IR 05000346/1988006
| ML20151T166 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 04/12/1988 |
| From: | Darrin Butler, Gardner R, Liu W, Westberg R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20151T151 | List: |
| References | |
| 50-346-88-06, 50-346-88-6, NUDOCS 8804280604 | |
| Download: ML20151T166 (29) | |
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s-w U.S. NUCLEAR REGULATORY COMMISSION
. REGION III Report No. 50-346/38006(ORS)
Docket'No. 50-346 License No. NPF-3 Licensee: Toledo Edison Company
Edison Plaza
.! 300 Madison Avenue l Toledo, OH 43652 Facility Name: Davis-Besse, Unit 1 Inspection At: Oak Harbor, Ohio Inspection Conducted: Februa 22-26 and March 7-11, 1988 G. W Inspectors: Rolf A. Westberg II/ 8N Team Leader Date Winston C. Liu Nlllf$8 Date
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David S. Butler //
Date k.bA 1 I Approved By: Ronald N. Gardner, Chief Plant Systems Section Date Inspecticn Summary
- Inspection on February 22-26 cnd March 7-11, 1988 (Report No. 50-346/88006(DRS))
Areas Inspected: Special safety inspection of activities with regard to review of allegations and resultant review of QA implementing procedures (51061B) and quality records (510658); design changes and modifications (37700, 37701, 37702);
i Licensee action on previously identified items (92701); and training (41400).
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Results: Of the four areas inspected, no violations or deviations were
- identified in two areas; two violations were identified in the remaining areas (failure to install oil sightglass in accordance with approved procedures (Paragraph 3.c.(4)(f)) and failure to apply design control measures to a specification change (Paragraph 3.c.(5)(b)).
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DETAILS t
1, Persons Contacted ,
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Toledo Edison Company (TED) t L. F. Storz, Plant Manager i P. C. Hildebrandt,-Engineering General Director !
L.~ 0. Ramseth, Quality Assurance Director S. C. Jain, Nuclear Engineering and Independent Safety Engineering Director a
D. S. Knaszak, Engineering Services Manager
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J. C. Sturdavant,-Licensing Principle ,
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G. Honma, Compliance Supervisor T. W. Anderson, Maintenance Planning and Outage Maintenance Superintendent ! L. Tillman, Design Process Supervisor L M. J. Knaszak, Design Engineer ,
U.S. NRC_ l P. M. Byron, Senior Resident Inspector R. N. Gardner, Chiof, Plant Systems Section
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1. J. Harrison, Chief, Engineering Branch t The preceding personnel attended the exit meeting at the Davis-Besse site on March 11, 1988. Other personnel contacted as a matter of routine during the inspection are documented in Attachment A to this repor '
- 2. Licensee Action on Previous Identified Iteme-
< i (Closed) Unresolved Item (346/86019-02): Review of clarification and !
j procede*al revisions to the Design Change Program. See Section 3.c.(2)
- of this report for detail . Facility Change Request (FCR) and Modification Review ,
' Inspection Scope f
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Th's three-week special safety team inspection of the FCR and j modification program reviewed the following areas: FCR closecut r commitment to NRC; design control program; audits; program
implementation relative to completed FCRs; and review of edifications scheduled for the fifth refueling outag In addition to document reviews and interviews, limited walkdcwns and verification of completed maintenance work orders (MW0s) and surveillance activities were accomplished. Team members also reviewed calculations in the civil / structural, mechanical, and electrical / instrument and control disciplines.
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b. Summary In general, the design change program was being effectively implemented; however, a number of program strengths and some j program weaknesses were identified. In addition, violations of NRC requirements were identified. Details of the violations are explained in the body of this repor The identified program strengths and weaknesses a.e summarized belo (1) Strengths l-t * 10 CFR 50.59 safety reviews were goo * FCR closecut was vastly improved since 198 * Good improvement was noted in the control of Measuring and Test Equipment (M&TE).
- Excellent instrument and control (I&C) instrument data string package * Implementation of the motor operated valve reliability program and MOVATS testing is a positive step in the resolution of MOV problem (2) Weaknesses
- Quality Assurance audits of Design were nontechnical, not activity oriented, and did not assess assurance QA activitie * Assurance QA was understaffed and overworked on nondesign related activitie * Too many procedures were required +.o implement the design proces * Examples were found which show inadequate design verification; however, the overall designs were acceptabl (3) Conclusions The licensee has been aggressively attacking the FCR backlog and should accomplish their goal. In addition, the design process appears to be improving with the advent of the new modification syste Finally, the introduction of more technical design audits should solve the design verification weaknes ,
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c. 064..iled Inspection Findings
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(1) FCR Closeout Commitment Review As a result of NRC inspection findings in 1985 (50-346/85031-02; '
FCR system is ineffective and 50-346/83035-05; MWO and FCR ,
systems require further evaluation and improvement), Toledo Edisoa Company (TED) commftted to the development of an action plan to reduce the FCR cloacut backlog. The subsequent action i plan identified 448 FCRs for u seout by the end of the fifth '
refueling outage (March 1988). Since the original commitment, i more FCRs were added to the listing and the total reached 54 During this inspection, of the 547 FCRs, only 85 remained ope Since this inspection preceded the outage, it appears certain ,
that the FCR closecut commitment will be me .
The responses to the 1985 NRC inspection findings also included commitments regarding procedure revisions, generation of a new procedure to control interdivisional'FCR activities, and an audit of the FCR system by a third party contracto The inspectors !
reviewed Procedure Nos. NEP-010, "Processing Facility Change ;
Requests;" NEI-010.2, "FCR Closeout Instructions;" NG-NE-0301, t
"Plant Modifications;" and the January 16, 1986 audit of the FCR system by Stone & Webste The inspectors also interviewed key members of the FCR closeout or0anization relative to FCR
- processing and prevention of recurrence of a backlog. The aforementioned reviews and interviews produced the following [
information:
l * Personnel directly responsible for FCR closeout have increased from six in 1985 to the present thirteen.
