IR 05000346/1993016
| ML20059B839 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 12/22/1993 |
| From: | Burgess S, Wright G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20059B818 | List: |
| References | |
| 50-346-93-16, NUDOCS 9401040264 | |
| Download: ML20059B839 (19) | |
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U. S. NUCLEAR REGULATORY COMMISSION
REGION III
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Report No. 50-346/93016(DRS)
Docket No. 50-346 License No. NPF-3 Licensee: Centerior Service Company c/o Toledo Edison Company 300 Madison Avenue Toledo, OH 43652 Facility Name: Davis-Besse Nuclear Power Station Inspection At: Oak Harbor, OH 43652 Inspection Conducted: September 27 through November 16, 1993 Inspection Team: S. Burgess, Team Leader A. Dunlop J. Hansen K. Riemer NRC Consultants: M. Shlyamberg, Parameter, In Approved By: MD BurgessFb b 6' Team Leader 12- 20 .53 Date Approved By: / M d~ b n/t VO G. C. Wright, Chief Date Engineering Branch Inspection Summary:
Inspection on September 27 throuah November 16. 1993 (Report N /93016(DRS))
Service water system operational performance inspection (SWSOPI) in accordance with NRC Temporary Instruction 2515/11 Results:
The team considered Davis-Besse's service water system design and operation to be effective. Engineering and technical support were adequate based on their involvement in the SWS design and operation and the quality of recently performed calculations. System design and engineering support strengths and weaknesses are provided in the Executive Summary of this report. One violation was identified regarding failure to perform a safety evaluation as required by the 10 CFR 50.59 (Section 5.1). Two unresolved items.were identified related to the licensee's position on single failure (Section 5.2)
and the ultimate heat sink seismic design (Section 5.3). One inspection foliowup item _ was identified regarding testing recommended by the licensee's System Test and Review Program (Section 8.1.1).
9401040264 931229 PDR ADOCK 05000346 '
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P TABLE OF CONTENTS I
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EXECUTIVE SUMMARY.................................................... .i l INSPECTION SCOPE AND OBJECTIVES................................ 2 ,
l LICENSEE ACTION ON PREVIOUS INSPECTION FINDINGS................ 2 .
- GENERIC LETTER 89-13 IMPLEMENTATION............................ 3 !
r SYSTEM DESCRIPTION............................................. 3 I
- MECHANICAL DESIGN REVIEW....................................... 4 PERATIONS...................................-.................. 7 MAINTENANCE............................................. ....... 8 l SURVEILLANCE'AND TESTING...................................... 11 QUALITY VERIFICATION AND CORRECTIVE ACTIONS. . . . . . . . . . . . . . . . . . . 13
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10.0 INSPECTION FOLLOWUP ITEMS..................................... 13 ;
l l 11.0 UNRESOLVED ITEMS........................ ..................... 13 l ;
12.0 EXIT MEETING.................. ......... ..................... 13 J i
Appendix A - Personnel Contacted
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l Appendix B - Generic Leiter 89-13 Action Items 1
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i Executive Summary During the period September 27 through November 16, 1993, a Region III inspection team conducted a service water system operational performance <
inspection (SWSOPI) at the Davis-Besse Nuclear Power Station. The inspection scope encompassed the service water (SW) and component cooling water (CCW)
systems. For these systems, the inspection included a mechanical design ,
review; detailed system walkdowns; review of system operation, maintenance, and surveillance; and assessment of quality verification and corrective actions required by Generic Letter 89-13, " Service Water System Problems Affecting Safety Related Equipment," as well as system unavailabilit The team considered Davis-Besse's SWS design and operation capable of i completing the safety function required by the design basis. In addition, the team concluded that, overall, the engineering and technical support organizations were qualified and involved in SWS design and operatio However, because of an inordinate number of inaccurate piping and instrumentation drawings (P& ids), operational schematics (OSs) and operational procedures, the team did not consider the systems' configuration control adequate. lhis could present unnecessary challenges to proper plant operation ,
and modification. Configuration control issues are documented in NRC report 50-346/93019. The team also identified the following weaknesses:
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the lack of a safety evaluation for the change in the valve alignment *
for the standby containment air cooler,
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failure to implement your own preventive maintenance actions in response to relay conditions identified in an information notice, and
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numerous errors in SW and CCW operating procedures and training lesson plan i The team identified the following strengths:
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SWS material condition appeared to be good,
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heat exchanger maintenance and performance monitoring were effective, i
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recently performed analyses and calculations were of good qualit !
