quality Assurance Program
- G. A. Hughes, Supervisor, Independent Safety Engineer Group -
C. S. Petzel, Quality Assurance Engineer l J. A. McGraw, Superintendent, System Engineering '
R. D. Affolter, Superintendent, Design Control
- H. Potter, ANII
- .Kuznetsova, Gosatomnadzor (Russian NRC)
- Adamchik, Gosatomnadzor
- Sapozhnikov. Gosatomnadzor
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- Semyakin, Gosatomnadzor
- Denotes those present at one or more exit interview I In addition, a. number of equipment operators, reactor operators, senior-reactor operators, and other members of the quality control, operations, maintenance, health physics, and engineering staffs were contacte l
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OPERATIONS IMPROVEMENT OPPORTUNITIES :
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+ Rotation of Licensed Personnel -
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Reactor Operators - Procedures
+ Non-Licensed Training Additional Requal Weeks j JPMs -
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MAINTENANCE IMPROVEMENT OPPORTUNITIES
+ Procedure Improvements
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+ Implement Maintenance Rule !
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+ Continued improvement in communications with Operations Continue to refine our system walkdown program q with Operations Upgrade our recurring issues program to be more timely and available
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+ Improved teamwork with Maintenance Include maintenance input when procuring or upgrading diagnostic equipment
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Continue enhancements to our MOVATS program implementation
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+ Streamlining the design change process to: !
maximize efficiency while maintaining a.high a level of quality
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Improve input from other departments throughout- i the design process ;
i Improve departmental management oversight.at 1
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+ Technical and professional development of the i Engineering staff j q
Continued involvement with nuclear. industry issues and programs Continue to develop close working relationship with ;
industry experts in the specialty fields; for example, I
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Automated Access Control- Electronic Dosimetry !
Radiation Survey Tracking / Trending .! Radiation Monitor System Tracking / Trending j i
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+ Resolve extent of future use of Radiological l Release Information System for dose assessment calculations >
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Drills and Exercises
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to enhance troubleoome EP areas j
+ Review independent assessment methods used l for EP )
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