> * Design Process Management reports to the Engineering -
General Directar This is a higher level of management i than the reporting chain which existed in 1985.
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- Management personnel with FCR responsibility have generally risen in level, that is, coordinator to supervisor and supervisor to manager.
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! * The depth of definition in the procedures has increase ,
- Coordination between participating organizations has improved l
with the development of "Motherhood Procedure," NG-NE-030 * Training has improved. Where it was prsviously reading only, now classroom training is provide * Better procedural control of revisions and supplements to FCRs will help to prevent future backlog of FCR closeouts.
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? * -Increased personnel dedicateit to FCR closeout and the creation of a field closeout organization will also help
to prevent future backlog l
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The inspectors concluded that the FCR closeout progrant had improved greatly between 1985 and the present and that it was ,
being implemented effectivel (2) Design Control Program The inspec. tors reviewed the TED design control program to verify that it was_in conformance with the QA program commitments and regulatory requirements. This review assessed the program procedures against the requirements of 10 CFR 50, Appendix B, Criterion III, and the TED QA manual commitment to the gooc4 work practices recommended by ANSI N45.2.11 "Quality r Assurance Requirements for the Design of Nuclear Power Plants."
The review of the program procedures, which are listed in '
Attachment C to this report, was completed with acceptable results. The review produced one program weakness comment, .
that a large number of procedures were used to implement the progra This comment was discussed with the Engineering General Director who indicated that this fact had been recognized and was the subject of a third party assessmen The inspectors also reviewed the unresolved item from Inspection Report No. 50-346/86019-02 and concluded that the previous concerns had been adequately addresse (3) DesignAudia The inspectors reviewed the audit program relative to design control to verify programmatic commitments and technical merit, i The iteas considered during this review included independence ,
of audit personnel, personnel qualifications, schedule, corrective action, and technical conten Thirteen internal design audits (TED) and.five external dasign audits (Architect / Engineer organization) conducted in 1985 ,
through 1987 were reviewed. In addition, six audits conducted :
in 1987 relative to test control, training, and corrective action were also assessed. For a complete listing of these ,
audits, see Attachment D to this repor !
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This review indicated that TED was implementing their commitment to audit design control; however, only two of the eighteen 3 design audits reviewed were activity oriented or assessed i technical items. The two audits were No. AR-87-BECHT-01 and No. AR-87-BECHT-02,
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That the design audits were nontechnical, programmatic audits was identified as a program weakness and discussions were held with TED QA. As a result of these discussions, the inspectors-reviewed the qualifications of the personnel who purformed the design audits relative to their' previous design experience. Of eleven TED auditors and lead auditors, three had previous design experience. Of seven technical specialists who participated in the audits, two had verifiable previous d.tsign experience. The inspectors also discussed the activities Which were not verified by the TED audits, such as: calculations 10 CFR 50.59 evaluations, implementation of safety class boundarles at transitions, and post modification test result As a result of the above discussions with QA, discussions were also held with Design Engineering and Engineering Assurance QA relative to assessment of the technical content of the design packages. Subsequently, the inspectors reviewed a draft procedure, No. EN-DP-01203, "Engineering Design Evaluation,"
and two QA Procedures, No. QA-EA-01102.03, "Quality / Technical Reviews," and No. QA-EA-01105.03, "Review of Facility Change Requests and Plant Modifications." In addition, two recently completed design evaluations conducted under Procedure N EN-0P-01203 were reviewe The inspectors concluded that although the procedures and evaluations were acceptable, there was insufficient independence of the involved personnel to independently assess the technical content of the design package Further discussions and interviews were conducMd and the plant personnel indicated that the evaluations and reviews by Assurance QA were not being presented as ar. audit program but as a means of assessing and improving the design process. The inspectors agreed and concluded that the evaluations were an excellent method for providing another independent design verification. These discussions did, however, point to another program weakness. Since Assurance QA is performing an in-line review function of design pack 1ges, the QA organization should be assessing them in their audits of desig Further, the interviews with Assurance QA personnel and their manager and a review of the scope of their intended function indicated that they were understaffe Nine people were performing procurement activities while only five were designated for design activitie Of these five, several were involved with various plant "task force" duties which take time away from design activitie This appeared to be another program weaknes (4) Review of FCRs in Closeout To assess implementation of the design process relative to completed FCRs, the team selected nine FCRs from the October 17,
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1985 Confirmatory Action Letter (CAL) (No. CAL RIII-85-13, Item 1.a(4)) relating to safety-related piping system operability and twenty-one others selected at random (see Attachment B for a complete listing). The FCRs were primarily reviewed for completion of maintenarce work orders (MW0s), the revision of procedures and the completion of required trainin The team also reviewed design do:umentation to assess des 4 n verification, 10 CFR 50.59 safety evaluations, calculations, records, and interface informatio In addition, items were walked down in the plant to verify correct installation per the FCRs. This review produced the following results:
(a) FCR 78-126: Modification to the drain lines from the j gsm generators to the condense Design changes were madu to allow their use at normal operating pressures and temperatures for feed and bleed during start-up and shutdown to maintain water chemistry. The inspectors reviewed
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documentation associated with this package including a i final design calculation filed with the package identified l as Calculation C-ECS-063-002, Revision 0, dated May 17, 193 The stated purpose of the calculation was to determine horsepower and torque switch settings for motor operated >
valves MS 603 and MS 611 which are containment isolation valves in the steam generator blowdown lines. The inspectors identified a number of deficiencies in the calculation:
- No substantiation or references were provided for any of the design input used in the analysi * The methodology used to calculate horsepower had no apparent technical basi * There was no basis provided for equations used to calculate valve stem factor or opening torque for the valv * The source of the opening torque used to calculate horsepower was not provided and was not based on any apparent results of the equation identified to determine torqu * The inspectors were unable to comprehend the methodology used since the originator was no longer employed at TE Despite these inadequacies, the calculation stated, "check performed in accordance with Exhibit VI, checking of calculations; results are satisfactory." The inspectors
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concluded that the results of this analysis were not meaningful and the analysis was not consistent with the requirements of ANSI N45.2.11 relative to design verification and design inpu In response to the inspector's concern, Toledo Edison informed the team that subsequent to the June 9, 1985 event (loss of all feedwater), Davis-Besse had initiated a comprehensive Motor Operated Valve Reliability Progra In accordance with this program, all safety related valve motor operators were tested using the MOVATS testing systen. In addition, calculations for all safety re7ated valves were performed to determine worst case design differential pressures. The results of these calculations were filed with the design data for each valve.
