IR 05000346/1990010
| ML20055G664 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 07/19/1990 |
| From: | Darrin Butler, James Gavula, Jeffrey Jacobson, Lougheed P, Phillips M, Yin I NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20055G663 | List: |
| References | |
| 50-346-90-10, IEB-79-14, NUDOCS 9007240042 | |
| Download: ML20055G664 (26) | |
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U.S. NUCLEAR REGULATORY COMMISSION L
REGION III
Report No. 50 346/90010(DRS)-:'
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DocketNo.(50-346 License No. NPF-3 Licensee: Toledo Edison Company-300 Madison Avenue-Toledo, OH 43652
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Facility Name:
Davis Besse Nuclear Power Station Inspection At: Oak Harbor, OH 43449 p
' Inspection Conducted * Ap il 30 thru May 18, 1990 b der 7 !/9
Inspectors:
M. Ja on
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f/1f96
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Date I!fd P. V 41/rp IT!@
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il ips, Chief 7!/9h9 Approved By: - Operational Programs-Sectionbate Insoection Summary
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~Insoection on Aoril 30.thru May 18. 1990. (Recort No. 50-346/90010(DRS).
- Areas insoected: -Routine announced safety inspection of the engineering and
- technical; support function. The inspection focused on' design (37700),
installation-and testing of moJifications (37828) and commercial grade f
, procurement (38703).
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Results: Of-the areas inspected, no violations or deviations were identified.
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The following unresolved items were identified:
lack of electrical backup
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protection for certain electrical penetrations (Paragraph 4); incorrect translation of. electrical setpoints developed by engineering into
- p implementation documents (Paragraph 3.c); questionable validity of IEB 79-14
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piping and support calculations used as design input for plant modifications (Paragraph 3.b); deficiencies associated with instrument and station air system modifications (Paragraph 3.b).
Within the functional area of engineering and technical support, the following observations were made:
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o Documentation,-from design through implementation, was considered
. generally thorough and.well-organized.L j
o Experience level of ~ the engineering staff was high and attitudes
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generally excellent.-
j Th'e efforts offthe Independent Safety Engineering and Engineering o
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Assurance-Departments are. indicative of-a-proactive attitude towards
-self assessment and improvement.
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o More thorough walkdowns and aggressive modification planning prior to
- the outage, combined _with the integration of modification coordinators into the design function have resulted in an improved engineering product.
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Areas of the design function still in need of improvement include.
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attention to_ detail and verification of calculations.
Documentation of
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the bases for engineering.judgements is also in need of reinforcement, o~
Weaknesses in the area of small bore pipe welding and inspection control
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were identified.
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L DETAILS
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Persons Contacted
- - -y Toledo Edison Company (TED)
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- D. Shelton, Vice President, Nuclear-L. Storz, Plant Manager
- G. Gibbs, QA Director-
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- S.'Jain, Engineering Director
- E. Salowitz, Planning and Support Director
'E. Caba, Performance Engineering Manager-q
- V Watson, Design Engineering-Manager-j D. Haiman, Engineering Assurance Manager
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C. Hawley, facility Modification Manager M. Stewart, Nuclear Training. Manager
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- J. Lash,_Indpendent Safety.Enginnering Manager
- J. Moyers, Quality Verification Manager
- D; T M s, Systems Engineering Manager
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- L Worley, Quality Systems Manager
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Schrauder, Nuclear Licensing Manager
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T.' Anderson, Maintenance Planning Manager i
- R. Brandt, P1 ant'Operatlons Manager-(
- B. :Walrath, Engineering Support Manager
- K. Prasad, Nuclear Engineering Manager
- G. Homma, Compliance. Supervisor R. Simpkins, Training Supervisor P. Jacobsen, Design Engineering Supervisor
. A. Zarkesh, System Analysis Supervisor
H. Stevens, Indepdendent Safety. Engineering Supervisor
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R. Collings, Audits;' Supervisor D. Staudinger, Engineering Services. Supervisor
- A. VanDenabeele, Engineering Assurance Supervisor
- E. Chimahusky, Test Projects Supervisor-J. Holden, Mechanical Design Supervisor
- C, Williams, Senior Projects Engineer
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P. Straube, Senior Design Engineer
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G. Hayes, Nuclear Engineer
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- R. Holliday, Licensing Engineer
- T. Swansiger,-Senior Engineer, Support V. Kumar, Senior Engineer, Mechanical
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P. Goyal, Senior Design Engineer M. Harris, Senior Design Engineer
- Denotes those attending the exit meeting on May 18, 1990.
l The inspectors also contacted other licensee and contractor employees.
