IR 05000346/1997002: Difference between revisions

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{{Adams
{{Adams
| number = ML20135F106
| number = ML20140D147
| issue date = 03/06/1997
| issue date = 04/17/1997
| title = Insp Rept 50-346/97-02 on 970113-17.Violations Noted.Major Areas Inspected:Operations,Maint,Qa & Engineering
| title = Ack Receipt of 970407 Ltr Informing NRC of Steps Taken to Correct Violations Noted in Insp Rept 50-346/97-02 on 970306
| author name =  
| author name = Grant G
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
| addressee name =  
| addressee name = Wood J
| addressee affiliation =  
| addressee affiliation = CENTERIOR ENERGY
| docket = 05000346
| docket = 05000346
| license number =  
| license number =  
| contact person =  
| contact person =  
| document report number = 50-346-97-02, 50-346-97-2, NUDOCS 9703110193
| document report number = EA-97-097, EA-97-97, NUDOCS 9704230014
| package number = ML20135F103
| package number = ML20140D153
| document type = INSPECTION REPORT, NRC-GENERATED, TEXT-INSPECTION & AUDIT & I&E CIRCULARS
| document type = CORRESPONDENCE-LETTERS, OUTGOING CORRESPONDENCE
| page count = 28
| page count = 2
}}
}}


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U.S. NUCLEAR REGULATORY COMMISSION
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REGION 111
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Docket No.: 50 346
:  License No.: NPF-3 Report No.: 50-346/97002(DRS)
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  . Licensee: Toledo Edison Company
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$  Facility: Davis-Besse Nuclear Power Station Location: 5503 N. State Route 2
:  Oak Harbor, OH 43449 Dates: January 13-17,1997
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h l  Inspectors: Andrew Dunlop, Reactor Engineer
 
Steven A. Eide, INEL
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Martin J. Farber, Rill (Tearn Leader)
Ronald K. Frahm, Jr., NRR (Staff Support)
,  Donald E. Jones, Reactor Engineer John H. Neisler, Reactor Engineer
;      .,,
Approved by: Wayne J. Kropp, Chief
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Engineering Specialists Branch 1 Division of Reactor Safety
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i 9703110193 DR 970306 ADOCK 05000346 PDR J
 
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;,    EXECUTIVE SUMMARY
!-
,  Davis-Besse Nuclear Power Station, Unit 1
.
NRC Inspection Report 50-346/97002(DRS)


This inspection included a review of the licensee's implementation of 10 CFR 50.65,
April 17, 1997    l
,
          ;
" Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants." '
FA 97-097        ,
j- The report covers a one week on-site inspection by regional and Office of Nuclear Reactor Regulation (NRR) inspector '
Mr. John K. Wood        ;
Ooerations      ,
          '
i
Vice President - Nuclear Davis-Besse Nuclear Power Station Centerior Service Company        ;
*
          '
l  Operators' knowledge was consistent with their responsibility for implementation of ( the maintenance rule (MR). There was~no indication that the maintenance rule
5501 North State - Route 2 Oak Harbor, OH 43449 SUBJECT: NOTICE OF VIOLATION (NRC INSPECTION REPORT 50-346/97002(DRS))    .
:
I Doar Mr. Wood:       l
detracted from the operators' ability to safely operate the plant. Using the Risk
          )
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This will acknowledge receipt of your April 7,1997 letter in response to ou:
Significant System Matrix chart helped operators monitor and limit risk.
March 6,1997 letter transmitting a Notice of Violation associated with the above    !
mentioned inspection report. This report summarized the results of the maintenance rule    i l
inspection at your Davis-Besse Plant. We have reviewed your corrective actions and have i
no further questions at this time. These corrective actions will be examined during future  ,
          !
inspections,        i
          !


i'
Sincerely,
l: Maintenance
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The team concluded that the Davis-Besse structures, systems, and components
!
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  (SSC) were properly scoped although documentation to support scoping decisions i  was weak.
    /s/ M. Icach (for)
 
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The expert panel was composed of well-qualified, experienced personnel. The i
Davis-Besse Individual Plant Examination (IPE) was used in conjunction with the panel's experience base to accurately assess the risk significance of the SSCs. No j-
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formal administrative procedures or guidance had been developed to govern expert panel activities.
 
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Based on the reviews conducted, the team determined that the licensee's approach to establishing the risk ranking for SSCs within the scope of the Maintenance Rule
,  (MR) was adequate. However, weaknesses in that approach included the use of an
{  outdated IPE and inadequate documentation of the expert panel's determinations.
 
!
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* *
e firey E. Grant, Director    !
The procedures for performing petiodic evaluations met the requirements of the rule
  . M 0 0 5 '?  Division of Reactor Safety Docket No. 50-346 Enclosure: Ltr 04/07/97, J. K. Wood,      .
, and the intent of the Nuclear Management Resource Council (NUMARC)
          '
l implementing guidance. The team noted that the reports addressed all of the
Centerior Energy, to US NRC w/enci See Attached Distribution
;  aspects specified in NUMARC 93-01 " Industry Guideline for Monitoring the
        )\
;
Effectiveness of Maintenance at Nuclear Power Plants," and 10 CFR 50.65(a)(3).


l  The team considered the preparation of system-specific periodic reports as a i significant strengt *
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The licensee's process for conducting the balance between reliability and availability
DOCUMENT NAME: G:\DRS\DAV041_7.DRS To receive e copy of thie document, indicate in the box 'C' = Copy w/o attach /enci "E" = Copy w/ attach /enci "N* = No copy l OFFICE Rill:DRS , , , g Rlll:DRP6\ % R ll:EICS lg Rlli:DRS lv NAME Gavula/kjKLP(  Jacobsqh\,h\,N Yayton C9g/ff8 Leach / Grant &4
: was considered acceptable. Changes to the preventive maintenance program were
  "
[ DATE 04/ /7 /97( '  04/p/91 )\ 04/f 7/97 /g 04//7/97
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OFFICIAL RECORD COPY  /
I 9704230014 970417 r j- gDR ADOCK 05000346 PDR .
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indicative of an ongoing effort to optimize reliability and availability.
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;  The team viewed the licensee's process for assessing plant risk resulting from multiple equipment outages to be appropriato. However, the tool used to assess
  '
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c'  plant risk while at power, the risk matrix, was viewed as weak in terms of making use of IPE risk information and providing user guidance.
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' The establishment of performance criteria and goal setting was considered goo !
The use of specific functional failure (FF) performance criteria for non-risk
- significant SSCs, along with maintaining plant level performance criteria, was a program strength. Performance criteria for reliability for safety significant SSCs, however, were deficient in that the FF criteria were not based on demands or run time. Based on the recent efficient use of outage times for several SSCs, the effectiveness of the unavailability performance criteria as an indicator may have
. been reduce ,
        !
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* The licensee adequately scoped buildings and enclosures as structures under the l rule. However, the structure monitoring program concentrated only on building j supports and did not consider system and internal component supports. Further, no i guidance was established for moving structures between the (a)(2) and (a)(1)
categorie *-
The team concluded that the licensee had properly integrated the MR into the existing industry operating experience (IOE) program. Adequate provisions had
' been made to incorporate information from the IOE program into periodic evaluations, goal development, and functional failure evaluation ,
*
The material condition of the plant systems examined was very good. With a few
. minor exceptions, the systems appeared to be well-managed and were free of corrosion, oil, water, steam leaks, and extraneous materia (
Quality Assurance (QA)      1
      :
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The team concluded that the licensee's self-assessment activities were appropriately conducted and identified worthwhile issues. Use of independent personnel was considered a strength of the 1995 self-assessment. The team further concluded that Audit 96-MAINT-01 was also a worthwhile effort although it did not identify a significant deficiency (lack of criteria for reclassifying a structure from (a)(2) to (a)(1)) in the structure monitoring program).
 
Enaineerina
*
System engineers (SE) had been trained and appeared qualified to provide oversight of the implementation of the rule for their respective SSCs. The team also noted that system engineers were actively involved in MR implementation and generally viewed the MR as a useful too i
.      l Report Details Summarv of Plant Status The plant was operating at full power during the inspectio Introduction This inspection included a review of the' licensee's implementation of 10 CFR 50.65,
" Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants."
 