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Toledo Edison provided the summary of the calculations (CME 3.01-204) for MS 603 and MS 611 which indicated that tne worst case design differential pressure was based on the steam generator relief valve set pressure, 1050 pli Valve testing (MOVATS) was based on this differential pressur The inspectors had no further questions relative to tte design differential used for these MOVs. However, the inspectors were concerned that the design basis calculation still on file (C-ECS-063-002) may be used in the future by engineers as a source of design basis data in the performance of design modifications or other safety related work. Further, it is not known whether other similar calculetions may exist which have been superseded by those in the Davis-Besse MOV reliability program files. These files have been updated to reflect new calculations and the results of M0 VATS testin Toledo Edison was unable to provide a resolution to this question during the inspection. Consequently, this item remains open pending TED resolution and subsequent NRC review (346/88006-01).
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(b) FCR 79-308: Facility change request which added a second independent and redundant miniflow recirculation line from the High Pressure Injection (HPI) puraps to the Borated Water Storage Tank (BWST). Previously, the recirculation line from each pump was cross connected into a single line returning to the BWST. In their review of the Safety Evaluation for this FCR, the Safety Review Board (SRB)
noted a concern that the modification could result in reduced HPI flow to the reactor coolant syste The response indicated that HPI flow would not be significantly altered. However, it was not obvious from the response to i
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. not result in reduced HPI flow. The SRB indicated that the response was inadequate and that more detail was required. Although further elaboration was provided, it was not clear that the issue had been addressed. The team could find no documented analyses to confirm that HPI flow would not be reduce ,
In response to the inspector's concern, Toledo Edison provided a draft calculation which indicated that the additional path provided for recirculation would not significantly reduce HPI flow since the orifice in both lines is the major resistance to flo ,
This item is not safety significant since the draft calculations provided to the team demonstrate that adequate HP.! flow is provided to the reactor coolant system with the modified systen. However, the item indicates a weakness of the part of the licensee in the documentation of design analyses relative to facility change requests. This item remains open pending completion of a finalized calculation confirming adecuate HPI flow to the reactor with the *
modified system and subsequent NRC Review (346/83006-02).
(c) FCR 85-160: Modification to install a drain line from the pressurizer power operated relief valve (PORV) loopseal to the pressurizer surge line. The FCR package incVuded a Design Review Caecklist as part of design verification to assure that app"opriate design considerations had been made and documented for the FC The inspectors reviewed :
the Design Verit'ication Checklist and found that the reviewer had annotated several items with the following comments:
1 "Assume that loads developed as a result of pzt spray actuation has been considered. (This would be a water slug rushing up the drain line when a delta-P is ,
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developed between steam space pressure and RCS pressure." (Item 6)
i 2 "Assumed to have been performed." (Hydraulic analyses -
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3 "Assumed" supporting calculations completed, checked, o and approve (Item 27) ;
The inspectors were concerned that the purpose of the checklist as a design verification tool was being r
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In response to the inspector's concern (TED memo to file dated March 10, 1988, NED 88-20156), Toledo Edison determined that a seismic analysis was performed for the line (Item 6) which adequately addressed the issued raised by the checklist and that appropriate calculations had been performed (Items 11 and 27). The inspectors had no further questions on this issue. However, this observation contributes to the team's concern that design activities were not adequately documented for FCR's performed at Davis-Bess (d) FCR 85-0204: Main Feed Pump Turbine (MFPT) high discharge pressure tri This FCR added a 0.1 second time delay relay to the nonsafety-related pump discharge pressure switch PSH-506 (MFPT 1-1) and PSH-582 (MFPT 1-2) trip l
logic. The time delay was added to prevent spurious trips due to high discharge pressure transients. The PSHs trip their respective MFPT on a pump discharge pressure of 1500 psig. The setpoint was based on the design pressure of the high pressure feedwater heaters (1500 psiH) which are located downstream of the pump The Instrument Information Sheet, for both PSHs, gave the PSH setpoint as 1500 psig i 15 psig and suggested the measuring and test equipment (MTE) be equivalent to a 0 to 2000 psig Heise gauge, t 2 psig. The "As Left" setpoint obtained from the instrument calibration records was 1500 psig for PSH-506 and 1495 psig for P5H-582. The Instrument Information Sheet provided for setting the PSH higher than the feedwe.ter design cressur The inspectors discussed with the licensee the need to factor in all the setpoint calibration errors to prevent exceeding the feedwater design pressure. Pending further ntview, this is considered an open iteta (346/88006-03),
(e) FCR 85-293: Change setpoint for pilot-operated rn_ lief valve (PORV) actuation to 2450 psi _g. The PORV is a safety grade valve with a non-safety grade actuator. Itu safety function is to retain pressure boundary integrity of the Reactor Coolant System (RCS). Operability of the PORV was intended to minimize lifts of the safety grade pressurizer code safety valves. Credit was not taken in Technical Specification (TS) Bases 3/4.4.3. for opening the PORV during any anticipated transient !