2.
Enaineerina and Technical Sucoort Function Overview The efforts of thi: portion of the inspection were directed at evaluating the effectiveness of the technical support organization at
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DavisLBesse. While interaction between departments was also considered the major _. focus was to the overal engineering product.
The' approach
- taken forLthis portion of. the inspection was a review.of engineering documentation and procedures coupled with interviews of department managers.
The Independent Safety. Engineering (ISE) group serves as an independent.
review body' utilized to evaluate such things as engineering, operating, maintenance, and modification activities.
ISE performs evaluations, analyses and reviews in support of both programmatic and design specific;
issues..The selective review of design changes, safety evaluations, and plant procedures is also part of the ISE charter.
The NRC-team reviewed the safety system functional. inspection (SSFI)
performed by ISE during Summer, 1989. This self-initiated inspection focused on the Station and Instrument AirJSystem and is discussed in
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. Paragraph 3 of this report.
The team also reviewed the safety system outage modification inspection (SS0MI) performed by ISE during the current outage. This SS0MI covered the design, implementation and post
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modification activities associated with selected modifications, simple configuration changes, and temporary modifications.
Two of the
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modifications included-in the licensee's SSOMI were reviewed in detail by the NRC team (FCR' 85-0063, " Install' Ground Fault Protection," and MOD 87-1198, " Replace Valves AF599 and AF608") and are discussed in Paragraph-3 of this report.
Both the.SSFI..and SS0MI were found to be good efforts on'the part of the licensee.
The activities performed by the ISE group are viewed as demonstrating a proactive attitude towards.self assessment and improvement.
The Performance Engineering group is a line organization responsible for operating experience assessment, station performance, and test j
engineering.
The operating assessment function collects and evaluates
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data from various industry networks, for applicability to the Davis
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Besse plant.
Also, under this function, all plant modifications are reviewed for human factors issues. The test engineering function is responsible for inservice testing of pumps and valves, pressure tests, j
and post modification testing. The station performance function is l
responsible for the implementation of the inservice inspection program
and predictive maintenance of valves, and rotating equipment.
The
.i performance group also collects thermographic data for evaluation and predictive maintenance.
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The NRC team reviewed the Performance Engineering group's efforts in the j
P area of air operated valve (A0V) testing. A diagnostic testing system
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was developed for. A0Vs by Performance Engineering in conjunction with Minerva Research Corporation. The test equipment is capable of I
diagnosing such problems as incorrect packing loading, inadequate air
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supply, and incorrect calibration of the voltage to pressure converter.
k Performance Engineering's innovative approach to A0V diagnostics has
resulted in the detection of several problems (e.g., turbine bypass
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valve testing) that could not' be identified earlier by conventional means.. This diagnostic capability should provide a definite benefit in
-increasing the reliable and: safe operation of A0Vs.. Performance d
engineering's initiatives in this area are representative of a general proactive attitude toward self-improvement.
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The Systems Engineering group was established in 1986 and-has steadily developed.into a very effective link between the Operating-and Engineering. Departments. System engineers are assig.ied to all major plant systems and are responsible for the continued, efficient performance of those systems. A system engineer is required to be
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intimately familiar with all aspects of a'particular system and provides a knowledgeable contact point for system problem resolution.
The concept of " system ownership" has been found to promote sound engineering input, resulting in effectively designed plant modifications.
The systems group is currently in the process of completing preparation of " system descriptions" for'approximately 80 plant systems. These documents will provide in-depth information about a particular system for both the designer and operator.
Systems-Engineering is also responsible for maintaining an appropriate
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preventive maintenance program for system components, developing troubleshooting plans, problem investigations, ar.d root cause-determinations.
The NRC team viewed the Systems Engineering group as an integral part of a well organized engineering and technical support function..
The Engineering Assursnce (EA) group is.an innovative spin off of the 10 CFR 50, Appendix B required Quality Assurance. The EA group however, reports to.the Director of Engineering and provides a dedicated self assessment function '.o the engineering' effort.
The EA major activities are specified by an aperating plan approved by the Director of
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Engineering and include reviews of procedures, engineering training activities, specific design efforts, and technical documents.
Additionally, EA serves as another management tool, available to.the Director of Engineering, to assess potential problem areas, within'the engineering function, as they are identified.
The NRC team reviewed the EA 1989 Annual Report and found the effort to j
be successful in improving engineering quality and effectiveness through
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critical self-assessment.