The report covers a one week on-site inspection by regional and NRR inspector I
      !
1. Operations 04 Operator Knowledge and Performance 04.1 ' Ooerator KnowledoeJf Maintenance Rule
.
i Insoection Scone (62706)
i
*
*
During the inspection of the implementation of 10 CFR 50.65, " Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," the team interviewed two licensed reactor operators and three senior reactor operators to determine if they understood the general requirements of the rule and their
; particular duties and responsibilities for its implementation,  i i Observations and Findinos
.
The team found that the operators had a general working knowledge of the
maintenance rule and their role in its implementation. They stated their primary i duties included review of maintenance activities and comparison of these activities j with the risk significant system matrix (RSSM) chart. They stated that they used the RSSM matrix chart to aid in identifying systems that were within the scope of
$
the station's maintenance rule program. In addition, they were tasked with the i
timely removal and restoration of equipment and accounting of equipment out-of-l service tim Tasks associated with the maintenanco rule that operators were responsible for included:
o Returning all SSCs to service as soon as possible in order to minimize unavailabilities
* Documenting SSC outages in the control room log for all SSCs under the scope of the maintenance rule
* Noting when equipment was taken out-of-service and when equipment was returned to service
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The operators indicated that the maintenance rule was integrated with their day-to-day activities, and that it did not impose additional administrative burdens that distracted them from their responsibility to safely operate the plan . Conclusions Operators' knowledge was consistent with their responsibility for implementation of the maintenance rule. There was no indication that the maintenance rule detracted from the operators' ability to safely operate the plant. Using the RSSM chart helped operators monitor and limit ris II. Maintenance M1 Conduct of Maintenance (62706)
The primary focus of the inspection was to verify that the licensee had implemented a maintenance monitoring program which satisfied the requirernents of 10 CFR 50.65, " Requirements for Monitoring the Effectiveness of the Maintenance l at Nuclear Power Plants," (the maintenance rule). The inspection was performed by a team of five regional and headquarters inspectors and a consultant from the Idaho l National Engineering Laboratory. Assistance and support were provided by one member of the Quality Assurance and Maintenance Branch, NR M1.1 SSCs included Within the Scooe of the Rule
      [
8 Insoection Scoce The team reviewed the licensee's scoping documentation to determine if the appropriate SSCs were included within their maintenance rule program in accordance with 10 CFR 50.65(b). The team used Inspection Procedure 62706, Maintenance Rule, NUMARC 93-01. and Regulatory Guide 1.160 as references during the inspectio Observations and Findinas The licensee's maintenance rule was described in procedure MRPM-02,
" Maintenance Rule Program Manual," Revision 2 (December 15,1996). This program described the methodology used to select the SSCs under the MR. The methodology considered whether SSCs were safety related, whether failures could cause accidents or transients, viere used in Emergency Operating Procedures (EOP),
whether SSC failure resulted in safety-related system failure, or failures could cause a safety-related system actuation and non-safety-related SSCs used to mitigate accidents or transients. Based on results of these evaluations, lists were developed of systems within the scope of the MR and of systems excluded from the scope of the M In general, the scoping of SSCs was conservative. The licensee considered about 290 sub-systems and components in the initial scoping phase. Through a process of consolidation, this was reduced to 89 SSCs: of these, about 60 SSCs were placed within the scope of the MR. The team noted that the licensee had little
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J. Wood  2  April 17,- 1997 cc w/o encl: J. P. Stetz, Senior Vice President - Nuclear J. H. Lash, Plant Manager J. L. Freels, Manager, Regulatory Affairs
documentation to support their decisions as to whether a system should be placed l within the scope of the MR. Documentation available for review consisted of SSC scoping sheets which included the four basic questions in NUMARC 93-01 with little documentation to support the answers to the question Conclusions    l l
      {
The team concluded that the Davis-Besse SSCs were properly scoped although documentation to support scoping decisions was wea M1.2 Safety (Risk) Determination. Risk Rankina. and Exoert Panel l l
cc w/ encl: State Liaison Officer, State of Ohio Robert E. Owen, Ohio
      , Insoection Scone i
   -
Paragraph (a)(1) of the rule requires that goals be commensurate with safet Additionally, implementation of the rule using the guidance contained in NUMARC 93-01, required that safety be taken into account when setting performance criteria and monitoring under paragraph (a)(2) of the rule. This safety consideration was to be used to determine if the SSC should be monitored at the system, train or plant level. The team reviewed the methods and calculations that the licensee established for making these risk determinations. The team also reviewed the risk determinations that were made for the specific SSCs reviewed during this inspection. NUMARC 93-01 recommended the use of an expert panel to establish safety significance of SSCs by combining Probabilistic Risk Assessments (
Department of Health     !
(PRA) insights with operations and maintenance experience, and to compensate for d the limitations of PRA modeling and importance measures. The team reviewed the composition of the expert panel and experience and qualifications of its member The team reviewed the licensee's expert panel process and the informa+ ion available which documented the decisions made by the expert panel. The team interviewed several members of the expert panel to determine their knowledge of the MR and to understand the functioning of the pane bservations and Findinas on the Expert Panel The licensee used an expert panel referred to as the " Maintenance Rule Working Group," in conjunction with a PRA ranking methodology to determine the safety significance of SSCs within the scope of the MR. The panelincluded a PRA analyst and representatives from operations, maintenance, engineering, and plannin Average nodear experience of the panel members was more than 12 years. In addition, the MRPM permitted subject matter experts to supplement the panel, as neede The expert panel held quarterly meetings to review the quarterly equipment windows and the periodic maintenance effectiveness assessment report to ensure performance monitoring and goal setting activities were proceeding as desired. The team reviewed minutes of the panel's quarterly meetings. The minutes appeared to provide an accurate description of the panel's general activities as the activities were described to the team by the panel member _ _ . _ _ .
C. A. Glazer, State of Ohio Public Utilities Commission Distribution:      ;
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Rlll PRR w/enci Rlli Enf. Coordinator w/ encl Docket File w/, encl _7 SRI, Davis-Besse w/ encl g_PUBLIC IE-01;w/ encl   TSS w/enci  -
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OC/LFDCB w/encI  LPM, NRR w/enci J. Lieberman, OE w/enci '
 
DRP w/ encl  A. B. Beach, Rlll w/enct J. Goldberg, OGC w/ encl DRS w/enci  C. D. Pederson, Rif t w/ encl R. Zimmerman, NRR w/enci
The team noted that no formal charter or procedure had been developed to govern the activities of the expert panel, nor was there any criteria for selection of panel members as to experience, training, or level of expertise. There was no quorum established for other than decisions to alter the implementation of the rule, nor l
were alternates identified for the five panel members,   l
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interviews with panel members indicated that they were knowledgeable in the requirements of the MR, system scoping,'use of the PRA in risk assessment, and
;  . were able to accurately assess risk significance of the SSC Conclusions on Expert Panel     l
 
1 The expert panel was composed of well-qualified, experienced personnel. PRA was I used in conjunction with their experience base to accurately assess the risk
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significance of the SSCs. No formal administrative procedures or guidance have been developed to govern expert panel activities.
 