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The TS requirement for the safety valves was a lift setting of.1 2525 psig corresponding to the ambient conditions of the valve at normal operating temperature and pressure. The F PORV cpening setpoint in TS was > 2390 psig. The licensee
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determined the margin between the_ opening of the PORV and the safety valves based on the hot setting of the safet ;
valve lift setpoint and the total PORV setpoint error
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(FCR 85-293). The licensee uses procedure MP 1401.02, '
"Pressurizer Code Relief Valve Removal, Disassembly, Repair, Assembly, Installation, Testing, and Reinsta11ation," in verifying the lift setpoin The procedure provides both a hot and cold setpoint method. At the time of this FCR, tha safety valves were set with the hot method. The licensee has developed a graphical signature (valve bonnet operating temperature vs. lift pressure) for each safety valve. The graph represents a linear function. The lift pressure de:reases as the bonnet temperature increases (conservative direction). The licensee is now using the ;
cold method (three lifts at 1 2525 psig) to provide the cold lift setpoin The hot setpoint can be determined from the graph for a safety valves' nominal operating temperatur The margin previously determined in this FCR has changed do to using the cold setpoint methodology. The inspectors discussed this item with the licensee including the need to determine the setpoint margin for all required TS Modes of operation, and for each installed PZR code safety valve and the PORV opening setpoin The licensee was also requested to include in the determination all i instrument / calibration errors (PORV instrument string)
and the pressure gauge error used in calibrating the safety
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document the determination each time a safety valve is teste Pending further review, this is considered an open item i
(346/88006-04). !
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The inspectors reviewed the last code safety setpoint calibration and determined that the safety valves in use !
were operat,le with a setpoint 5 2525 psi (f) FCR 78-024: Containment Spray (CS) pump bearing oil sightglass installatio Prior to June 1977, the licensee !
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noted that during the operation of CS pumps 1-1 and 1-2 a ;
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oil level As a result, the oil levels were being kept i too low and bearing damage occurre To make it easier to monitor the oil levels, the licensee decided to install oil sightglasses on each pum This occurred in May 1977, !
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undtr MWO 77-079 l i
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The inspectors reviewed the documents contained in the ;
above FCR package. The inspectors noted that the licensee :
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performed the installation of the oil sightglass on the CS pumps without licensee commitments and NRC requirements being met in that:
1 No design drawings or detailed drawings were use No~ procedures / instructions were found for installation and inspectio ~
No documented design criteria / instructions were utilized for seismic qualification evaluatio The above findings were discussed in detail with licensee representatives. No dissenting comments were received from the licensee. The inspector informed the licensee that these findings-were examples of a violation of 10 CFR 50, Appendix B, Criterion V (346/88006-05).
In addition to the aforementioned findings, the inspectors noted that in the time period between 1977 and 1986, the installed oil sightglass assemblies were not seismically qualified although the plant was in operation. The NRC inspectors held discussions with licensee representatives regarding the potential reportability and operability requirement At the time of this inspection, the licensee was not able to complete the evaluation of reportability and operability. Pending further review, this matter is identified as an unresolved item (346/88006-06).
The inspectors reviewed seismic qualification Calculation No. C-ME-61.01-076, dated February 19, 1986, and the revised Calculation No. C-CSS-61.01-102, dated March 4, 1968, for the installation of CS pump oil sightglass assemblies. The inspectors noted that the calculated stresses were well below the allowable stresses set forth by the applicable ASME code. Consequently, the installed oil sightglass assemblies were seismically qualified by the above calculations. The inspectors found a numerical error in the pump mass ratio calculation. However, it does not affect the outcome of the calculation. The inspectors concluded that the installed oil sightglass assemblies could perform their intended function during a seismic event; however, the error in the calculation was another example of inadequate design verificatio (g) FCR 85-010: Support modification on the Auxiliary Feedwater syste The inspector's review determined that the material identification for Item No. 8 was missing on Support Drawing No. GC-EBB-4-H11, Sheet 1 of 4, Revision T This was a further example of inadequate design verificatio _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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This FCR modified a snubber support on the Auxiliary Feedwater piping to steam generator 1-2. NCR 85-003 identified two concerns. One was the interference of the snubber on support GC-EBB-4-H11 with its end bracket. The other was the support assemblies that were not installed in accordance with the applicable design drawings. The as-built support assemblies were reviewed by Bechte It was determined that the support assemblies were to be acceptable in terms of interim operation. For long term ,
operation minor modifications were required. The modifications contained in this FCR involved rotating the *
two snubber end brackets 90 degrees and the rework of support members so as to permit snubber position to be horizontal during normal operation. In addition to the above support modification, the pipe local stresses were also evaluated due to the complex support assemblie On the basis of the above review, the inspectors concluded that the support modified could perform its intended functio In general, the team found the review of documentation provided .
with the FCR packages to be cumbersome and not well organize ,
Although the team was able to assess the closeout of maintenance work orders for the FCRs reviewed, the team could not assure -
that any FCR package contained all required documentation, or that appropriate checklists had been executed to assure completion of the closeout process. Specifically, the team could not confirm that all procedures had been revised or training conducted as required. However, the team was encouraged by the initiation of a modification procedure which seeks to correct many of the problems contributing to the cumbersome closeout proces .