Of particular note were the procedure
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reviews, design evaluations and procurement document reviews.
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two technical comments and 213 procedural comments were generated out of l
83 procedures reviewed by EA.
Four " vertical slice" design evaluations l'
were performed to assure tbt the design basis inputs of selected modifications had been appropriately defined and implemented in lower level documents. The resultant 42 concerns identified during these evaluations served to raise engineering management's awareness of the
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need for more careful checking of the engineering product before
issuance.
Procurement document reviews involved the review of 13,000
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documents in 1989.
By the end of 1989, the unsatisfactory document rate had been reduced to 5 percent from the previous'15 percent.
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The overall contribution of EA to the engineering and techr.ical support function is considered to be a very positive aspect of the licensee's program.
As an indicator of the constructability of completed engineering efforts, Design Engineering performed a trending evaluation of the issuance of Field Problem Reports (FPRs) from the fourth qu.arter,1988 through the first quarter, 1990.
This treading, based on FPRs per 100 craft man hours, indicates a reduction to about one half the previous rate for this refueling outage.
The improvement is attributed to factors such as an increased ratio of TED to contract personnel, changes to the modification process to require a constructability walkdown, and integrating six modification coordinators into the design process.
In an effort to increase the engineering presence in the plant and to monitor the effectiveness of the engineering function, a management plant tour program was established. Su)ervisors of the various design disciplines are required, on a monthly sasis to select a project of their choice and conduct a plant tour of the implementation of that project. This program serves as an aid to increase the level of awareness of the engineering management and obtain feedback, from the end users, on the quality of the engineering product.
Another activity which demonstrates a proactive attitude towards improving the engineering function was the development of an Engineering Department Policy Manual.
This document serves to state, and place in the forefront the company policy regarding such engineering functions as the use of engineering judgement, engineering evaluations, and arocedures.
The manual attempts to address those engineering issues (nown to be industry weaknesses.
The overall conclusion from this portit,n of the inspection with respect to organization, management controls, and general attitudes was positive.
If the licensee continues to fully implement the established programs and continues to actively pursue better methods of providing
. engineering and technical support, the current improving trend is
' expected to be maintained.
3.
Modification Review /Desian Control A.
Signe This portion of the inspection was directed at evaluating a sample of plant modifications in an effort to ascertain the quality of the engineering product. The evaluation consisted of such tHngs as verification of appropriate engineering methodologies, compliance with applicable codes and standards, validity of associated 10 CFR 50.59 reviews, correctness of supporting calculations, and adequacy of testing. Alse where possible, an n
attempt was made to verify that the modification was installed in
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accordance with the final design by performing a walkdown of the installation.. The evaluation included permanent mechanical, electrical, and instrumentation and controls madifications, as well as, a sampling of temporary modifications.
B.
-Mechanical Modifications (1)
Modification No. 89 0066
" Change Normal Makeup flowpath from Loop 2-1 to Loop 2-2" As documented in PCAQR No. 88 0576, the thermal sleeve on
one of the high pressure injection /make up(HPI/MU) nozzles j
to the enld leg reactor coolant (RC) piping failed due to thermal cycling. The TED letter of June 19, 1989 to the NRC
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committed to re route the makeup flow piping to an alternate
HPI nozzle.
This would eliminate the possibility of thermal cycling on the present makeup /HPI nozzle caused by the cold.
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makeup flow. The modification entailed cutting and capping the existing 2 1/2 inch diameter makeup connection and
adding several feet of new piping in order to reconnect to l
another HPI line. Other changes-associated with this work
included Modification No. 88 0145 which replaced valve MV
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169 with a less restrictive type of check valve, and Modification No.88-165 which replaced two HPI thermal sleeves.
The following calculation was reviewed for compliance with
NRC requirements and licensee commitments.
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Calculation No. 56B2. " Pipe Stress Analysis Report for Hiah
Pressure In.iection.JTrain 1) Pipina." Revision 6.
September 26. 1089.
l During this review the following discrepancies were noted in this calculation:
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o The peak pressure term for the piping down stream of valve HV HP2B was analyzed as 1650 psig instead of the
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specified 2500 psig. This resulted in an increase in
tie calculated stress by approximately 1600 psi.
F Although the resulting stresses still met the
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allowable stress limits the Equation 11 stresses increased from 96.2% to 99.7% of allowable, o
for the 2 inch diameter piping adjacent to valve MU 169, the pipe wall thickness was modeled as.375 inches instead of the specified.343 inches.