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DETAILS l Inspection Scone and Obiectives Nunerous problems identified at various operating plants in the country have called into question the service water systems' (SWSs) ability to perform their design function. These problems have included: inadequate heat removal capability, biofouling, silting, single failure concerns, erosion, corrosion, ;
i insufficient original design margin, lapses in configuration control or
) improper 10 CFR 50.59 safety evaluations, and inadequate testing. NRC management concluded that an in-depth examination of SWSs was warranted based ,
on the identified deficiencie i The inspection team focused on the mechanical design, operational control, maintenance, and surveillance of the SWS and evaluated aspects of the quality assurance and corrective action programs related to the SWS. The inspection's-primary objectives were to:
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assess SWS performance through an in-depth review of mechanical systems functional design and thermal-hydraulic performance; operating, maintenance, and surveillance procedures and their implementation; and operator training on the SWS,
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verify that the SWS's functional designs and operational controls are capable of meeting the thermal and hydraulic performance requirements and that SWS components are operated in a manner consistent with their design bases,-
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assess the licensee's planned and completed actions in response to !
Generic Letter 89-13, " Service Water System Problems Affecting Safety Related Equipment," July 1989, and
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assess SWS unavailability resulting from planned maintenance, surveillance, and component failure The areas reviewed and the concerns identified are described in Sections through 8.0 of this report. Conclusions are provided after each sectio Personnel contacted and those who attended the exit meeting on November 16, 1993, are identified in Appendix A. Details pertaining to Generic Letter (GL) 89-13 action items are attached as Appendix .0 Licensee Action on Previous inspection Findinas (0neni Inspection Followup Item 50-346/93017-02(DRP): During an inservice test performed on September 13, 1993, valve CC-1495 failed to fully close and engage a detent that assured the valve maintained in its safe position (closed) following a design basis accident (DBA). The test was performed wit full flow conditions. The valve was previously tested under partial flow conditions with bypass valve CCW-43 open during the tes Valve CC-149S isolates non-essential component cooling water (CCW) loads from the safety related portion of the CCW system during accident conditions to-
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assure that an adequate supply of CCW is provided for the safety related component Failure of this valve to close during a DBA would lead to an unanalyzed condition, should the flow rates to the safety related components decrease below their analyzed value The operability determination of this failure, performed under potential condition adverse to quality report (PCAQR) 93-0429, concluded that the valve would have performed its safety function during a DBA event. The justification for this conclusion was based on the CCW system flow distribution analysis. During the test the valve went to 95% closed. lhe engineering analysis determined that safety related components would receive required flows up to the limiting valve closure of 78%. However, the operability determination failed to document the consequences of the loss of instrument air to the operator, which could cause the valve to re-ope Modification 88-0066 specifically added a detent to the air operator to assure that the valve would remain closed; therefore, the operability determination was inadequat The licensee stated that the loss of instrument air was considered, but not documented, since the potential for re-opening did not exist. The licensee's position regarding the valve re-opening was based on an EPRI A0V documen The licensee also stated that modification 88-0066 was initiated to prevent re-closing of similar A0Vs. In response to the team's concern, the licensee committed to reperform the operability determination and to document the loss of air evaluation. This item remains ope .0 Generic Letter 89-13 Implementation The NRC issued GL 89-13, " Service Water System Problems Affecting Safety Related Equipment," requesting that licensees take certain actions related to their SWS. These actions included establishing the appropriate frequencies for testing and inspecting safety related heat exchangers over three operating cycles, to ensure the operability of SWSs that are credited for cooling safety related equipmen The team considered Davis-Besse's initial response adequate in addressing GL 89-13 concerns. See Appendix B for details pertaining to each GL 89-13 Action Ite .0 System Description The safety related service water was comprised of the SW and CCW system The SW system passes cooling water to the following major safety related components:
. CCW heat exchangers' tube side that provided heat removal from the CCW system,
. containment air coolers (CAC) that provided containment heat removal, and
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emergency core cooling system (ECCS) room air coolers that provided heat removal from the equipment housed in the ECCS room '
The SWS is a once-through system that supplied cooling water to the heat exchangers listed above as well as provide an emergency source of supply and makeup for the auxiliary feedwater and CCW systems respectively. The SWS also provided cooling for the non-safety related loads. The non-safety related loads were automatically isolated during accident conditions. The system consisted of two independent trains with one train needed to provide all ,
safety related services. There were three SW pumps, each rated for 10,250 gpm ;
at 69 psid. Only one pump per train was required; the third pump can be manually aligned to either train. Similarly there were three CCW heat j
exchangers and three CAC heat exchanger l The CCW system recirculated cooling water to the following major safety i related components:
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. CCW heat exchangers' shell side that provided the heat transfer to the ..
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decay heat removal (DHR) heat exchanger that provided the decay heat !