The team found that the implementation of M0 VATS testing and the motor operated valve reliability program was a positive step in the resolution of MOV problems at Davis-Bess (5) Review of New FCRs/ Modifications To assess implementation of the design process relative to new ,
! FCRs and Modifications, the team selected three design change l packages at random from those scheduled for the fifth refueling
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outage. Two of the packages were FCRs No.84-002, "Reactor
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Vessel Head to Hot Leg Vent Line Piping," and No. 86-0432,
"Feed and Bleed Enhancements." The other package was a ;
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modification, No. 87-1107,'"Recommended Improvements to Steam
- and Feedwater Line Rupture Control System _(SFRCS)."' In addition, the scheduled replacement of Station Batteries No. 2P and No. 2N was reviewed even though it was neither an FCR nor a modificatio The reviews of FCR No.84-002 and modification No. 87-1107 were completed with acceptable results. The reviews of FCR No. 86-0432 and the battery replacement generated the following results:
(a) FCR 86-432: Enhancements to feed and bleed capability Toledo Edison committed to the NRC (Serial 1382, dated June 25, 1987) to enhance the present feed and bleed
, capability through a modification to the makeup syste Feed and bleed is not part of the design basis for Davis Bess However, the upgraded makeup system would provide increased flow and add independent flow paths for each pum Subsequent to these improvements, feed and bleed could accommodate failure of either makeup pump or the PORV, in addition to multiple failures of steam generator cooling systems which would necessitate the initiation of feed and bleed operations. All new equipment is to be purchased and installed as nuclear safety-related, Seismic Class Where possible, existing equipment will be upgraded to meet nuclear safety related requirements. These modifications are to be made in two phases during the fifth and sixth refueling outage. Feed and bleed enhancements are being made under FCR 86-432; however, the design is
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not yet complet The inspectors reviewed correspondence contained in the correspondence section of the FCR file relative to feed and bleed enhancements to be made under FCR 86-432. A listing of equipment located in the makeup pump room which identified qualification requirements was included in this packag The listing indicates that the startup lube oil pump motors for the makeup pump will not be qualified for the environment in the makeup pump roo The reason given for this was that the startup pumps are used only during startup of the makeup pumps and, at that time 3 normal temperatures prevail. However, in Bechtel Calculation 540-72-22501, Revision 0, dated September 25, 1987 (see i
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discussion in design Analysis, below), credit was taken for shutting down one of the makeup pumps to minimize the heat load in the roo The inspectors concluded that shutting down one of the makeup pumps essentially eliminates the proposed feed and bleed enhancements as committed to the NRC in Serial 138 A failure of the redundant pump would render the plant
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with no means of feed and bleed since the shutdown pump could not be restarted in the high temperature environment with the unqualified startup lube oil pump moto In response to the team's concerns, Toledo Edison advised that they were investigating the possibility of qualifying the startup lube oil pump motor for the temperatures determined in the pump room heat up analysi If this is determined not to be feasible, the room heat up analysis will be revised to consider both pump motors operating, and the effects on temperature will be determine The licensee stated that all required equipment will be evaluated for operation in this environment and procedures will be written to reflect the need to start (and continued operation of) both pumps if require This item is not safety significant since the feed end bleed enhancements are outside the design basis for the plan However, the observation indicates that the NRC commitment to enhance feed and bleed capabilities may not be achieved if the startup lube oil pump motor cannot be qualified for its operating environmen Further, this observation indicates a weakness in the coordination of dnsign basis information developed in design analysis with interfacing groups (e.g., equipment qualification). This item is considered unresolved pending completion of these evaluations and review by the NRC (346/88006-07).
The inspectors reviewed design documentation related to the feed and bleed enhancements made in FCR 86-432 including a number of design analyses. The team found that these calculations contained flawed methodologies, did not always consider worst case operating modes, and were based on unverified or unsubstantiated assumption The following were examples of these deficiencies:
1 MPR Associates, Inc., calculation "Options for Increasing Makeup System Capacity at Davis-Besse,"
dated December 30, 1986, was performed to determine the makeup system flow capacity for both the existing and modified system The results of this calculation (flows at various reactor pressures) were used as
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input for the B&W analysis (32-1168039-00) of feed and bleed capability for Davis-Besse. The inspectors identified the following concerns with the MPR calculation:
a The methodology used to determine the makeup flow to the reactor coolant system consisted of determining the system resistance curve for the existing and modified system for several
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, different cases, and algebraically modelling the Recirculation flow to the
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makeup pump curv BWST was assumed to be 35 gpm and a flow of 32 gpm was assumed to the reaccor coolant pump seals. Based on these assumptions, the flow to the reactor coolant system was calculated in each cas However, the inspectors found that flows in this type of flow network must be determined based on an analysis of the resistance in each flow path. Depending on the relative resistance of each path, flow to the reactor coolant system could vary significantl The inspectors noted that flow to the reactor coolant pump seals is determined by a flow control valve. Thus, flow in this path should be based on the worst case control valve differential pressure plus other valve and line losses and for the modified system, recirculation flow is to be isolated, b There was no basis given for the 10 psi suction pressure use c The modelling used for the modified system (second line) was not clear. The team was advised that the same piping losses were assumed for the second line as the existing makeup line which implicitly assumed the same line lengths for the new lin In response to the inspector's concerns, TED indicated that a final calculation of flows in the upgradeo system has been recently completed, and the calculation was being reviewed. The results of this calculation indicated higher reactor coolant system flows than those used as input for the B&W analysi TC) also indicated that this calculation would evaluate NPSH provided to the makeup pumps to assure that vendor NPSH requirements are satisfied during feed and bleed operations since this has not been don The inspectors had no further questions on this calculation. However, this item remains open pending final review and issue of the TED calculation of feed and bleed flows for the enhanced system (346/88006-08).
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2 Bechtel Calculation 540-72-12501, Revision 0, dated September 25, 1987, "Analysis of the Makeup Pump Room (Rm 225) Heat Up," is a calculation to determine the temperature response of the makeup pump room during feed and bleed operations. The results of this calculation serve as the basis for equipment qualification temperatures in this room since the room coolers for the room are not safety grade and are not being upgraded to safety grade equipmen The inspectors had the following concerns with this calculation:
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a The calculation assumed 30% of the room and 10%
of the adjoining vestibule were congested (i.e.,
occupied with equipment, piping, ductwork, etc.).
No basis was given for these assumptions. If the congestion were actually more than assumed, room temperatures could be higher than calculate In response to this concern, Bechtel provided data which indicated that actual volume occupied by equipment and piping was less than 20% of the room volum b A Bechtel memorandum attached to the calculation indicates that a 1.15 multiplier (applied to pump motor heat load), originally used to account for piping and lighting heat loads, was delete Instead, actual piping in the room was modelled as a heat sink. However, no consideration of heat loads due to lighting in the room was include c
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The calculation took credit for heat transfer to cooler piping in the room via natural convectio The team found that heat transfer to cool piping via natural convect'on is not conservative and may not be effectiv ANSI N45.2.11 requires that assumptions should be verified or adequate substantiatiori provided to justify the assumptions. This item remains open pending resolution of these concerns and revision of the calculation to document the basis for the assumed room congestion and the other undocumented assumptions (346/88006-09).