This resulted in under estimating the stresses at these locations by approximately 5%.
o The socket welded joints for valve MU 169 were modeled L
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I without stress intensification factors. This resulted
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in under estimating the stresses at these locations by
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a factor of 30%.
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o The model dimension for the location of support i-No. 33C-CCB 3 H2 was different from the dimension i
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given on the hanger location drawing DCN-HL 231D-4 by 11 inches. Although a direct field-measurement was i
not made during the inspection,,a visual approximation indicated that the model was not correct.
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difference exceeded the pipe support location tolerance of i 6 inches specified in the TED " Design i
Criteria Manual," Revision 0, but no documented.
i reconciliation was noted in the calculation package.
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Preliminary evaluations performed by the licensee during the inspection indicated that the system was still within code l
allowables considering all of the above discrepancies.
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Several of the above discrepancies were considered to be due to licensee inattention to detail by the NRC team.
However, since
most of these errors were attributable to mistakes;in the previous revision to the calculation; the assumption made by the licensee regarding the validity of the existing Bechtel piping calculation may not be justified.
Both the number and apparent magnitude of these errors indicate that a more detailed verification is
- warranted prior to using these calculations as design input.
Based on discussions with the licensee, some calculation reverification normally occurs prior to any revision. The issue then becomes how much reverification is necessary in order to establish confidence that the discrepancies in the existing i
calculation will not be safety significant. Although there were no examples of where these discrepancies caused the facility to
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exceed the design limits, this is attributable more to the existing conservatism in the calculation rather than the i
insignificance of the discrepancies. On this basis, the concern at this point is toward modifications where the overall design r
margins are potentially being reduced.
Pending a review of the licensee's action to establish the appropriate level of reverification for piping calculations used as design input for modifications, this is considered an Unresolved item (346/90010-01).
(2)
Modification No. 87-1198
" Replacement of AF599 and AF608 With 900 Pound ANSI Class Valves."
During testing of the motor operated Auxiliary Feedwater (AFW)
Valves AF599 and AF608, higher than normal thrust loads were required for seating and unseating the valves.
Although the cause L
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L has not yet been determined, the valve manufacturer recommended replacing the valves discs with a reinforced model. The licensee-decided to replace both valves with higher pressure class valves to provide additional design margin for over pressurizatien concerns and to sustain the thrust loads required to operate at the maximum system pressures. Another modification associated with this work was FCR 85-154 which, in part, removed the steam and feed rupture control system (SFRCS) Safety Actuation Signal for SG 1 solation Valves AF599 and AF608. This modification was intended to minimize the probability of inadvertent complete i-isolation of the Auxiliary Feedwater to the steam generator.
The following documentation was-reviewed for compliance with NRC.
requirements and licensee commitments:
Calculation No. 002. "Pioe Stress Analysis Auxiliary Feedwater System." Revision 01. February 9. 1988.
Although the model had been properly changed to account for the difference in valve weight and length, the original input data was incorrect with regard to the LOCA thermal expansion temperature.
Instead of using 120 F, the analysis was run at 90 F.
Subsequent licensee evaluation determined that this error did not cause the piping to exceed the design criteria.
MWO No. 2-87-1198 11 Closed May 15. 1990 Because of the modification to the H1 hanger, the drain line had to be changed to allow installation of the pipe cap. This change was originally requested under FPR 87 1198-016, May 3, 1990, and necessitated the installation of a new socket welded elbow and a short length of pipe. During the review of the implementation documentation, there was no record available as to what size of fillet weld needed to be made for the socket welded fitting. The licensee's response referred to MM09245 General Welding Procedure which specifies the size of the weld as 1.09 times the nominal wall thickness of the pipe. Using this approach requires the welder to perform a calculation using the nominal pipe wall in order to determine the weld size.
This same calculation would be required by the QC inspector in order to verify that the appropriate size weld was made.
Documentation was not available in the package to show what size weld should have been made or actually was made, it was the opinion of the NRC team that specifying the fillet weld size was a design function and should be specified in the design-
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documents as opposed to relying on the welders and QC inspectors to calculate the code required size.
Pine Support Calculation No. 6C EBD-14-H33. Revision 5. April 26.
1988, i
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This analysis was somewhat confusing in that both Revision 4 and i
Revision 5 of the calculation incorporated the same changes due to L
FCN 8022.
Revision 4 was a single statement conclusion that the addition of more stiffeners to the baseplate strengthens the plate and that the current ME035 computer run was conservative.
i Revision 5 was-a complete reanalysis with a new computer analysis
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using the new loads from Revision 3 and the configuration from Revision 4.