, Observations and Findinas on Risk Determinations
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j b.2.1 Analvtical Risk Determinina Methodoloav i
e During the inspection, the team reviewed the Davis-Besse Individual Plant Examination (IPE), Individual Plant Examination of External Events (IPEEE), and l Safety Evaluation Report (SER) of the IPE, and interviewed the IPE representative C The IPE was a small event tree and large fault tree model, and the "CAFTA" computer. code was esed to develop and quantify the model. The team judged the ( '
IPE to be acceptable f or support of the MR. (It should be noted that the SER concluded that the ir E met the requirements of Generic Letter 88-20, " Individual '
Plant Examination 'or Severe Accident Vulnerabilities," and associated guidance in NUREG-1335, " Individual Plant Examination: Submittal Guidance.") However, the IPE, submitted in February 1993, had not been updated to support the M Therefore, IPE design, unavailability, and unreliability data were representative of the plant prior to 1991. Not having updated the IPE to reflect more recent plant data was considered by the team to be a weakness in the implementation of the M b.2.2 Adeauacy of Expert Panel Evaluations The licensee's process for establishing the risk significance of SSCs within the scope of the MR was documented in the Davis-Besse MRPM, Section 5.6 and Attachment 7, intra-company memorandum NEN-95-10115, and minutes of the Working Group (expert panel) meetings. These documents were reviewed and found to adequately describe the process of determining risk significanc For SSCs modeled in the licensee's IPE, three importance measures were evaluated (core damage frequency contribution, risk achievement worth, and risk reduction worth), as recommended in NUMARC 93-01. The licensee first evaluated the IPE basic events relative to the core damage frequency contribution criterion. Then the remaining IPE cut sets and basic events were evaluated with respect to the other two importance measure criteria. If a basic event's importance measure met one or
 
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more of the criteria, then the SSC associated with that basic event was judged to be potentially risk significant. Because NUMARC 93-01 guidance indicated that a i basic event was potentially risk significant if any of the three importance measure i criteria were met, the approach used by the licensee was acceptable. The Working {
Group then made the final determinations with respect to risk significance. Several '
SSCs indicated to be risk significant from the IPE importance measures were downgraded to non-risk significant. Documentation of these cases was provided in Attachment 7 of the MRPM and NEN-95-10115. The team found that sufficient I information was presented in those documents to justify the downgrading of those !
SSC :
For SSCs not modeled in the IPE, the Working Group determined the risk j significance. Although the final risk significance results were appropriate, the i details of the Working Group determinations were iot adequately dccumente l Finally, the original IPE (February 1993) was used to support the risk significance determinations. That IPE generally reflected plant conditions before 199 Therefore, a weakness in the licensee's risk significance determinations was the use of an IPE that had not been updated since its original submitta !
I Conclusions on Risk Determinations Based on the reviews discussed above, the team determined that the licensee's approach to establishing the risk ranking for SSCs within the scope of the MR was adequate. However, weaknesses in that approach included the use of an outdated :
IPE and inadequate documentation of the Expert Panel's determination .
j M1.3 (a)(3) Periodic Evaluations    I Insoection Scone Section (a)(3) of the rule requires that performance and condition monitoring {
activities and associated goals and preventive maintenance activities be evaluated, !
taking into account where practical, industry wide operating experience. This evaluation was required to be performed at least one time during each refueling cycle, not to exceed 24 months between evaluations. The team reviewed the ;
procedural guidelines for these evaluations and two periodic report ' Observations and Findinas The licensee's instructions for conducting periodic evaluations were contained in procedure DB-PF-00002 " Preventive Maintenance Program," Revision 00, Section 6.4. The procedure generally provided adequate guidance for preparing evaluations which would meet the requirements of 10 CFR 50.65(a)(3) and the intent of NUMARC 93-01. One identified exception involved the MR requirement to I incorporate industry experience into the periodic evaluation. DB PF-00002, Section 6.4.1.e specified three sources to be used, thus procedurally eliminating other, equally good sources of industry experience. The team noted that the licensee's operating experience assessment program (discussed in section M1.8)
was not as restrictive. After discussing this with the MRC, the team concluded
 
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I that this was inadvertent. The MRC indicated that the step would be corrected to remove the restrictio The team reviewed the following:
* Cycle 9 Periodic Maintenance Effectiveness Assessment Period (July 1993 - November 1994)
* Cycle 10 Periodic Maintenance Effectiveness Assessmelt Period
; (November 1994 - May 1996)
The team noted that the cyclic periodic reports contained an overall plant summary and individual, system :,pecific periodic reports for all the systems within the scope
,
of the MR. Both the summary and the system-specific reports addressed all of the aspects specified in NUMARC 93-01 and 10 CFR 50.65(a)(3). The team considered the preparation of system-specific periodic reports as a significant i strengt j l Conclusions
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The procedures for performing periodic evaluations met the requirements of the rule !
I h
and the intent of the NUMARC implementing guidance. The team noted that the reports addressed all of the aspects specified in NUMARC 93-01 and 10 CFR 50.65(a)(3). The team considered the preparation of system-specific periodic reports as a significant strength,    i M1.4 (a)(3) Balancina Reliability and Unavailability Insoection Scooe Paragraph (a)(3) of the MR requires that adjustments be made where necessary to assure that the objective of preventing failures through the performance of preventive maintenance was appropriately balanced against the objective of minimizing unavailability due to monitoring or preventive maintenance. The team reviewed the licenset's plans to ensure this evaluation was performed as required ;
by the rul i Qbservations and Findinas    ;
 
The licensee's basis for balancing reliability and availability was contained in Section 13 of the MRPM and reflected a qualitative method that was consistent with NUMARC 93-01 guidance and industry practic The licensee's a;proach for optimizing availability and reliability was achieved by properly scheduling testing and maintenance activities to minimize the overall unavailability of the SSC and to maximize the reliability. During refueling cycle 10, l the licensee made approximately one thousand changes to preventive maintenance activities to optimize SSC reliability and availability. Scheduling efficiency was I continually improved by applying lessons learned from operating cycle mini- ;
schedules as found in Work Process Guideline WPG-1 R04. The team also noted I
 
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that evaluations of the reliability / availability balance were included in both the  i j  overall and plant-specific periodic evaluation i
; Conclusions i
 
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The licensee's process for conducting the balance between reliability and ava: lability- l
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was considered acceptable. Changes to the preventive maintenance program were l  indicative of an ongoing effort to optimize reliability and availabilit M1.5 (a)(3) On-line Maintenance Risk Assessments
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j Insoection Scope
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*
Paragraph (a)(3) of the MR requires that when removing plant equipment from a
service the overall effect on performance of safety functions be taken into account.
 
_
 
The guidance contained in NUMARC 93-01 required that an assessment method be developed to ensure that overall plant safety function capabilities were maintained  l i  when removing SSCs from service for preventive maintenance or monitorin Observations and Findinas The licensee's process for determining plant safety when equipment was taken out of service was documented in the' MRPM, Section 8 and Attachment 2. Whilc the plant was at power,' the licensee's process was outlined in the Work Process (
Guideline, WPG-1, which is supported by intra-company memorandum NEN-95-  i 10113. When the plant was shut down, the process was outlined in Outage 4  Nuclear Safety Control, NG-PS-0011 While the plant was at power, the " Risk Significant System Matrix" (Attachment 9 in W.PG-1) was used by work planners and Senior Reactor Operators (SROs) to evaluate plant risk for concurrent outages of risk significant SSCs. The risk matrix was developed with the support of the IPE, as outlined in NEN-95-10113. The licensee used a 12-week rolling schedule for planning surveillances and preventive  ,
maintenance. The work planners used the risk matrix to prevent planned concurrent equipment outages that would place the plant in a potentially high risk situation (indicated by "X"in the risk matrix). The SROs performed a final evaluation of planned outages against the risk matrix before each work week began. For combinations of equipment outages not covered by the risk matrix, SROs and work planners used their experience and judgment to evaluate the plant risk. (A review of requests for risk determinations identiRd :ase in which a work planner requested a risk determination for planned cc'  :tages of SSCs not all covered by the risk matrix.)
 
The risk matrix used by the licensee was considered t  . in terms of effective use of IPE information to evaluate plant risk frorn concurrent equipment outages. Specific issues were the following:
  * The matrix provided no explicit guidance for assessing plant risk when three or more pieces of equipment were out-of-service at the same time. Such combinations might place the plant in a high risk situation without the user
 
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i  realizing this. (It should be noted that the users of the risk matrix indicated that they would request a risk determination for outages of three or more risk significant SSCs. However, that guidance was not in the procedure.)
 