3 Bechtel Calculation 34.34, Revision 0, dated September 25, 1987, was performed to "determine the pressure in the makeup pump while operating the pumps piggyback with the decay heat removal pumps." The
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calculation determined the maximum pressure imposed on the makeup pump suction and discharge piping. The maximum suction piping pressure was based on static head due to the BWST at maximum level and decay heat removal pump shutoff head. To determine the maximum makeup pump discharge pressure, 2691 psig, the makeup pump head at 170 gpm (approximate feed and bleed flow) was added to the maximum suction pressure. Baned on these pressures, the calculation concluded that maximum pressures were less than design. However, the inspectors found that the worst case discharge pressure would result from operation of the makeup pump at shut off head (e.g., against a closed pump discharge valve) while in the piggyback mode. The inspectors independently determined that maximum pressure could reach approximately 2900 psig in this case which is significantly greater than the pressure determined in the calculation. Since the design pressure for this line is 3050 psig, there is no safety concer The inspectors concluued that none of the concerns related to these calculations are ,afety significan However, these observations contributed to the team's concern that design verification of analysis for facility changes is a weakness at Davis-Bess (b) Battery Replacement The inspectors elected to look at the scheduled replacement of Statien Batteries No. 2N and No. 2P because of their important function in the safe shutdown of the plant. Tne batteries were scheduled for replacement because they were approaching the end of their service lif Station Batteries No. 2N and No. 2P are 60 cell, 1500 amp hour, lead calcium batteries. These batteries, together with D.C. Motor Control Center 2 and tne battery chargers make up Train B of the 250/125 volt bus, as described in the Davis-Besse Technical Specification Prior to entering the plant, the inspectors reviewed Purchase Order No. EN1Q-010783ST to GNB Batteries Inc., a synopsis of MW0s 1-87-1182-00 and 3-87-1184-00 for the battery replacement, and Specifici. tion Change Notice No. 01-03 to Specification No. 17.501-E-19Q, Revision 1,
"Technical Specification for Operational Phase for 250/125 Volt Station Batteries." The change notice changed the load profile for the battery performance discharge tes The inspectors interviewed key personnel responsible for the battery replacement and the performance discharge
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test.- lhis interview led to the review of Test Procedure No. ST 5084.02, "Station Battery Service and Performance Discharge Test." The inspectors noticed that the. load profile in the test procedure was more conservative than-the load profile required by the specification change. This information was given to Engineering and they were asked to explain the differenc Engineering subsequently issued a memorandurd which documented why the specification was changed and issued procedure change request No. 88-0336 to correct the procedure. However, it did not explain why the procedure was not changed when the specification change was processed, a time difference of six month Discussions were held with Engineering and the following sequence of events was constructed:
1 Surveillance Report Number 84-36 and NRC Open Item No. 50-346/82-21-14 questioned the battery capacit ~
Calculation C-EE-002-004, Revision 1, dated December 1984 determined the capacity was acceptabl FCR No. 82-0029, Revision E, changed the DC distribution Technical Specification to adopt the Standardized Technical Specification ,4, The revised Technical Specification was approved by the NRC on March 12, 1987, in Amendment 100 to License No. NPF- Specification Change 01-03 to Specification No. 12501-E-18Q, dated August 3, 1.987, should have required changes to the following interfacing documents:
a DB-ME-09200 b
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DB-ME-09201 DB-ME-3000 ii DB-ME-03002 (ST 5084.02)
j DB-ME-01001 6_ NRC inspection on February 26, 1988, determined that the above procedures had not been change Procedure Change request for DB-ME-3002 (ST 5084.02)
was processed on March 3, 198 This sequence of events made it obvious that Procedure No. NEP-021, "Specifications," was violated, in that, the required "Interfacing Document Worksheet" which would have
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accomplished the required changes was filled out but never processed. Further discussions were held with engineering management relative to the use of NEP-021 as a means for changing the specification. It then became '
clear that the above problem was an intermediate caus The root ~cause was that the procedure was misused, in that the battery load profile change was treated as if it were a minor change, such as correcting a typo, without applying full design change controls. Since the change involved technical matters, an FCR should have been processed for the change to ensure that the proper controls were implemented. Failure to apply design control measures to a specification change is a violation of Criterion III of 10 CFR 50, Appendix B (346/88006-10).