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The results of the Revision 5 calculation stated, "Since more stiffeners mean more strength to the plate and the current ME035 run is conservative, the baseplate and. anchor bolts are okay by inspection." The basis for this statement was questioned by the NRC team _since the current ME035 run used an applied X moment of 9.2828 inch kips whereas the Revision 5 X moment was reported to be 20.631 inch kips.
It was not-obvious how the analyst could
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state that the current ME035 run was conservative based on a
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moment increase of over 2.
This was considered an instance of undocumented engineering judgement by the NRC team.
Pioe Suonort Calculation No. 6C-EBD-14 H01. Revision 5.
December 24. 1987 with Addendum A dated April 17. 1990.
The purpose of the calculation was to " evaluate the support due to increase in load per Modification 87-1198." The evaluation consisted of the statement: " Comparing the existing calculation
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and STRUDL output, 6.7% increment in fz load will have no significant effect on structural integrity." However, in addition to a load increase there was also a configurational change to the
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support. The structural attachments to the base) late were shifted over to the edge of the plate. This change to tie configuration
was not addressed in the calculation nor in any other document in the modification package.
It was unclear to the NRC team if the analyst had reviewed this aspect and concluded that the change was not significant or if he did not realize that the support configuration had changed.
This is another example of an
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undocumented engineering judgement and is considered as a poor
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design control practice.
As a general comment, the NRC team found the support calculation packages to be somewhat burdensome. Of the approximately 250 pages associated with each of the above calculations only the last 50-60 pages were valid. The rest of the pages were voided or superseded but were still carried along with each revision to the calculation. Also, these calculations were recently turned over to loledo Edison by Bechtel and for some reason the cover sheet documenting this turnover was issued as Revision 0.
Now the cover sheet says Revision 0 and the calculation itself stands at Revision 5.
Although this didn't appear to confuse the Toledo Edison personnel it made it somewhat difficult for an outside auditor to follow.
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(3)
Modification 87-1243: "Reolace Reactor Coolant Pumo Seals with Different Desian*
This modification replaced the seals on the reactor coolant pumps (RCPs) with ones which were designed for better seal life-expectancy and to better withstand station blackout / Appendix R
concerns.
The new seals were supplied by Byron-Jackson, manufacturer of the RCPs.
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The modification was broken into five parts to allow for
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F installation on one seal at a time.
Four of the segments dealt l
with wiring changes as the new seals monitored seal temperature data through use of RTDs rather than with thermowells.
The final package provided for manufacture and installation of spool pieces j
to allow for piping differences between the old design and the new
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The licensee installed one seal package during the fifth I
refueling outage, and the other three during the sixth refuel.
The licensee originally planned to allow for use of either the old or the new seal designs, through swapping of the spool
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pieces.
However, during the outage, they decided to use only the
new !.eals to alleviate Appendix R concerns over seal life.
-The team reviewed the design package for the modification, which included the 50.59 safety evaluation. The basis for the change i
was a recommendation from the pump manufacturer and the nuclear steam system supplier (NSSS).
The instrumentation provided was non safety related, with the only safety concerns being seismic
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and the tie in to the non-lE electric containment penetration.
The team also reviewed the implementation package for the
modification, especially the electrical packages.
The team noted i
that the implementation packages were complete and well documented.
There were no problems with thia modification.
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Modification 87-1074:
" Remove Internals of Check Valves AF11 and AF12" This modification removed the internals of two lift check valves as part of the licensee's check valve maintenance program.
The check valves in question (AF-ll and AF-12) were on the inlet for the auxiliary feedwater (AF) pump bearing, the turbine bearing, and the turbine governor cooling coils.
This line normally was connected to the discharge from the AF pump, with an isolated connection to the service water system.
The check valves were provided to prevent backflow from the service water system into the AF system. The licensee evaluated the situation and
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determined that the only time that backflow could occur was during i
the turbine overspeed test. This was handled by adding a step to j
the overspeed test procedure requiring the operator to close the
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l manual gate valves prior to opening the service water valves.
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The team reviewed the modification, including the 50.59 safety.
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C, analysis and the appropriate testing procedures. The team found i
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the safety evaluation adequate and noted that the procedures had been changed to reflect the modification.