'
s There was no guidance for recovery from high risk configurations resulting l  Trom emergent failures (no guidance on determining which piece of
;  equipment to return to service first).
 
l The procedures used by the licensee for plant shutdown conditions appeared to be
 
.the standard industry approach, based on NUMARC 91-06, INPO guidelines for
!
outage management, and EPRI guidance.
 
i        l i Conclusions i
 
;~ .The team viewed the licensee's process for assessing plant risk resulting from multiple equipment outages to be appropriate. However, the tool used to assess plant risk while at power, the risk matrix, was viewed as weak in terms of making use of IPE risk information and providing user guidanc l M1.6 (a)(1) Goal Settina and Monitorina and (a)(2) Preventive Maintenance
:
' Insoection Scoce The team reviewed program documents in order to evaluate the process established d to set goals and monitor under (a)(1) and to verify that preventive maintenance d (PM) was effective under (a)(2) of the rule. The team also discussed the program '
with appropriate plant personnel. The team reviewed the following systems:
  (a)(1) systems Switchyard / Transformers Freeze Protection Component Cooling Water Control Room Emergency Ventilation System Safety Features Actuation System (a)(2) systems Turbine-generator Auxiliary Feedwater Radiation Monitors Service Water The team reviewed each of these systems to verify that goals or performance criteria were established in accordance with safety, that industry wide operating experience was taken into consideration where practical, that appropriate monitoring and trending were being performed, and that corrective actions were taken when an SSC failed to meet its goal or performance criteria or experienced a maintenance preventible functional failure (MPFF). The team also reviewed performance criteria for SSCs not listed abov .. _ . . _ . _ _ - _ __ .. . _ . ~ . - _ . _ _ _ _ ._ __.
 
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The team reviewed the licensee's process to evaluate onsite passive structures for
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inclusion under the MR. Structures evaluated by the team included buildings, enclosures, storage tanks, earthen structures, and passive components and ,
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materials housed in the aforementioned. . in addition, the team assessed by what .
means performance of structures determined to be within scope were monitored for
[  degradation.
 
' Observations and Findinas
:
I  The team found that the plant MRPM, Revision 2, provided appropriate guidelines
;  for establishing performance criteria / goals for SSCs scoped under the MR. The
}  licensee had established performance criteria and/or goals for all systems
;  . designated within scope. The performance criteria / goals were documented and i  -retrievable. System engineers were knowledgeable of the performance
)-  criteria / goals for the assigned systems. The team did not identify any functional l failures that were not previously identified by the licensee. However, as discussed j  below, the team identified cases where reliability related performance criteria had not been properly established for certain risk significant SSCs.
 
ll 4 Performance Criteria for Unavailability
)  Section 9.3.2 of NUMARC 93-01 recommended that risk significant SSC t'
performance criteria be set to assure that the availability and reliability assumptions used in the risk determining analysis (i.e., PRA) were maintained. The team .
evaluated the licensee's performance criteria to determine if they had been -
;  adequately set under (a)(2) of the MR, consistent with the assumptions used to
;  establish SSC safety significance.
 
.
!!  The team reviewed the 29 licensee-specified risk significant SSCs, and identified
'
that reliability performance criteria had been set (the adequacy of these reliability
). criteria is discussed below). For ten risk significant SSCs, the licensee had not I established specific unavailability criteria. For these SSCs, any unavailability would l  have caused a plant trip or required a plant shutdown (e.g., DC power, main steam, l  reactor coolant, steam generators) and, as such, the licensee considered monitoring i
unavailability as not meaningful. Problems with these systems would be adequately
;  monitored by functional failures. This was considered reasonable and acceptable.
 
,  The team noted that the majority of unavailability goals and performance criteria
.
established for high safety significant SSCs were less conservative than the
!  unavailability values assumed in the PRA.~ Unavailability performance criteria i generally were twice the PRA value. The licensee had recalculated tha Core
!  Damage Frequency (CDF) value using the MR performance criteria unavailability
,  values. The results of Calculation No. C-NSA-99.16-19, Rev.1, indicated that the CDF changed from 6.57E-5 to 5.76E-5. Based on this slight increase (11%) in the i  CDF, the team determined that the unavailability performance criteria established for high safety significant SSCs were acceptable.
 
l  The MR ur$ availability performance criteria in most cases were based on a 3-year average unavailability multiplied by two. As a result of the rule, system engineers have become more attuned to PM on systems taken out-of-service, such that PM
,
;    12
:
i-
 
_ _ _ . _ __ _ _ _ . _  . . _ . _ _ _ _ . . _ . _ _ . _ _ _ . . _ _  -_
.
J N
l . activities were being consolidated, frequencies revised, and scheduled during i
system min-outages to reduce system unavailability. As a result, unavailability was reduced on a number of systems, however, the performance criteria were still j
based on historic data. Although unavailability criteria did not significantly affect
:
'  CDF, the licensee should ensure the performance criteria established remains reaIsonable based on the improved PM methodology.
 
i l Performance Criteria for Reliability i
The licensee established, as reliability related performance criteria, no repetitive FFs,
;  and either greater than zero or greater than one FFs per cycle for 26 of the 29 risk
:  , significant SSCs. The use of FFs was considered better than MPFFs. However, at
.
the time of the inspection, the licensee had not performed a sensitivity analysis that
;
;
demonstrated that the performance criteria used for reliability preserved the assumptions used in the PRA, or that the use of these reliability performance j
criteria did not have an adverse impact on risk ranking. The team noted that there
,
was no relationship established between these criteria and the failure probability l  assumptions in the PRA, since the number of function demands and/or equipment
:  run times were not considered.' Thus, widely different SSC reliability estimates
  - (probability of failure upon demand or per hour) could result from the same number 3  of FFs in a specified time period if the number of demands or operating times varied i
between periods. The licensee failed to determine the sensitivity of the PRA to
,
these unreliability performance criteria.
 
!
;  As such, the failure to couple the number of FFs to the failure probability
:
assumptions in the PRA was considered to be a violation of 10 CFR 50.65(a)(1),
failure to define performance monitoring criteria which demonstrate acceptable
;
.
performance commensurate with safety (VIO 50 346/97002-01(DRS)).
;  For two SSCs, emergency diesel generator (EDG) and standby blackout diesel
*
generator, a 95% reliability performance criterion was established. For the
;  containment system, the reliability performance criteria were based on leakage l  rates. These criteria were considered acceptable reliability performance criteria.
 
. Performance Criteria for Non-risk Sianificant Normally Ooeratina SSCs The licensee had established specific performance criteria (FFs/ cycle) for on-risk
,  significant normally operating SSCs, although the guidance in NUMARC 9:.-31 only required plant-level performance criteria for these SSCs. This was considered a significant improvernent over the NUMARC guidance in that the FF criteria would better assess the SSC being monitored. The placement of the freeze protection system into fo)(1) due to several FFs was the result of establishing SSC specific criteri The licensee also established six plant-level performance criteria that included the following:
o unplanned automatic reactor trips per 7000 critical hours e unplanned unit capability loss factor e unit capability factor
 
-- , , - -    .- . . _ - - -
 
  -.  . .
    . ._ . _ .. - _ .. ~ _ _ . _ _ _.. _ _ _ _ ~
        '
a .
.
E
.
        :
! ' e adjusted heat rate
 
e containment leakage e non-risk significant (a)(2) functional failures
.
If.a plant level performance criterion was exceeded, an evaluation woud be i
performed to determine the cause. If the evaluation could determine a specific SSC as the root cause for the performance criterion being exceeded, ther that SSC
! would be considered for transfer to category (a)(1). l
*
q
: b.4 - Goals Established for (a)(1) SSCs The goals established for all SSCs placed in (a)(1) were the same as the previously established performance criteria for that SSC. The licensee had determined that the performance criteria were adequate to monitor the SSC's corrective actions such -
that additional goals were not necessary. The team concurred that in these cases the performance criteria were acceptable goals. The licensee also stated specific goals would be established if the established performance criteria would not  i effectively monitor the SSC's corrective action The licensee had dispositioned two SSCs from (a)(1) to (a)(2) by redefining what constitutes a FF of the SSC. This change, which required a system / train failure versus a component / channel failure resulted in not classify:ng several component / channel failures as FFs, such that the SSCs' performance criterion was -
not exceeded. The team did not have a concern with the licensee's position on what was considered a FF in these cases. These examples are further discussed in section M2.1.b of this repor t 1 Structures and Structure Monitorina The team reviewed procedure M/S DG-26, " Design Guideline for Maintenance Rule Evaluation of Structures," and other associated licensee programmatic controls to determine which onsite structures were evaluated for inclusion under the rul ,
Additionally, a review of the' performance criteria and monitoring established for '
structures within scope was performe The team identif!9d several good aspects of the structure monitoring program including: 1) scope included the structures required to be monitored by the MR, 2) contained specific checklist guidance for evaluating structures, including some quantitative performance criteria,3) required qualified individuals to conduct the walkdown, including a licensed professional engineer,4) divided structures within ,
scope of the rule into individual rooms to ensure all areas would be inspected, and !
5) used existing programs to support structural monitorin The licensee's initial structural walkdown frequency was selected as every four years and would be revised based on the results of the baseline walkdowns. Some structures, such as the cooling tower and intake structure, were already on a refueling outage frequency based on previous existing programs. Although the baseline structural walkdowns have not been completed, the licensee did have a  !
~ draft schedule to complete the remaining walkdowns by 1998. The team reviewed  I the walkdown checklists completed for the containment building during the last
 