Within the areas inspected, two violations, two unresolved items, and six open items were identifie ,
4. Allegation (RIII-87-A-0170) (Closed)
On December 30, 1987, Region III received an anonymous allegation concerning the manner in which backlogged Field Change Requests were '
resolved under the Course of Action submitted to the NRC in the wake of the June 9, 1985 event. The allegation stated that actions were taken which did not follow established administrative procedures and in which documentation was mishandled and fabricate Three broad problems and one specific problem were identified: FCRs were submitted for closure which did not meet the criteri Closure was based on work requests which were voided or which were listed as open and work in progres FCRs were modified or consolidated which caused a delay in their implementatio Unrelated FCRs were consolidated and given an implementation priority not commensurate with their safety significance, FCRs which modified the Containment Spray Pump Oil sightglass ,
isolated it from the oil reservoir for extended periods of time with plant operational and surveillance being conducted on the system. This was not evaluated for LER reportability by the Licensing Department as it should have bee NRC Review To assess these problems, the inspectors reviewed the FCR program procedures, selected and reviewed a sample of FCRs closed between the present and implementation of the Course of Action Program in 1985, and reviewed the Containment Spray Pump oil sightglass FCR. The results of the inspectors review are as follows:
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.(1) FCRs were submitted for closure which did not meet the criteri The inspectors reviewed the FCR program procedures to determine the criteria for FCR closeout. The following procedures were reviewed: NEP-010. "Processing Facility Change Requests;"
NEI-010.2, "FCR Closeout Instructions;" and AD-1845.03,
"Facility Changr Request Implementation." The inspectors also '
interviewed key members of the FCR closeout organization
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relative to criteria for FCR closeou To assess whether FCRs were closed without meeting the established. criteria thirty-three FCRs out of the original 448 FCRs in the FCR closeout commitment to the NRC were reviewed in detail with special attention paid 1c the use of Maintenance Work Orders (MW0s) in the closecut proces ,
The inspectors were not able to substantiate this concern. No extmples were found where the closeout criteria were not me Specifically, no cases were found of FCRs closed by MW0s that were later voide (2) FCRs were modified or consolidated which caused delays in their implementatio The inspectors were able to substantiate this concern but only in the time period preceding the 1985 closecut commitmen The alleger was referring to the practice of continually adding supplements to FCRs so that they became so large that final closecut was difficult. This fact was recognized by the NRC and resulted in a violation in Inspection Report No. 50-346/85031-02 which stated "continually adding ,
supplements to some FCRs (changing scope of work) partially I contributed to some nonconforming or deficient conditions not being corrected in a timely manner." This violation was corrected and the violation was closed in Inspection Report No. 50-346/86004. The alleger was referring to a problem t identified by the NRC whose correction was tracked by the NR Therefore, this item has no safety significanc (3) Unrelated FCRs were consolidated and given an implementating priority not commensurate with their safety significanc The inspectors review of the program procedures documented in (1)
above indicated that FCRs were not allowed to be consolidate However, unrelated supplements; such as; mechanical, electrical, and I&C were routinely combined under a common FCR number. This item was previously identified by the NRC as a i
violation and tracked to closure. See Section (2) abov The inspectors determined that this item was not safety significant.
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i (4) Containment Spray Pump Oil Sightolass. This concern was substantiated. See Section 3.c.(4)(f) of this report for detail It will be tracked until closure as Unresolved Item No. 346/88006-6 pending NRC review of the licensee's evaluation
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S. High Pressure Injection (HPI) Direct Current (DC) Lube Oil Pump Review l
i During a configuration management walkdown, the licensee discovered that
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, the thermal overload relay was installed upside-down for each HPI pump DC lube oil pump. Also, it was discovered that the installed control circuit fuses were rated at 1 Amp while the elementary drawing (E528 - Sheet 64, "Reactor Cooling System HPI Pump DC Lube Oil Pump")
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specified a 10 Amp fuse; and the installed supply fuses were rated-at 15 Amp while the one line diagram (E-7, "250/125V DC and Instrumentation AC") specified a 10 Amp fuse. The licensee conducted a search of
- maintenance history records which showed no equipment replacement, system modifications, or control circuit changes. The licensee believed this condition had existed since initial equipment installatio The inspectors reviewed the licensee's corrective actions. Tests were conducted to determine the motor and control circuit inrush and nominal operating current. These are listed as follows
i In-Rush Time of Operating Circuits Current In-Rush Current Current DC Motor 12.8a 0.6 second 2.6a Control 0.7a Not Determined 0.22a l These values support the original design. The supply fuse should be
10 Amp and the control fuse should be 1 Amp (determined from fuse curves
, and size availability). The licensee issued Design Change Notice E528-442 which changed the 10 Amp fuse to 1 Amp on drawing E52B, installed the 10 Amp fuses in the plant, and correctly positioned the overload rela Further review of drawing E528 revealed the overload relay contacts were
- not connected to the control circuit. The inspectors verified these
{ items were correct per the "As-Built" plant design. The Electrical Superintendent indicated this was a common practice at Davis-Besse, that safe shutdown equipment supplied with overload relays would not have the relay contacts wired into the circuit.
i The inspectors reviewed the initial findings and corrective actions to determine HPI operability. The overload relay had no affect on the HPI
- operabilit The inspectors determine the supply fuse (15 Amp) installed
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since initial equipment installation provided adequate motor (overload)
, and wiring (fault) current protection. The 15 Amp fuses were coordinated r with the upstream fuse (800 Amp). Review of the 10 Amp fuse melting
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time-current curve indicated the fuse would open in approximately
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60 seconds on a 128% overload current. This is greater than the inrush cut rent time of 0.6 seconds and will assure the fuse will not open during i the til pump start cycle.
I The inspectors have no further questions on this item and have concluded that botn HPI pump DC lube oil pumps were operable from initial plant
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. Followup on Items from OSTI The inspectors performed a follow-up review on two items from the OSTI (50-346/87-24) as follows: Procedures are not updated in a timely manner following plant modifications. The inspectors reviewed this program weakness and found that the OSTI report was based on only one example. The review of 30 previously closed FCRs during this inspection did not support this concern. This item is considered close A large backlog of engineering responses to Corporate Nuclear Review Board questiuns to 50.59 evaluations exists, primarily related to plant modification The number of safety evaluations requiring response at the time of the OSTI was 12 As of March 1, 1988, this number has been icduced to 27. This item is considered close . Open Items Open items are matters which have been discussed with the licensee, which will be reviewed further by the inspector, and which involves some action on the part of the NRC or licensee or bot Open items are discussed in Paragraphs 3.c.(4)(a), (b), (d), and (e); and 3.c(5)(a) 1 and . Unresolved Items An unresolved item is a matter about which more information is required in order to ascertain whether it is an acceptable item, an open item, a deviation, or a violation. Unresolved items are discussed in Paragraphs 3.c.(4)(f) and 3.c.(5)(a). Exit Interview The inspectors met with lice.1see representatives denoted in Paragraph I during and at the conclusion of the inspection on March 11, 1988. The inspectors summarized the scope and results of the inspection and discussed the likely content of this inspection repor The licensee acknowledged the information and did not indicate that any of the information disclosed during the inspection could be considered proprietary in natur Attachments: Personnel Contacted Facility Change Request Review Procedure Review Audits Reviewed
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i Attachment A: Personnel Contacted Toledo Edison -
T. W. Anderson Maintenance Planning and Outage Management Superintendent i
M. L. - Borysiak Senior Instrument and Control Engineer I 0. R. Breese Facility Modification Supervisor W. S. Delicate CNRB Administrator I R. C. Elfstrom Nuclear Specialist l
G. N. Ferguson Senior Engineer 1
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0. J. Harris Quality Systems Manager P. C. Hildebrandt Engineering General Director l
G. Honma Compliance Supervisor T. R. Isley Lead Instrument and Control Engineer S. C. Jain Nuclear Engineering and Independent Safety Engineering Director.