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(5)
Field Chanae Reouest 86-432:
"Uparade the Makeun System to
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Provide Better Feed and Bleed Caoability" This modification made a number of changes to the makeup system
and to the Power Operated Relief Valve (PORV) in order to upgrade the licensee's ability to cope with a beyond the-design basis loss t
of all secondary side cooling. Among these changes, the licensee moved the suction for the makeup pumps, when pulling from the
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r Borated Water Storage Tank (BWST), to the suction for the the High i
Pressure Injection (HPI) pumps. This provided the capability to
" piggy back" the suction of the makeup pumps to the discharge of the Low Pressure Injection (LPI) pumps.
The team reviewed the 50.59 safety evaluations and two supporting calculations from the modification 6asign package.
The majority of the design effort was performed by the NSSS vendor, B&W. The
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licensee performed a calculation to show that there was adequate NPSH for both the HPl pumas and the makeup pumps in the piggy-back.
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configuration. _ The team lad no problems with the portion of the modification reviewed.
(6)
Field Chanae Reauest 83-035: " Replace Oxvaen Monitor on the Waste Gas Surae Tank and Clean Waste Receiver Tank to Provide More Accurate Monitorina.
This FCR documented the continuing problems the licensee has had with the oxygen (0,) and hydrogen (H,) monitors on the waste gas surge tank (WGST) and the clean waste receiver tanks (CWRTs).
In j
1983, the modification was written to remove the two existing 0, analyzers and replace them with three 0 /H analyzers in the discharge of each tank line. The first sup,plement "0" was written
to install taps-and isolation valves in the lines from the WGST and CWRTs.
In 1984,. a supplement was written to add the taps and isolation valves from the CWRTs.
In 1985, the new monitors were added under supplement "2".
In late 1985 supplement "3" provided flow switches and indicating lights to the system.
In 1986 a Revision "B" was written to remove the system just installed and
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to replace it with a new redundant-system. Additionally, the location for the analyzers for the WGST was to be moved to a lower i
radiation area. (A revision "A" had been written, but was never implemented),
in late 1986, Supplement "5" removed the WGST analysis panel and replaced it with a "more reliable one".
In 1989, a second WGST system was added under supplement "6".
This supplement also revised the ranges of the instruments to provide
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greater accuracy.
system is a non-safety related radwaste system. However, The 0,/H' a technical specification limit on the amount of 0, in there is i
the WGST, which is monitored by the 0,/H, to an action statement.
analyzer, such that failure of the analyzer puts the plant in j
The team reviewed the design anc' implementation packages for this
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extended package. Although there was no technical problem with L
the modification, the team noted that this package exemplified the
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licensee's "old way" of handling modifications - to add continual
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supplements to an FCR package, rather than to close it and handle l
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new work as a new modification.
The team expressed the opinion to the licensee that additional work was necessary to see that
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these old FCRs were closed and-the 1987 modification program was
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implemented.
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(7)
Modification 87 1062 i
This modification replaced the emergency instrument air compressor
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(EIAC), installed a new centrifugal station air compressor (SAC),
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removed the diesel engine driven SAC for temporary use (not done),
t reduced the amount of moisture in the station air header, and
replaced service connection isolation gate valves with ball valves
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to reduce line leakage.
The modification intent was to improve the instrument and station air system reliability, capacity, and air quality.
The team reviewed the documentation, observed the installation,
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and had the following findings:
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(a)
EIAC and SAC Sizina Calculation The original EIAC and SAC sizing calculations were:
o Bechtel Calculation No. 14.4, " Instrument Air System,"
dated January 15, 1972.
o Bechtel Calculation No.14.2, " Station Air-Requirements," dated December 27, 1972.
t These calculations were not included in the TED document,
" System Description for Station and Instrument Air System,"
dated November 17, 1987, and were not used as design basis documents when sizing the new EIAC and SAC.
The present sizing of EIAC and SAC was based on conservative end user demand estimates. The subsequent.oversizing of the centrifugal SAC resulted in an operational problem that.
required installation of a modulated air pressure relief system to mitigate the problem.
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10 CFR 50.59 Safety Evaluation
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The performance characteristics of the centrifugal SAC were not evaluated in the licensee safety evaluation,
dated December 22, 1987, and resulted in operating problems as reported in the licensee Field Problem / Resolution Form, FPR 87-1062-103, dated E
April 28, 1989.
The permanent installation of the diesel driven SAC was recorded in facility Change Request, FCR 77-318,
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dated September 12, 1977, without a safety evaluation.