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diesel generator (DG) jacket water heat exchanger that provided the diesel jacket water heat removal, and
. DHR and make-up pumps that provided lubrication heat remova CCW is a closed loop system that provided cooling water to the heat exchangers i listed above as well as non-safety related loads. The non-safety related I loads were automatically isolated during the accident conditions. The system ;
consisted of two independent loops with one loop needed to provide all safety related services. There were three CCW pumps, each rated for 7860 gpm at 65 psid. Only one pump per loop was required; the third pump can be manually
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aligned to either loo .0 Mechanical Desian Review
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The mechanical design review of Davis-Besse's SW and CCW systems included determination of whether the systems' design bases, design assumptions, 3 calculations, analyses, boundary conditions, and models met licensing !
commitments and regulatory requirements. This review also included an ;
assessment of a single failure impact on the ability to perform required '
safety function. Also reviewed were the SW and CCW systems' capability to meet the thermal and hydraulic performance specifications during accident or abnormal conditions. The team also reviewed the systems' seismic qualification, design vulnerabilities, flooding mitigation characteristics, and selected modification package ;
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1 Chanae in the Alianment of CAC isolation Valves The CAC valve alignment was changed during the eighth refueling outage (8RO) i for the standby CAC unit. The supply motor operated valve (MOV), SW-1368, position was changed from normally closed (NC) to normally open (NO), whereas ,
discharge air operated valve (A0V) SW-1358 was changed from N0 to NC and also !
the position of manual isolation valve SW-325 downstream of the ADV was {
changed from N0 to NC. This alignment change resolved a water hammer :
condition experienced during 8R i This change in alignment met the definition of a change in the facility as ,
described in the safety analysis report provided in paragraph 4.2.4 of !
procedure NG-EN-00304, " Safety Review and Evaluation," Revision This !'
paragraph defined a change in the facility as a modification to the updated safety analysis report (USAR) text or drawing that was anything other than an i editorial correction or clarification. The CAC alignment was shown in USAR Figure 9.2-1, Revision 1. Also, this change in alignment resulted in a !
trapped volume between the A0V and the manual isolation valves. Additionally, this alignment change constituted a change in manner and location of containment isolation of the standby CAC unit. Prior to this alignment change, the standby CAC unit had supply containment isolation capability, i which was negated by this chang l The original safety review (SR) for the alignment change, dated March 1,1993, was identified as an operational procedure change and failed to identify that l this change constituted a change in the facility as described in the safety '
analysis report. Also, the SR failed to identify the trapped volume between l the discharge isolation valves. Furthermore, the original SR did not provide any description of valve position changes; the only reference to actual changes were limited to identification of the applicable Technical Specification and USAR section number ;
The subsequent SR, dated September 2,1993, identified the trapped air volume i presence, but similarly failed to identify that this change constituted a change in the facility as described in the safety analysis report and did not address the change in containment isolation functio The team considered the failure to perform a safety evaluation for a change in i the facility as described in the safety analysis report to be a violation of 10 CFR 50.59 (50-346/93016-01). The licensee subsequently performed a safety ;
analysis per 10 CFR 50.59, which concluded the CAC valve alignment did not constitute an unreviewed safety question or require a Technical Specification i chang The handling of this issue also exemplified the team's concern related to l adequacy of configuration control, since the position of these valves was i different on the operation schematics (05), piping and instrumentation drawings (P& ids), USAR Figure 9.2-1, Revision I and operations procedure DB-0P-06261, Revision 00. Only the operating procedure was revised to reflect ;
the new alignment. The team's concerns regarding configuration control are i documented in NRC report 50-346/9301 i
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'l Potential Non-Compliance with Sinole Failure Criteria Recuirements
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-Section 9.2.1.2 of Davis-Besse's USAR provided details of the SW system's :
automated capability to switch the discharge to the seismically qualified portion of the intake structure if the seismic event caused the blockage of service water as a result of a non-seismic line failure. This action was :
initiated by means of pressure switches PSH 2929 and PSH 2930. These devices :
enabled opening of the normally closed discharge isolation valves SW-2929 and
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SW-2930 if the pressure sensed by these switches exceeded the setpoint value ;
of 50 osig. The two independent PSHs controlling two independent valves (each .
switch and its corresponding valve was powered by a different Class lE power i source) constituted two trains and, hence, provided the single failure protection, should the normal discharge through the non-seismically qualified j piping become unavailable. However, both pressure switches could be isolated by a single common root valve, SW-2930A. A potential failure of this common root valve, whether inadvertently left closed, or mechanical in nature, would disable both train '
The licensee did not consider the misalignment of valve SW-2930A to meet the i single failure criteria compliance; however, prior to completion of the '
inspection the valve was locked open. Pending review of this issue by NRR,-
this is considered an unresolved item (50-346/93016-02). . Seismic Desian of the Ultimate Heat Sink -
The containment and emergency core cooling transient analysis was performed i utilizing the non-seismic portion of the ultimate heat sink (UHS). This ;
analysis was performed assuming that the SW temperature remained constant for :
the duration of the analyzed events. The SW temperature can remain constant ;
only if the interconnections between Lake Erie and the seismic class I intake ,
structure were intact. These interconnections were not classified as seismic j class 1. According to the information presented in USAR Section 9.2.1.2.a, ;
the supply SW temperature following a DBA loss of coolant accident (LOCA) ;
would increase from 76 F to 131of in 14.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> if only the seismic class l portion of the UHS was availabl ,
The use of non-seismic class I interconnections appeared to be in conflict >
with the following commitments stated in the USAR:
- Section 9.2.1.1 states "The portion of the system required for emergency operation, including the intake structure, is designed to the ASME Code, l Section III, Nuclear Class 3 and Seismic Class I, as applicable."