, J. J. Johnson Operations Engineering Supervisor J. R. Kasper Operations Superintendent
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j J. B. Keagler Associate Nuclear Engineer - Electrical D. S. Knaszak Engineering Services Manager M. J. Knaszak Design Engineer D. R. Lightfoot Facility Modification Department i' Superititendent
! D. J. Mominee Engineering Assurance Design Supervisor
- J. E. Moyers Quality Verification Manager J. A. Nevshemal Mechanical Engineering Manager i
L. O. Ramseth Quality Assurance Director l T. B. Ridlon Nuclear Technologist
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Attachment A 2
R. R. Rinderman Quality Verification Supervisor E.-M. Salowitz Planning Superintendent E. D. Schock Assistant Nuclear Engineer - Electrical R. W. Schrander Nuclear Licensing Manager L. - F. Storz Plant Manager J. C. Sturdavant Licensing Principal R. J. Swain Assistant Nuclear Engineer T. S. Swim Civil / Structural Engineering Manager G. L. Tillman Design Process Supervisor V. Watson Design Engineering Director A. G. Weedman Engineering Assurance Manager J. A. Wells Engineering Assistant Analyst / Modifications B. L. Wrightman Engineering Assistant
A. K. Zarkesh Independent Safety Engineering Manager Bechtel
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C. A. Leskovar Design Engineer l M. L. Murphy Senior Engineer
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- Attachment B: Facility Change Request Review FCR Number Description 78-0024 Containment Spray Pump 011 Sightglass 78-0126 Modify Drain Lines for Feed and Bleed Use During Startup and Shutdown 79-0308 Add Mini Recirculation Line From HPI Pumps to BWST 80-0221 Relief Valves SW3962 and SW3963 Setpoint Change 83-0063 HPI Lube Oil Pump A.C. Motor Replacement 83-0136 Replacing Existing AFW Pump Turbine Governors 84-0002 Reactor Vessel Head to Hot Leg Vent Line Piping
84-0111 SW Pump Pressure Switch Setpoint Change l 85-0010 AFW Piping System Support Modification 85-0086 MS Piping System Support Modification-85-0126 Removal of Core Drill Sealant Material and Replacement with Non-Shrink Grout 85-0159 SFRCS Low Steam Pressure Trip Logic
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85-0160 PORV Loopseal Drain Line Installation 85-0164 AFPT Overspeed Trip Mechanism Replacement 85-0167 SFRCS Full Trip Alarm 85-0201 ICS Steam Generator Low Level Setpoint 85-0204 MFPT High Discharge Pressure Pump Trip 85-0224 MS Line 'A' Snubber Addition 85-0239 EDG Room High/ Low Temperature Alarm 85-0242 LPI Flow Trans:;.itter Replacement 85-0263 RCP Seal Leakage Instrumentation Removal 85-0293 Raise PORV Opening Setpoint
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Attachment Makeup and Purification' Flow Instrumentation
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85-0328 86-0002 RPS Fan Noise' Suppression 86-0016 SFRCS Pressure Switches 86-0093 EDG Overspeed Trip Modification 86-0148 AFW Press to Test Lights 86-0162 Auxiliary .'eedwater Controls 86-0174 Essential Steam Generator Level Control 86-0235' Flow Transmitter.FT-4909 Replacement 86-0300 SFRCS/AFW Integrated Test 86-0432 feed and Bleed Enhancements 87-0015 EDG Thermostat Setpoint Change
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Attachment C: Procedure Review Procedure Number Title Revision AD 1844.02 Control of Work 3 AD 1844.03 Facility Change Request 0
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Implementation AD 1844.03 Facility Change Request 1 Implementation AD 1845.04 Facility Change Request 0 Closeout DB-ME-3002 Station Battery Service 8 and Performance Discharge
. Test EN-DP-01203 Engineering Design Draft A Evaluation NEI-01 FCR Closeout Instructions 3 NEI-02 Instructions for PICA Forms 0 NEI-20 MOD Processing Instructions 0 l NEP-010 Processing Facility Change 1 Requests NEP-011 Conceptional Design 0 NEP-020 Design Work Packages 0 NEP-200 Processing Plant Modifications 1 NG-NE-0301 Plant Modifications 1 QA-EA-1102.03 Quality / Technical Reviews 0 l
QA-EA-1105.03 Review of Facility Change 2 Requests and Plant
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Hodifications I
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I Attachment 0: Audits Reviewed :
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Year- Audit Report Number
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1985 1349 1371 i 1344 l 1439 i
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1986- 1486 ,
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1511 1516 1572 AR-86-DESIGN-01 1987
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AR-87-DESIGN-01 AR-87-DESIGN-02 AR-87-DESIGN-03 AR-87-DESIGN-04 AR-87-BECHT-01 !
AR-87-BECHT-02 !
AR-87-BWNPD-01 L AR-87-B&R0E-01 -
AR-87-SWENG-01 i AR-87-TESTC-01 t
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AR-87-TRAIN-02 AR-87-TRAIN-03 3 AR-87-TRAIN-04 .
AR-87-CORAC-01 l AR-87-CORAC-02 [
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