Subsequent safety evaluations documented in "USAR Section 9.3 Revision," dated December 12, 1986, did
not address the possible contamination of the breathing air supply due to the proximity between the diesel engine exhaust and air compressor. intake (TED t
System Review and Test Program Report Problem No. SIA-NRR-001, dated November 20,1985), and did not
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consider the design operating period for the diesel engine (lack of stored diesel fuel quantity
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requirement in operating procedure DB 0P 06521, dated
February 6, 1990).
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(c)
Post Modification Test
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.The team reviewed the Test Requirement Sheet, dated March 27, 1989, and the test results documented in Periodic Test Procedure DB SS-04013, completed on Jurs 22, 1989, and considered the capacity test to be adequate.
The team also reviewed the periodic air system moisture removal records, and had no adverse comments.
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(d)
As-Built Verification The team inspected the automatic drain system on the new centrifugal type SAC aftercooler (shown on Bechtel as-built drawing FSK-M-JBD-47 1, dated February 5, 1990), and observed a number of piping dimensional discrepancies between the as built and design.
The team also inspected J
the tubing system connecting to the pressure indicating
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controller (shown on.TED drawing J-1009, dated August 2, 1989), and observed that the installation was within acceptable tolerance. The licensee issued PCAQR No. 90 0417
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on May 16, 1990 to document the nonconforming conditions, (e)
Licensee Self-Initiated SSFI The licensee's self-initiated SSF1 of the station and instrument air system was conducted during May through
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August 1989. The team concluded that the effort was
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substantial af ter a review of the executive summary and some of the findings.
However, the team considered that the scope was compromised in the following areas:
o The lack of sizing calculations for the EIAC and the
SACS was not identified by the SSFI due to the TED Independent Safety Engineering (ISE) management-decision to let the production engineering groups evaluate the centrifugal air compressor operational problems. No plans for followup were indicated in the
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SSF1 report.
o The lack of an adequate safety evaluation for the l
diesel driven SAC was not observed due to the ISE's belief that it would be removed soon.
However, the
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diesel driven SAC continued to be needed due to the centrifugal SAC operational problems.
I Des)ite the above shortcomings, the SSFI did identify a num)er of significant safety issues, such as backu)
r.. umulators design function not demonstrated, tecinical
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errors and deficiencies in the operating procedure, and EDG load tables that Fad not accounted for all loads.
(f)
Licensee Corrective Actions During the last refueling outage (SRF0) in December 1988 time frame, the licensee's self-initiated SSOMI identified a number of system modification design control deficiencies, t
such as:
o Design packages not completed prior to commencement of the refueling outage.
o-Control valve venturi sizing calculations using non-conservative pressure drop assumptions, o
Applicable loads and timing sequence had not been factored into battery sizing calculations, o
incomplete review of design package, weak design i
interface, inadequate drawing update, and software control difficulties.
During the present ongoing refueling outage (6RFO), the
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licensee's self-initiated SS0MI further identified that
there was a lack of a safety evaluation for the elimination of the backup safety grade accumulator in the CCW instrument air system.
These design control problems, together with the team findings on the lack of sizing calculations for the EIAC and
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l SACS, and inadequate safety evaluations in some of the past modifications prompted the licensee to issue an Action Plan
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on May 8,1990, during the course of the NRC team J
inspection. The Action Plan will evaluate specific issues regarding Modification 87 1062, as well as sample six to ten modifications for each of the civil engineering, mechanical
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engineering, I&C engineering, and electrical engineering disciplines.
The team plans to followup on the licensee's Action Plan and the
PCAQR upon their completion and resolution.
This is an unresolved item (346/90010-02).
(8)
Modification 87-1280 i
This modification replaced the two Emergency Diesel Generator (EDG) air compressors (DACs), added a swing DAC, and installed aftercoolers and moisture removing devices. The purpose of this J
modification was to improve the EDG air start system reliability, capacity,.and air quality.
The licensee in a document (NED-90-20300, dated hay 15,1990),
committed to include this modification in its Action Plan (refer
to paragraph 3.b.(7)) as a part of the review for design change
control.
The team will perform a followup upon licensee
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com)1etion of the Action Plan. The team performed a general wal(down observation at the EDG and the new DACs, and had the following comT.ents:
o The EDG fuel oil strainers and filters were observed to have excessive leak:.
The licensee indicated that a new type of strainers and filters will be replacing the existing ones.
The team did not consider the present condition to be a fire hazard due to the distance between the leaks and the diesel engine block.
o The new installed DACs and the moisture removal equipment
were observed to be in good physical condition.
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A hold tag was placed at one of the EDG starting air system.
o pressure regulators. Air contamination resulted in the
regulator malfunctioning and was documented in a failure
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analysis report (NES-89-00215, dated March 28,1989).