         ,
         ,
 
h
.
        {
        ;
e 4 .
outage and noted that minor deficiencies that did not affect the structures function were being documented, and resolved if required. No major concerns were Identified by the licensee on the containment structure * where walkdowns have .
been complete l
        !
 
Although there appeared to be a good foundation for the structural monitoring l
program, two issues were also identified. The program's baseline inspections only concentrated on the buildings' wpports (i.e., walls, ceiling, structural steel) and not i  supports (e.g., pump pedestals, pipe supports) contained within the building, inspection of these supports was left to the system engineers; however, the i
system engineering handbook did not specifically address this requirement when I  conducting walkdowns. Although several system engineers indicated they look at supports during the weekly walkdowns, the team was concerned that these walkdowns may not be adequate to baseline these supports for the MR.
 
;' s
;  The second issue involved dispositioning a structure from (a)(1) to (a)(2) or (a)(2) to
'  (a)(1). The program stated these decisions would be evaluated by the expert panel, however, there was no guidance in the program for the expert panel to use to make
;
this evaluation. Concerns with the lack of expert panel guidance were previously j  discussed in section M1.2.b.1 of this report.
 
i'
Since the baseline inspections for the structure monitoring program were not
.
sufficiently completed to evaluate whether these two issues conform with
 
regulatory requirements or industry guidelines, this was considered an unresolved (
.
item pending further NRC review of the baseline inspections and the dispositioning (
!  of structures between the (a)(2) and (a)(1) categories under the Rule (URI 50- ;
346/97002-02(DRS)).
.
; Conclusions
,
s
.
Ti3e establishment of performance criteria and goal setting was considered (  satisfactory. The use of specific FF performance criteria for non-risk significant SSCs, along with maintaining plant level performance criteria was a program i  strength. Relhbility performance criteria for safety significant SSCs, however, were deficient in that the FF criteria were not based on demands or run tim Based on the'recent efficient use of outage times for several SSCs, the affectiveness of the unavailability performance criteria may have been reduced.
 
The licensee had adequately scoped buildings and enclosures as structures under the Rule. However, the structure monitoring program concentrated only on building supports and not internal component supports and no guidance was established for moving structures between the (a)(2) and (a)(1) categories under the Rule.
 
t i
M1.8 Use of Industrv-wide Oneratina Exoerience
  , Insoection Scoce
:
'
Paragraph (a)(1) of the rule states that goals shall be established commensurate
. with safety and, where practical,iaking into account industry-wide operating j  experience. Paragraph (a)(3) obthe rule states that performance and condition
.
(
 
,
i
 
    -. .- . . . - . . . -.
.
.
monitoring activities and associated goals and PM activities shall be evaluated at least every refueling cycle. The evaluation shall be conducted taking into account industry-wide operating experience. The team reviewed the licensee's program to integrate industry operating experience (IOE) into their monitoring program for maintenanc Qhservations and Findinos on Use of Industrv wide Ooeratina Experience The licensee procedure NG-NA-00305, Revision 01, " Operating Experience Assessment Program," provided the administrative guidelines to integrate industry-wide operating experience, in addition to the processing of in-house experience that may be of interest to the industr Interviews and the review indicated that a structured process existed for evaluating and processing IOE. The process transferred information to the SEs, about events that were received by the operating experience (OE) group, in addition, OEs were also screened separately by the Nuc!cm Network Coordinator. AL OEs were also summarized in a routine OE Report (OER) which was prepared by Independent  i Safety Engineering and distributed to a wide range of interested and cognizant individuals. This process also included a monthly memo to file which documented ;
the disposition of all operating experience reports that were recsived the preceding l month. This memo also identified OERs deemed not applicable to Davis-Bess l The formal review included the completion of an OER Screening Checklist by the independent safety engineering reviewer. The team found that SEs were able to discuss the program, formal and informal, and how they used the information to identify system improvements as well as the mechanism to process the sharing  i with the industry of in-house informatio I Conclusions for Use of Industry wide Ooeratina Exoerience The team concluded that the licensee had properly integrated the MR into the existing lOE program. Adequate provisions had been made to incorporate  ,
information from the IDE program into the periodic evaluation, goal development, j and functional failure evaluation '
M2 Maintenance and Material Condition of Facilities and Equipment (61703,71707)  l M2.1. _ General System Review Insoection Scop _q The inspectors conducted a detailed examination of several systems from a MR perspective to assess the effectiveness of the licensee's program when it was applied to individual system Observations and Findinas for the Switchvard/ Transformers The team reviewed the performance criteria for the switchyard and transformers and noted that the licensee had established a reliability performance criterion of one
    '
 
.
functional failure per cycle. The system had been correctly classified as risk significan The system 'was placed in category (a)(1) in the second quarter of 1995 due to experiencing two failures that were initially classified as FFs. Breaker 34562 i
' developed an air leak which rendered the breaker inoperable. The licensee rebuilt all five of the 345kV breakers. Although breaker 34562 was inoperable, ability to supply power from the ring bus was not lost, so the licensee determined that the switchyard did not experience a functional failure. Another failure for the SSC was I the deenergization of startup transformer XO2. A mayfly swarm, attracted by i lighting in the vicinity of the transformer, initiated a flashover on the A phase I lightning arrestor. The licensee cleaned and tested the transformer and returned it -
to service. Lighting in the vicinity of the transformer was redirected to avoid ;
attracting mayfly swarms. The licensee reclassified this as a nonfunctional failure
      .
I since the SSC was still able to perform its function of supplying power to both essential and non essentialloads. Based on the reclassification of these events the system was moved to category (a)(2). The team noted that the material condition of this SSC was good and that it was functioning 'as designe l c.1- Conclusions for the Switchvard/ Transformers    '
The licensee's initial classification of this system as (a)(1) was conservative; l subsequent reevaluation of the FFs and reclassification of the system as category (a)(2) was appropriat G
      $ Observations and Findinos for the Freeze Protection System The team reviewed the established performance criteria for the freeze protection / heat trace system. The team found that the licensee had established a performance criterion of three or less FFs per operating cycle. As noted in paragraph M1.6.b.3, plant level performance criteria would have been acceptable; however, as discussed below, the application of specific criteria resulted in better system monitorin The freeze protection system had been dispositioned from category (a)(2) to (a)(1)
due to experiencing six FFs in cycle 10, including repetitive FFs caused by blown fuses. The SSC was still experiencing FFs in operating cycle 11 due to fuse failures. Most of the FFs had been caused by personnel plugging portable  ,
equipment into the duplex receptacles with the freeze protection circuits. The i licensee had labeled the freeze protection associated circuits to avoid usage that could cause fuses to fail. The SSC was stillin category (a)(1) at the end of this inspectio Conclusions for the Freeze Protection System l
The application of specific performance criteria to this low risk, normally operating !
system resulted in identification of system problems, categcrization as (a)(1), and the development of focused corrective action. At the close of the inspection, the effectiveness of this action had not yet been determine . . - . . .. - .- - . --
    ., _-.. - . . - - - _ - - .- - - .. - -. . -
 
        !
, .        l l',
1 Observations and Findinas for the Turbine-Generator System
,
The team reviewed the performance criteria for the turbine-generator system and
!
noted that the licensee had established a criterion of less than or equal to one FF per cycle, with no repetitive FFs.
 
f This SSC had been in category (a)(2) for the previous two operating cycles. The l  system had been operating successfully w!th no automatic trips for the two
;  previous cycle l Conclusions for the Turbine-Generator System    '
The team concluded the turbine-generator system was appropriately classified and that system performance was good.
 