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I 1 Section 9.2.1.2 states "If a LOCA and seismic event should ~ occur when I the forebay level is above elevation 564 feet, the actions of the i operator to ensure the return of the service water to the forebay are dependant on the consequences of the seismic event." r Appendix 30, item 3D.2.27 does not explicitly limit the definition of :
the UHS safety function to the "(1) dissipation of residual heat after 1 reactor shutdown". The safety function definition provided in Safety l l
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Guide 27 in addition to (1) above-included "and (2) dissipation of :
residual heat after an accident". l The UHS design review provided in SER, NUREG - Ol3C, Section 9.3.3, did not ;
discuss the ability of UHS to dissipate the residual heat after an acciden '
Pending review of this issue by NRR, this is considered an unresolved item (50-346/93016-03).
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' Nodification Packaaes i
The team determined that modifications performed on the SW and CCW systems !
were thorough and contained appropriate safety analyses. However, the 7 modification process allowed the closure of modification packages without the j appropriate changes to drawings and procedures. This was a contributing cause to some of the configuration problems identified by the team. Additional i discussion related to the configuration control issues is provided in NRC i report 50-346/9301 I Enoineerina Calculations l Recently performed SW and CCW calculations were of good quality and were I technically adequate. However, original calculations were not revised or i annotated when new calculations had been performed. This presented a ;
challenge in determining which calculations reflected the current design j basis. The licensee stated that consideration would be given to cross !
referencing the calculation :
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In general, the design of SW and CCW systems exhibited an appreciable margin i which allowed these systems to overcome some degradation and still perform !
their intended function in accordance with the licensing basis for the plan I With the exception of the unresolved issues, the team concluded that the SW and the CCW systems' mechanical design was effective and each system was ';
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capable of performing its safety related functions under the DBA scenari i Operations !
l The team reviewed plant operations to assess operator knowledge and the 1 accuracy and completeness of procedures and training with regard to the SW and CCW systems. The team performed system walkdowns; reviewed procedures for normal, abnormal, and emergency conditions; assessed conduct of operations in the field and control room; and evaluated training manuals, lesson plans, and operator actions on simulated SWS/CCW malfunction The team determined the licensee was operating the SWS in an appropriate manner. Although weaknesses were identified, operating procedures were considered adequate and plant equipment labeling was good. The operator training program relating to the SW and CCW systems was considered a strengt I
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Operator knowledge of SW and CCW equipment operations and procedures was
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Discrepancies in the P&lDs, OSs, and operational procedures were considered :
configuration control weaknesses. These concerns are discussed in NRC report '
50-346/9301 ;
! Coerations Sgenarios/Trainino i
Operator responses to simulated SW/CCW system related alarms and malfunctions !
were effective in mitigating the events. The operators were knowledgeable about necessary local actions and where those actions would be performe Equipment operators were knowledgeable of assigned procedural steps and understood SWS functio !
Training materials were adequate with SW scenario and job performance measures' '
(JPM) development a strength; however, several deficiencies were noted in the ;
lesson plans. Discrepancies in the training lesson plans indicated that when i the effort to match the P& ids and operations checklists lapsed, upkeep of '
training documents suffered. The following were examples of deficiencies identified in the lesson plans and system descriptions: ,
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SW system description SD-018 had not been updated concerning the spare CAC valve lineup resulting from a water hammer even !