, Observations and Findinas for the Comoonent Coolina Water (CCW) System j
'
The team reviewed the performance criteria for the CCW system and noted that
'
reliability criteria (less than one FF/ cycle) had been established on a system basi I Unavailability criteria had been established on both a pump / system basis (less than i
109 hours / year) and average ventilation train basis (less than 147 hours / year). The
:
system was initially placed into (a)(1) category due to averege pumo unavailability
;  being greater than the performance goal of 109 hours per year. This was due i
primarily to the past practice of performing work only on one shif i
*        ;
The system was changed to category (a)(2) based on decreased system i  unavailability, due in part to performing work on an around-the-clock basis to
*
minimize outage duration. There were no functional failures during cycle 1 .
l
,
The team walked down portions of the component cooling water system areas with  l l
' '
the system engineer and noted that the material condition of the equipment and the housekeeping were satisfactor ' Conclusions for the Comoonent Coolina Water System The team determined that reclassification of the system from category (a)(1) to (a)(2) was appropriate. The material condition of the system was satisfactor Observations and Findinas for the Auxiliarv Feedwater (AFW) System
        >
The team reviewed the performance criteria for the AFW system and noted that both reliability (less than or equal to one FF/ cycle) and unavailability (less than 130 hours / year) criteria had been established. The MR requirements were met by the AFW system, with average train unavailability being maintained less than the MR requirement. No FFs had been identifie The team walked down portions of the AFW system with the system engineer and noted that the material condition of the equipment and the housekeeping were satisfactor __
 
- . - . -~ ._ -. . . .- -, . -  . -- -. -. . . _
  .
. Conclusions for the Auxiliary Feedwater Svstem The team concluded that the AFW system had been appropriately addressed under the licensee's MR program.
 
j Observations and Findinas for the Radiation Monitors  *
The team reviewed the performance criteria for the radiation monitors and noted
 
that the licensee had established a criterion of less than or equal to three FFs per cycle with no repetitive FF The age of the Geiger-Mueller tube and ion chamber detectors made them susceptible to failure, however, the current failure rate was considered by the licenses to be acceptable. The detectors were replaced on an as-fail basis.
 
Due to the poor performance of the containment radiation rnonitors, the licensee was considering requesting a Technical Specification change to delete the monitors'
trip function or a modification to replace the monitors with a reliable model.
 
!  The team walked down portions of the radiation monitor system with the system engineer and noted that the material condition of the equipment and the
,
housekeeping appeared satisfactory, Conclusions for the Radiation Monitors i
        (
The team concluded that the radiation monitoring system had been appropriately
.
addressed under the licensee's MR program.
 
,        ,
! Observations and Findinas for the Safetv Features Actuation System (SFAS)
 
The team reviewed the performance criteria established for the SFAS and noted
!
that the licensee had established less than or equal to one risk significant FFs per
,  cycle or less than or equal to two non-risk significant FFs per cycle as reliability performance criteria. No unavailability performance criterion was established as the system was considered operational basically 100% of the time. The licensee did not establish FFs per run time as the reliability performance criteria for this high
 
safety (risk) significant system.
:
;  The SFAS system had been placed in (a)(1) due to FFs exceeding the performance q  criterion. The FFs concerned failures of radiation monitoring channels. The l  licensee subsequently revised the criteria for what was considered FFs. The evaluation determined that the failure of one channel would not prevent the system 1  from operating, such that a FF would now be defined as the failure of one train
,
  (two channels). This definition would also be applicable to the reactor protection system (RPS) and steam and feedwater rupture control system, and consistent with the other systems within the MR scope. Based on this revision, there were no
'
longer any FFs of SFAS and the system was returned to (a)(2). The team did not j  have any concerns with the evaluation or with returning the SSC to (a)(2).
 
19 l
 
_ _ . _ . ._ _ - . ~
.  .
i -
.
  . . Conclusions for the Safetv Features Actuation Systen
 
The' team concluded that the licensee's reclassification of the SFAS from category i
  (a)(1) to (a)(2) was appropriate and that the system was being properly addressed n  by the licensee's MR program.
 
,
: Observations and Findinos for the Service Water (SW) System
;.  .
;F
.
The team reviewed the established performance criteria for the SW system and noted that the licensee established a reliability performance criterion of less than or g  equal to one FF per cycle, an unavailability performance criterion of 283 hours per
;
'
year average for.the three SW pumps, and an unavailability performance criterion of 865 hours per year for the dilution pump. The team found that the licensee did not
)-  establish FFs per number of demands or per run time as the reliability performance l  criteria for this high safety (risk) significant system.
 
;  The unavailability hours were determined by a 3-year average and multiplied by two for both the SW and dilution pumps. The majority of the unavailability hours came
.
from scheduled 3-week PM every five years taken on each of the pumps. This was l-  the reasoning behind the performance criterion being the average unavailability for F
the three pumps, as the PM outage would automatically have a single pump exceed
!  the performance criterion. The licensee also stated that if one SW pump exceeded the 283 hours and was not the result of the 3-week PM, that pump would be n  considered for (a)(1). For Cycle 11, the SW pumps averaged 105.3 hours per year,  {
i  while the dilution pump averaged 107.1 hours per year. The unavailability criterion 4 for the dilution pump appeared high based on the recent data. Although the high
        .
r omber was understandable because of the 3-week PM, there could be a masking  '
effect of pump problems that would not be evaluated by the unavailability criterion
 
estvo!ished.
 
I Although there were no FFs associated with this system during the last cycle, there
'
 
were a number of significant issues that were being addressed by engineering and included the following:
!  * Several Velan gate valves had experienced severe corrosion / galvanic action
[    to the valves' stem / wedge interface that resulted in reduced flows to the emergency core cooling system (ECCS) room coolers. The licensee replaced I
a number of these valves to ensure a flow path for the coolers and intended to replace other valves in the system that also may be affected.
 
j  * The identification of zebra mussel shells in the CCW heat exchangers due to the loss of the chlorination system for 30 days in June 1995. Although there .was no operability issue with the SW system resulting from shell i    blockage, the licensee conducted component inspections and flushed the
!
system using higher than normal concentrations of Sodium Hypochlorite. As
;  long term actions, the chlorination system was rebuilt and included as part of the SW SSC with respect to the MR, including possibly developing
 
specific performance criteria for the chlorination syste l.
 
l      20
 
4
, , . , .,e.an v , - , ,  ,.,....,r,. , , ,-.  . , , - . - ,_,._,n
 
. .-  . -. ~. . - - . .. - - . . . _ _.- .-.
.
.
s o  e As the result of intermittent low /high level alarms for the intake screens, the licensee identified 2-4 feet of silt in the intake bay. The bays were dredged
'  to remove the silt and a refueling outage inspection / cleaning was instituted ,
for the intake structure. A recent inspection by the licensee identified i
[  degradation of the intake crib in Lake Erie, which provided make-up to the
,
;
ultimate heat sink. Licensee action on this issue had not been formulated at the time of the inspectio ,
' Conclusions for the Service Water System    j i
The team noted that a number of issues represented potential performance problems for the SW system. A!though the licensee's unavailability criteria were
 
'  acceptable under NUMARC 93-01 guidance and MRPM guidance, there was
*
potential for system problems to be masked by incorporating the 3-week PM outage into the unavailability criterion.
: Observations and Findinas for the Control Room Emeraency Ventilation System
'
The team reviewed the performance criteria established for Control Room i  Emergency Ventilation System (CREVS) and noted that the licensee had established  j
:  a reliability performance criterion of less than or equal to two FFs per cycle and an
;  unavailability performance criterion of 204 hours per year per trai CREVS was an (a)(1) system due to three FFs during cycle 10. Two FFs were
!
refrigerant leaks, both of which were considered repetitive. The first concerned a .
.
diaphragm failure on a valve actuator. The same valve had failed a year earlie The failures were attributed to a manufacturing defect and additional degradation !
during shipping, storage, or valve assembly. A PM was generated to replace the diaphragms every five years and to inspect the diaphragms prior to installation for defects. The second set of refrigerant leaks occurred due to bumping of tubing
,
during work in the CREVS room. A major modification scheduled for the next i  outage will relocate the control panels to reduce the number of instruments and
.
'
refrigerant tubing lengths, and provide protection for remaining components. The
  . third functional failure was a ground on the #2 chiller condensing fan motor that
;  caused'the loss of both trains of control room normal ventilation. The fan motor  i electrical fault was repaired. Corrective action completed or planned appear adequate to address the problems identified with CREVS.
 
l System engineering continued to reduce system unavailability by consolidating maintenance activities. By performing mini-outages and improving personnel
,  awareness of unavailability requirements, unavailability was reduced from 204
;  hours per year per train to 115 hours per year per train.
 
4 Conclusions for the Control Room Emeraency Ventilation System The team concluded that the classification of the CREVS as category (a)(1) was
;. appropriate and that the licensea's corrective actions, both in progress and planned,
;  appeared adequate. The team considered the reduction of system unavailability a l  direct result of properly implementing the M l
, -  -  _ ._
 
- . . . . . . _ _- . _ . _ _ . _ . _ . . -._ . _..-m ._. -_ . . _ . ~ _ _. _ . . . . .-. . .      -
,
!
.
!      ,
!
  . .
            .
:  M2.2 Material Condition i
t inspection Scone
 
!-
In the course of verifying the implementation of the MR using Inspection Procedure j  62706, the team performed walkdowns using Inspection Procedure 71707, Plant j  Operations, to examine the material condition of the systems listed in section M1.6.
 
, . Observations and Findinas
!'
{  ' Except as noted in Section M2.1, the systems were free of corrosion, oil leaks,
_
'
water leaks, trash, and based upon external condition, appeared to be well-maintaine ,
!-
l  c.' Conclusions i
In general, the material condition of the systems examined was very good,
>
i M7 Quality Assurance in Maintenance Activities (40500)
!  M7.1' Licensee Self-Assessments of the Maintenance Rule Proaram (            ,
            ' Insoection Scone
;-
The team reviewed the report of an assist visit by the Nuclear Energy Institute (NEI)
3  which took place in 1995. The team also reviewed a sample of licensee self-l  assessment activities associated with maintenance. The documents reviewed i  included:
I j'  * Memorandum dated May 19,1995: Maintenance Rule Assessment i  * Memorandum dated September 5,1995: Maintenance Rule Self-Assessment
!.  (Response)
i  * Toledo Edison Quality Assessment Audit AR-96-MAINT-01, dated January 3, 1997 i
j Observations and Findinga    ,
!
!  The May 1995 self-assessment was conducted by a multi-disciplined team, which included members from other utilities. -This approach provided an independent viewpoint which added to the overall quality of the assessment. The assessment i  was thorough and identified issues across a wide range of MR implementation aspects. The inspectors considered the findings in the " Program Concerns" and
'
  " Regulatory Concerns" to be especially valuable. Of particular note were the
<
'
concerns expressed regarding the guidance for expert panel activities. While some corrective action was taken by the MR staff, the inspectors independently noted
. that guidance for and documentation of expert panel activities were still wea The team noted that MR implementation was a portion of Audit AR-96-MAINT-01, j-  which was a comprehensive examin'ation of the station's maintenance progra :  The team noted that NRC laspection Procedure 62706, NUMARC 93-01, and other
.
A
 
,
v4 x  s- -
es-.-- -  ,, .- ,-.n-.,--,e-. -, e .m..a e-.m., um-,-, ,.- -- m e-
 
__
  - . - . -. -.-
,
*
i-
,
current guidance were used in conducting the' audit. The team reviewed the portion of the Executive Summary pertaining to the MR and the two specific findings (AR-96-MAINT-01-01 and 02) and noted that while it correctly identified the need to
,
schedule and conduct baseline structural inspections, it did not identify the need to
!.
establish criteria for reclassifying structures from category (a)(2) to (a)(1). The team noted that corrective actions for the findings were either in place or in
{ ' progress.
 
!
- Conclusions i
The team concluded that the licensee's self-assessment activities were appropriately conducted and identified worthwhile issues. Use of independent
    .
personnel was considered a strength of the 1995 self-assessment. The team further concluded that Audit 96-MAINT-01 was also a worthwhile effort although it did not identify a significant deficiency in the structure monitoring program.
 
*'
ll1.- Enaineerina
 
,
E4 Engineering Staff Knowledge and Performance (62706)
 
; E Enoineer's Knowledae of the Maintenance Rule j InsoecticrtS00De (62706)
l-  The team interviewed system engineers and managers to assess their
!  understanding of PRA, the MR, and associated responsibilities.
 
'
l Observations and Findinos The team interviewed the system engineers assigned responsibility for selected SSCs, walked down systems with them, and determined that they were i  knowledgeable of their systems, the MR performance criteria for their systems, and j  where their systems stood with respect to the prformance criteria. They had
!  received training in the use of the PRA in risk assessment fo- their SSCs and
;-
appeared to be able to use applicable PRA data in assessing risi.s involving their i
'
systems. Each of the system engineers appeared to be knowledgeable concerning
*
MR requirements for their respective systems. This was espec. ally apparent in their understanding of unavailability, how to track it, and how to us9 it most effectively
'
in monitoring system performance. System engineers also inJicated thct the MR was a useful toolin bringing poorly performing systems to the attention of station y  management.
 
t Conclusion _ .The team noted that system engineers were generally experienced and
;  knowledgeable, and their understanding of the MR and PRA was good. The team ,
, also noted that system engineers were actively involved in MR implementation and
;
-s
  ,
generally viewed the MR as a useful tool.
 
l
,-
 
  . .- -.- . -_ - .- .- . .  . .. .-
.
J V. Manacement Meetinas
      ,
l X1 Exit Meeting Summary    l
^
The team discussed the progross of the inspection with licensee representatives on a daily basis and presented the inspection results to members of licensee management at the conclusion of the inspection on January 17,1997. The licensee acknowledged the fint!ings presante The team asked the licensee whether any materials examined during the inspection should be considered proprietary; none was identifie a
      .
      )
 
  ,
 
.. .. -..- ..~ - . - --  - - . - _ - _
      . . - ~ . , . - .
; -
p .
f
 
PARTIAL LIST OF PERSONS CONTACTED I      ,
! Licensee
,
J. Arora, Senior Engineer      1 j' D. Alley, Quality Assurance Supervisor      {
M.' Beier, Manager - Quality Control      {
,
W. Bentley, Shift Manager K. Byrd, Senior Engineer i
*
R. Coad, Superintendent - Radiation Protection
#
L. Dohrmann, Manager - Quality Services R. Donnellon, Director - Engineering and Services
-
D. Eschleman, Manager - Operations
! J. Fawcett, Operations Coordinator
'
J. Freels, Manager - Regulatory Affairs l B. Gallatin, Test Engineering i G. Gillespie, Superintendent - Chemistry
-
T. Gulvas, Shift Manager _
 
L. Hughes, Manager - Davis-Besse Supply l C. Kraemer, Engineer - Licensing 1 D. Kuhtenia, Senior Engineer B. Lakis, Shift Manager T. LeMay, Project Coordinator J. Lash, Plant Manager S. Lewis, IPE Consultant      44 D. Lockwood, Supervisor - Compliance      i A. McCallister, Maintenance Rule Coordinator J. Michaelis. Manager - Maintenance J._O'Neill, Supervisor - Quality Control
  .
J. Pearson, Reactor Operator M. Roder, Operations Work Control Supervisor J. Rodgers, Manager - Plant Engineering R. Uebbing, Maintenance Advisor J. Wood, Vice Present - Nuclear      !
G. Wolf, Engineer - Licensing F. Zurvalec, Senior Engineer - Compliance R. Zyduck, Manager-Design Bases Engineering    l N. tl A. Dunlop, Reactor Inspector, Rlli S.~ Eide, Consultant, INEL M. Farber, Reactor inspector, Rlli R._Frahm, Jr., Reactor Operations Engineer, NRR D. Jones, Reactor inspector, Rlli J. Neisler, Reactor inspector, Rlli S. Stasek, Senior Resident inspector, Rill
 
._    __ - --
      .. - - _ . -. , _ _ .
 