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TCCS room cooler bypass MOVs SW-89, SW-97, SW-105, SW-ll3, and SW-121 ;
were incorrectly identified as locked open in the system descriptio ,
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SWS lesson plan, OPS-SYS-I305.00, and SWS student handouts incorrectly ;
indicated that valves SW-14, SW-15, SW-36, and SW-132 were locked ope i
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The CCW/SW cross connect tell tale drains SW-235 and SW-237 were incorrectly listed open in CCW system description 50-016 and SWS lesson pla l
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SWS lesson plan incorrectly listed the wrong pressures for some pumps and strainer DP The licensee agreed to correct the document .0 Maintenance I
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The tsam reviewed maintenance procedures, work history, completed work request packages, LERs, Potential Condition Adverse to Quality (PCAQ) reports, preventive maintenance (PM) tasks for selected components, performed detaile systen walkdowns and observed selected maintenance activities to determine if the SV and CCW components and piping were adequately maintained and system equipnent that required frequent maintenance was detected. The team also ,
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evaluated implementation of GL 89-13 commitments in the maintenance are '
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. SWS Preventive Maintenance The team determined that the PM program, as described in the licensee's initial response to address GL 89-13 concerns was adequate. The team noted, however, that seven leaks and two instances of silting have occurred in the SWS since 1990. Five leaks were located in the same 120 foot section of idle SWS piping. The leaks were caused by corrosion, most likely microbiological 1y influenced corrosion, and were not discovered by ultrasonic testing (UT). The licensee subsequently replaced the entire run of affected pipin Modifications implemented by the licensee in the SWS included the installation of a spoolpiece in this run of piping and the generation of a PM to visually inspect the piping internals. Also during the seventh refueling outage (7RO),
the licensee discovered silting in the SWS suction supply line to auxiliary feed pump 2 when the piping was cut to install an inspection related modification. As a result, a PM event was generated to visually inspect the line. No problems have been identified with this line subsequent to 7RO. The team considered the enhancements made to the existing PM program to be an ,
effective means of meeting GL 89-13 requirement .
The team requested copies of all SWS PM deferral forms from 1991 through 199 included in the request were those SWS PMs deferred during 8RO. The team
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concluded that justification for the deferrals were technically adequate, in ,
accordance with licensee procedures, and eventually performed within the 25%
grace perio However, one exception was noted where CCW-1 heat exchanger's .
PM was deferred several times and later failed to meet testing acceptance !
criteria as discussed in Section 7.2. The licensee documented the failure circumstances in PCAQR 91-0335, and the team concluded that the deferral appeared to be performed in accordance with station procedures and this occurrence was an isolated incident. Based on current PM program and associated procedure reviews, the team concluded that the PM program problems of the mid to late 1980's didn't exist today. Adequate administrative controls and management involvement were in place to accomplish PMs in a timely manner. Even if PM program problems existed then, the team determined I that no safety significant concerns existed toda .2 Heat Exchanaer Inspection and Cleanina i
Based on the team's observation of CCW-2 heat exchanger conditions, as well as the historical review of inspection results, the team concluded that the licensee's program for meeting GL 89-13 requirements was effectively performe ;
The team noted that the initial implementation of heat exchanger inspection i and cleaning to satisfy GL 89-13 commitments was adequate with improvements i added based on inspection results. However, due to a leaking valve that ;
prevented isolation of CCW-1 heat exchanger, the inspection PM for CCW heat i exchanger was deferred several times. During the refuel outage CCW-1 heat -
exchanger failed the TS performance test, and inspection revealed an excessive amount of fouling in the heat exchanger's service water side. As a result, l the licensee changed the CCW heat exchanger PMs to require a. tube cleaning, on l
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a staggered basis, every 18 months. The team observed the PM performance for !
CCW-2 heat exchanger and noted no significant silting or fouling. In addition to reviewing the photo history of prior CCW heat exchanger inspection results, the team also viewed a selective. videotaped film history of prior SW5 piping inspections. No concerns were note !
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. Acceptance Criteria The team identified the lack of acceptance criteria for inspection PMs for various heat exchangers and other components. This weakness was also noted by ,
the licensee in their SWS review. However, the licensee adequately documented ;
the inspection PH results;'in addition, the team determined that trending based on photo and videotape results would be effective in identifying the !
need for additional corrective action :
' SWS Pioe t.eak
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During the SWSOPI inspection, a one gallon per minute leak developed in the !
SWS pump 1 strainer discharge line. The licensee had not performed UT of this piping due to inaccessibility. When the section of piping was replaced, initial investigation results determined that the leak occurred at a weld and '
was not due to erosion. The leak was the result of a hole in the weld, most ,
likely due to weld porosity combined with localized pitting or particulate !
corrosio :
^ In response to the leak, the licensee attempted a new method for perforn.ing UT :
of this particular portion of piping. The licensee tested the similar location in the other two SW pumps and identified some pitting but all '
readings were within the nominal band for pipe wall thickness. This data was j identified as a new baseline and incorporated into the corrosion erosion monitoring and analysis program. The licensee was evaluating whether to generate a new PM, and the frequency, to UT this portion of piping. Initial determination by the licensee indicated that this data would probably be used ,
to identify the need to replace this piping on an "as required" basi The i team reviewed the work associated with the piping replacement and discussed preliminary testing results with the licensee and had no concern .5 SWS Valve Failure
The SW supply isolation valve to AFW pump 1-2, SW-1383, failed to open when !
required during the performance of a surveillance test performed on September 29, 1993. The licensee determined the most probable cause was the failure of a normally energized Agastat rela ,
The team reviewed the licensee's response to Information Notice (IN) 84-20, :
which discussed the service life of Agastat relays in safety related system The IN was received by the licensee; however, the plans were never implemented because of miscommunications between engineering and maintenanc Even though the Agastat relays discussed in IN 84-20 were a different series than the Agastat relay that failed in SW-1383, the team concluded that the licensee's failure to implement actions to address-IN 84-20 could have prevented the possibility of predicting and averting the failure of SW-138 The licensee planned to identify all time delay Agastat relays in safety applications or applications that could affect continued power operation This list would then be reviewed to acknowledge those relays that have adequate testing or those relays that would be expected to fail in their :
safety position. An appropriate relay life for non-EQ Agastat time delay i
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relays, for both energized and de-energized applications, would be evaluated based on information from Agastat relay failures at Davis-Besse and other sites, vendor information, and related equipment qualification data. The licensee also planned to initiate replacement PMs for applicable relays with an appropriate replacement frequency. The team considered these actions acceptabl .6 Conclusions Overall, the team determined that maintenance performed on SWS and CCW components was effectively accomplished and that system problems were properly addressed and corrected. The team also concluded that recent licensee efforts have greatly improved the SWS housekeeping and material condition. While the initial response to GL 89-13 was adequate, subsequent enhancements to the PM program have improved the licensee's GL progra .0 Surveillance and Testina The team reviewed preoperational test procedures, surveillance procedures, and the licensee's inservice test (IST) program and implementing procedures to determine if sufficier.c testing had been conducted to confirm system design requirements and system operabilit .1 Inservice Testina of Pumps and Valves 10 CFR 55.55a required the SW and CCW systems to be in compliance with the American Soc;ety of Mechanical Engineers (ASME) inservice pump and valve testing specified in Section XI of the Boiler and Pressure Vessel Code, except where relief had been granted by the NRC, or where alternative testing was justified in accordance with Generic Letter 89-04. The team reviewed the implementing TS surveillance procedures and identified the following issue . Procram Scope The SWS IST program scope was adequate with the following exceptions: .
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The IST program did not contain a closed safety function for the SW pump strainer blowdown valves; even though, flow could be reduced to safety i related components if the valves failed open. Although this function was not in the program, the required Code testing was performed and 4 trended. The licensee committed to add the closed safety function to the IST progra .
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SW to CCW cross-connect valves SW-232, SW-233, SW-234, and SW-236 were not included in the IST program even though SW provided the only seismic backup water supply to the CCW system. The licensee stated that no credit was taken in the licensing basis for the valves' operation during a seismic event. The valves were available for operator convenience in the event the other two makeup water supplies were not available. Since makeup to CCW was normally less than 100 gallons per month, it should not be required during a seismic event. The licensee, however, committed to include exercising these valves in the augmented program
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every refueling outage to ensure valve reliability. Based on this information and discussion with NRR, not including the valves in the IST program was determined to be acceptabl i The licensee's System Test and Review Pro" ram (SRTP) for the SWS included a recommendation that these valv;s be tested every refueling outage. The valves were added to _the refueling surveillance procedure; however, when the procedure was revised, the valve testing was delete Testing required to be performed prior to the plant restart from the 1985 AFW event was not placed in a commitment tracking system unless it pertained to a TS or required an LER. As such, testing recommended from the SRTP could be changed or deleted without understanding why the testing was being performed. This will be considered an inspection followup item pending a review to determine if other testing recommended i by SRTP was also inappropriately deleted (346/93016-04). i
8. Missed Testina l The team identified the following example of missed inservice testing-
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Stop check valves CCW-148, CCW-149, CCW-150, and CCW-151 were not adequately exercised on a quarterly basis to the full open position to fulfill their safety function as stated in the IST program. The valves were added to the IST program in June 1992 when this portion of the system was reclassified. The valves were required to open to supply CCW to the decay heat pump bearing motor housing cooler. The quarterly tes verified that when the decay heat pump was operating the pump bearing !
temperature would not exceed its limiting value of 170 f. This testing l methodology did not meet the intent of GL 89-04, position 1 or 2, to verify the valve opened by passing full flow or by ,
disassembly / inspectio The test, however, could be considered a partial stroke tes The licensee committed to continue the quarterly partial stroke test, l
perform a disassembly / inspection during the next refueling outage, and l
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document these changes in the IST program. If the disassembly and inspection was unfeasible due to the welded bonnet, modification 86-251 would replace the stop check valves with manual valves. Initial plant design changes no longer required the valve to close on reverse flow and I was now only considered a maintenance isolation valve. The team considered these actions acceptabl . Pumo Testina Pump testing was performed in accordance with the ASME Code requirements with the following exceptions: 1
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When pump testing was performed, recorded data from the test was averaged over the 30 minute test period and then compared with the reference value. This did not appear to meet the Code's intent and was identified during a previous IST inspection. The licensee agreed to I submit a Code inquiry to resolve the concern and to request relief from the NRC in the interim.
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The acceptance criteria for the SW pumps was based on pump curves that meet the Code limits; however, the design requirements for the pumps may be more limiting. This issue was identified by the internal licensee SW safety system functional review (SSFR) and was resolved by changing the !