- .  . . .. -. . . _ -. = -- .
'
        ,
        '
.
LIST OF INSPECTION PROCEDURES USED IP 62706: Maintenance Rule
: IP 40500: Effectiveness of Licensee' Controls in identifying, Resolving, and Preventing j  Problems
'
IP 71707: Plant Operations    j LIST OF ITEMS OPENED 50-346/97002-01(DRS) VIO " Reliability Performance Criteria" 50 346/97002-02(DRS) URI " Structure Monitoring Criteria" i
l  LIST OF ACRONYMS USED t
AFW Auxiliary Feedwater CDF Core Damage Frequency CF Code of Federal Regulations
,
CREVS Control Room Emergency Ventilation System
'
CCW Component Cooling Water DRS Division of Reactor Safety ECCS Emergency Core Cooling Systems EDG Emergency Diesel Generators EOP Emergency Operating Procedure
] EPRI Electric Power Research Institute FF Functional Failure    (
        ;
,
'
IFl inspection Follow-up Item    '
INPO Institute of Nuclear Plant Operations lOE Industry Operating Experience
"
IPE Individual Plant Evaluation
: IPEEE Individual Plant Evaluation of ExternaIEvents i MPFF Maintenance Preventable Functional Failure 2 MR Maintenance Rule MRC Maintenance Rule Coordinator
!
MRPM Maintenance Rule Program Manual
'
NEl Nuclear Energy Institute NOV Notice of Violation
 
#
NUMARC Nuclear Management Resource Council NRC Nuclear Regulatory Commission NRR Nuclear Reactor Regulation
'
OE Operating Experience OER Operating Experience Report PDR Public Document Room i PM Preventive Maintenance PRA Probabilistic Risk Assessrnent-QA Quality Assurance RPS Reactor Protection System RSSM Risk Significant System Matrix
-
SE System Engineers
,
SER Safety Evaluation Report
 
SFAS Safety Features Actuation System
~
 
i
 
- .- - -- _ . - . - - .- . . - . - - - - .. .- - . . . - . _ - . _
.
a!
 
LIST OF ACRONYMS USED (cont'd)
SRO Senier, Reactor Operator
:
"
SSC Structures, Systems or Components URI Unresolved item LIST OF DOCUMENTS REVIEWED Maintenance Rule Program Manual, Rev. 02, December 15,1996 Individual Plant Examination for the Davis-Besse Nuclear Power Station, February 1993  l l
Individual Plant Examination of External Events for the Davis-Besso Nuclear Power Station, December 1996 l
Staff Evaluation Report Individual Plant Examination Davis-Besse Nuclear Power Station,  !
Unit No.1, Docket No 50-346, October 1996 Work Process Guideline, WPG-1, R04, June 28,1996
    ,
Outage Nuclear Safety Control, NG-DB-00116, December 13,1995 Equipment Out of Service Evaluation, NEN-93-10113, April 21,1995 C
Risk Significant Structures, Systems, and Components (SSCs), NEN-95-10115, May 1,  q 1995      ~
Reliability Performance Criteria, DBE-97-00011, January 9,1997 Monitoring Reliability for the Maintenance Rule, EPRI Technical Bulletin 96-11-01, November 1996 Maintenance Unavailability Sensitivity Analysis (Performance Criteria), C-NSA-99.16-18, Rev.1, January 8,1997 Maintenance Unavailability Analysis (Measured Performance), C-NSA-99.1619, October 15,1996 DB-PF-00002, Preventive Maintenance Program NG-QA-00702, Potential Condition Adverse to Quality-M/S .DG 26, Design Guideline For Maintenance Rule Evaluation of Structures Cycle 9, Periodic Maintenance Effectiveness Assessment Report Cycle 10, Periodic Maintenance Effectiveness Assessment Report Windows Report, Third Quarter 1996
 
      -.
 
__ _ _ . . . . _ _ _ . . . . _ . . - _ . _ _ _ . . .
"
,
'
 
LIST OF DOCUMENTS REVIEWED (cont'd)
 
-System Engineering Handbook  '
      ,
l C-NSA 99.1619, Maintenance Unavailability Analysis (Measured Performance)
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Revision as of 04:39, 28 June 2020

Ack Receipt of 970407 Ltr Informing NRC of Steps Taken to Correct Violations Noted in Insp Rept 50-346/97-02 on 970306
ML20140D147
Person / Time
Site: Davis Besse Cleveland Electric icon.png
Issue date: 04/17/1997
From: Grant G
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To: Jeffery Wood
CENTERIOR ENERGY
Shared Package
ML20140D153 List:
References
EA-97-097, EA-97-97, NUDOCS 9704230014
Download: ML20140D147 (2)
v • d • e


Text

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r

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l

April 17, 1997 l

FA 97-097 ,

Mr. John K. Wood  ;

'

Vice President - Nuclear Davis-Besse Nuclear Power Station Centerior Service Company  ;

'

5501 North State - Route 2 Oak Harbor, OH 43449 SUBJECT: NOTICE OF VIOLATION (NRC INSPECTION REPORT 50-346/97002(DRS)) .

I Doar Mr. Wood: l

)

This will acknowledge receipt of your April 7,1997 letter in response to ou:

March 6,1997 letter transmitting a Notice of Violation associated with the above  !

mentioned inspection report. This report summarized the results of the maintenance rule i l

inspection at your Davis-Besse Plant. We have reviewed your corrective actions and have i

no further questions at this time. These corrective actions will be examined during future ,

!

inspections, i

!

Sincerely,

!

/s/ M. Icach (for)

!

e firey E. Grant, Director  !

. M 0 0 5 '? Division of Reactor Safety Docket No. 50-346 Enclosure: Ltr 04/07/97, J. K. Wood, .

'

Centerior Energy, to US NRC w/enci See Attached Distribution

)\

'

DOCUMENT NAME: G:\DRS\DAV041_7.DRS To receive e copy of thie document, indicate in the box 'C' = Copy w/o attach /enci "E" = Copy w/ attach /enci "N* = No copy l OFFICE Rill:DRS , , , g Rlll:DRP6\ % R ll:EICS lg Rlli:DRS lv NAME Gavula/kjKLP( Jacobsqh\,h\,N Yayton C9g/ff8 Leach / Grant &4

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[ DATE 04/ /7 /97( ' 04/p/91 )\ 04/f 7/97 /g 04//7/97

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OFFICIAL RECORD COPY /

I 9704230014 970417 r j- gDR ADOCK 05000346 PDR .

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_ _

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J. Wood 2 April 17,- 1997 cc w/o encl: J. P. Stetz, Senior Vice President - Nuclear J. H. Lash, Plant Manager J. L. Freels, Manager, Regulatory Affairs

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cc w/ encl: State Liaison Officer, State of Ohio Robert E. Owen, Ohio

-

Department of Health  !

C. A. Glazer, State of Ohio Public Utilities Commission Distribution:  ;

Rlll PRR w/enci Rlli Enf. Coordinator w/ encl Docket File w/, encl _7 SRI, Davis-Besse w/ encl g_PUBLIC IE-01;w/ encl TSS w/enci -

OC/LFDCB w/encI LPM, NRR w/enci J. Lieberman, OE w/enci '

DRP w/ encl A. B. Beach, Rlll w/enct J. Goldberg, OGC w/ encl DRS w/enci C. D. Pederson, Rif t w/ encl R. Zimmerman, NRR w/enci

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