IST administrative procedure to ensure when pump curves were developed ;
the pump design and Code requirements were me ' Conclusions i
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The surveillance and testing programs were adequate to verify the SW and CCW systems operability. The deletion of SRTP recommended testing could be a concern if determined to be a pervas-ive issu .0 Ouality Verification and Corrective Actions t The team evaluated SW and CCW systems, and GL 89-13 program implementation assessments, technical audits and audits pertaining to recent configuration ;
control issues, reviewed the corrective action tracking system to ensure adequate treatment of related items, and reviewed SWS and CCWS operational history to assess the adequacy of root cause evaluations, i
While the quality assurance (QA) organization conducted effective performance '
based audits, resultant corrective actions have not resolved all design configuration and plant equipment control problems. These concerns are documented in NRC report 50-346/9301 i 1 Inspection Followup Items j
Inspection followup items are matters that have been discussed with the :
l licensee, which will be reviewed further by the inspector, and which involve i some action on the part of the NRC or licensee or both. An inspection j followup item disclosed during the inspection are discussed in Section 8. ]
1 Unresolved Items i An unresolved item is a matter requiring more information in order to ascertain whether it is an acceptable item, a violation,-or a deviatio Unresolved items are identified in Sections 5.2 and .0 Exit Interview The team conducted an exit meeting on November 16, 1993, at the Davis-Bess Nuclear Power Station to discuss the major areas reviewed during the inspection, the strengths and weaknesses observed, and the inspection result Licensee representatives'and NRC personnel in attendance at this exit' meeting are documented in Appendix A. The team also discussed the inspection report's likely informational content with regard to documents reviewed by the team during the inspection. The licensee did not identify any documents or processes as proprietar !
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APPENDIX A I l
Davis-Besse Nuclear Power Station ;
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L.' Storz, Vice Presidant Nuclear J. Barron, Supervisor, Test / Performance !
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S. Byrne, Manager, Plant Operations j R. Donnellon, Manager, Plant Engineering G. Gibbs, Director, Engineering .
l C. Hawley, Manager, Quality Control J. Holden,- Manager, Design Engineering l j
S. Jain, Director, Nuclear Services l J. Lee, System Engineer !
N. Peterson, Licensing l
K. Prasad, Engineer, Nuclear Engineering
- A. Rabe, Supervisor,' Quality Engineering l i
J. Rogers, Manager, Maintenance G. Shukla, Nuclear Licensing P. Smith, Supervisor, Compliance E. Wilds, Lead Engineer, Nuclear Engineering J. Wood, Plant Manager, Davis-Besse l'
D. Wuokko, Supervisor, Nuclear Regulatory Affairs ;
A. Zarkesh, Supervisor, Safety Analysis l
U.S. Nuclear Reaulatory Commission G. Grant, Director, Division of Reactor Safety [
T S. Burgess, Team Leader
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J.-Shine, Resident Inspector, Davis-Besse l l
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APPENDIX B l
Generic letter 89-13 Action Items ! I Biofoulina Control and Surveillance Technioues l
Action I of GL 89-13 requested that licensees implement and maintain an I ongoing program of surveillance and control techniques to significantly reduce the incidence of flow blockage problems as a result of biofouling. The action !
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request included intake structure inspections, chemical treatment of service ,
water systems, and periodic service water system flushing / flow testin ,
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The team reviewed the actions taken by the licensee to address the generic d
letter request. Intake structure and service water bay inspections were conducted in accordance with PMs 2694, 4595, 4596, and 4597. The PMs inspected for sediment build-up, debris, macrofouling, zebra mussels, and damage to the intake structure. The PMs were weak.in providing guidance on
intake structure and SW bay inspections for algae build-up, SW pump column '
inspections, and inspection documentation requirements. The licensee stated ,
the inspection guidance would be reviewed and indicated that the SW pump years. inspections were also performed when the SW pumps were rebuilt every 5 column
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No concerns were identified during the last pump inspectio ,
i A continuous chlorination program was in place to arrest the growth of algae, slime, and zebra mussels. This program was effective in minimizing zebra mussels in the service water bays. Four injection points were installed in i i
the intake structure; three between the trash racks and traveling screens and one downstream of the dilution (backup SW) pump in bay 3. This last injection l
point was normally used when SW pump 3 was in service to reduce the amount of ;
chlorination returned to Lake Erie by the dilution pump. Injection would i
automatically start in the bay where the SW pump was operating. The first intake canal inspection in 1991 identified significant zebra mussel
accrulation on the trash racks. The zebra mussels were mechanically removed -
and the 1993 inspection identified only a small accumulation of zebra mussel ;
The licensee's actions for biocide injection appeared responsive to the generic letter reques i t
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The licensee stated in their initial response to GL 89-13 there were no idle I or stagnant loops in the SW system since it was in continuous operation. The supplemental response to the GL, dated September 9,1993, identified three l infrequently used lines: supply to auxiliary feedwater pumps, hydrogen ;
dilution blowers, and the control room emergency ventilation system. These lines were inspected or flushed on a refueling outage basis. The team *
considered additional lines that should have been considered idle loop i These included the idle CAC, and the CCW heat exchangers not in operatio .)
The CACs were rotated on a monthly basis, while one CCW heat exchanger was inspected every six months as a result of past silt buildup problems. These i actions appeared to minimize silt accumulation in the SW system in idle or stagnant line i
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I Monitorina Safety Related Heat Exchanaer Performance r Actier.11 of GL 89-13 requested that licensees irplement a program to !
periodically verify the heat transfer capability of safety related heat !
e.xchargers cooled by the SWS. The test program should consist of an initial i test rogram and a periodic retest progra i
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APPENDIX B 2
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' Davis Besse's test program consisted in some cases of both inspecting / cleaning i and performance testing of safety related heat exchangers. Performance ;
testing was conducted on the CCW heat exchangers, CACs, ECCS room coolers, and i the decay heat and low pressure injection system decay heat coole Inspection / cleaning was conducted on the CCW heat exchangers, ECCS room !
i coolers, and the control room emergency condensing units. In addition, monitoring tests were conducted for the CACs and ECCS room cooler !
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The performance portion of this program provided very good tools for the heat exchanger performance monitoring and compensatory maintenance activity;
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however, more explicit acceptance criteria needed to be developed. This ;
condition was further exacerbated by a lack of definitive values for the :
minimum required design flows. The licensee was aware of this condition and was in a process of developing more definitive acceptance criteria for heat ,
exchanger performance and system flow testin i f
The team concluded that the licensee's response to Action II was adequate and the GL intent was me i 11 h Routine Inspection and Maintenance !
i Action III of GL 89-13 requested that licensee implement a routine inspection and maintenance program for open-cycle SWS piping and components. This program should ensure that corrosion, erosion, protective coating failure, :
i silting, and biofouling cannot degrade the performance of the_ safety related systems supplied by the SW .'
t In their initial response to GL 89-13, the licensee utilized the existing PM 1 program for routine SW inspections and maintenanc i The licensee reviewed the PMs to determine initiated to inspectifthe enhancements intake bays, could be made and three additional PMs were in a supplemental response to Gt 89-13,
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i dated September 9,1993, the licensee implemented several SWS modifications that provided additional inspection locations. These modifications, i
implemented during 7R0 (August - November,1991) and 8R0 (March - April, j 1993), installed removable spoolpieces in the SWS. The licensee performed
visual inspections of the previously inaccessible piping and videotaped the '
results to obtain a baseline for inspection PMs of the piping intervals. The supplemental response also stated that two additional access points were ,
scheduled to be installed during 9R0 (October,1994). Additional baseline t
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data would be collected and additional inspection PMs would be generate The team considered the licensee's initial response to GL 89-13 Action item I
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Ill to be adequate. Changes to the program including installation of l inspection modifications, the generation of new inspection PMs, and i
videotaping of inspection data, should improve the program's effectivenes i I Desian Furction Verification and Sinole Failure Analysis Action IV of GL 89-13 requested that licensees confirm that the SW and CCW i i
systems basis forwill theperform plant. their intended function in accordance with the licensing This confirmation should include a review of the ability to perform failur required safety functions in the event of a single active to.nonent
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The team reviewed the SW and CCW systems' ability to perform their intended i
function in accordance with the licensing basis for the plant. The review !
included system configuration, flood and tornado protection, seismic design, i emergency power supply, and functional logic evaluation. Specific '
consideration was given to identify failure of any single component that could potentially affect the performance of safety related systems served by either (
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The team concurred that the SW and CCW systems would perform their functions
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i in accordance with the licensing basis for the plant. However, the team j identified a potential vulnerability related to the UHS design. Davis-Besse's j containment and emergency core cooling transient analyses were performed utilizing non-seismic portions of the UHS. This concern is discussed further ,
in Section i i
The team's assessment of the licensee's response related to single failure j analysis is pending the disposition of the unresolved issue listed in Section l Traininq Action V of GL 89-13 requested that licensees confirm that maintenance :
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practices, operating and emergency procedures, and training that involves the 'j
' SWS were adequate to ensure that safety related equipment cooled by the SWS' t will function as intended and that operators of this equipment will perform effectivel !
Licensee commitments to this item included the review of all procedures and .j training relating to the SW and CCW systems for accuracy and adequac !
Based on the team's review of maintenance practices, operating and emergency I
procedures, and training documentation, the team concluded that overall, !
Action V was adequately accomplished but that the response was weak due to the j number of errors identified in the operating procedures and training lesson plan :
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