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o The analysis identified that sludge, formed by water, oil, and rust caused improper regulatory function.
The team inspected the discarded and disassembled regulator internals, and observed shinny metal particles, which were not addressed in the failure analysis. Additional consideration for filtering out this particulate was discussed with the licensee.
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o The team reviewed the Deltech automatic drain valve (ADV)
vendor manual, and noted that a strainer and filter are recommended in the ADV system; however, none of these were observed during the team walkdown.
In discussion with the licensee, the staff reported the ADV experienced both i
failure to open and failure to close problems after
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installation.
To address the team's concern, TED staff stated that the ADV failure to open can be detected during
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frequent operational shift surveillance; and the ADV failure to close will be modeled as a continuous end user load demand in the Action Plan sizing calculations for all the air com)ressors. The team concuried with the licensee's.
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approac1.
(9)
Modification 88-0203 Modification of Diesel Generator Fuel til System
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Power supplies for fuel oil transfer pumps P1951 and 195-2 l
are routed through fire areas BN and Q.
In the event of a-i fire in these areas, the transfer pumps could become inoperable.
The purpose of this modification was to provide an alternate means of filling the emergency diesel generator
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(EDG) day tanks from the onsite fuel storage tank in the event of a loss of the dedicated pumps.
The NRC team reviewed the safety evaluation and found-it to
be acceptable. Approved drawing changes (implementation phase) were found to be properly filed with the applicable drawings.
Procedures, potentially affected by the modification, were found to have been reviewed and procedure
change requests issued where applicable.
(10) Modifiction 88-0251
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Installation of Steam Generator Inspection Openinas
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Steam generator secondary side fouling due to debris and deposit accumulation has resulted in increased pressure drop and subsequent power limitations.
This modifiction provided for the installation of two inspection openings through each
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steam generator shell and shroud. These openings permit
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visual inspection of the secondary side to monitor the extent of fouling and predict required maintenance.
The NRC
team reviewed the safety evaluation and found it acceptable.
The design of the openings and closure hardware was in accordance with B&W standard design.
C.
Electrical Modifications
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(1)
FCR No. 85-0063. " Install Ground Fault Protection f1E MCC)"
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Provide electrical ground fault protection and_ eliminate the lack of breaker coordination for safe shutdown'480 Vac essential power MCCs.
The Davis-Besse design cascades MCC to MCC through molded
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case breakers. The breakers did not provide adequate coordination and ground fault protection to prevent the loss of Appendix R associated circuits. This modification installed a high resistance ground system, removed six (6)-
breakers and hard wired the MCCs together.
The documents reviewed included the safety evaluation, finalized design, design change documents, design reviews,_
seismic adequacy of the installation, ten (10)
implementation work orders, and testing. The team concluded the following:
the design was adequate; the modification was being adequately controlled; and the modification was being adequately implemented and tested.
(2)
RFM No. 87-1045. " Station Battery Dicharoe/0vercharoe Detection"
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Provide undervoltage (VV) and overvoltage (OV) detection to ensure Technical Specification operability of the station batteries for a discharge to 110 Vdc and an overcharge to 150 Vdc. This modification installed General Electric Model 196 relay meters on each of the de distribution MCCs.
The documents reviewed included the safety evaluation, finalized design, design change documents, design reviews, seismic adequacy of the installation, two. (2) implementation work orders, testing, and observation of the actual installation. The team concluded the following:
the design was adequate; the modification was being adequately controlled; and the modification was being adequately implemented, except for the setpoint calibration tolerances.
Design engineering recommended an UV setpoint of 12611 Vdc and an OV setpoint of 148+0,-2 Vdc. However, the relay setting change notice did not specify a setpoint tolerance.
The plant assigned a setpoint tolerance of 13 Vdc which was based on a relay meter accuracy of 12%. As a result, there exists the potential for the station battery Technical Specification to be exceeded.
in addition, the measuring and test equipment (MTE) accuracy was not considered in the setpoint which adds additional setpoint uncertainty. Design engineering was not cognizant of the plant's implementation of the setpoint tolerance.
This is considered an unresolved item (346/90010-03) for the licensee to review the implementation of electrical
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is interfacing with design engineering so that the designer l
is cognizant that the intent of the design is being l
fulfilled.
The as-left llV setpoint was 125.7 Vdc and 148 Vdc for the OV setpoint for MCC No. 2 2N, and 125.9 Vdc and 148.3 Vdc respectively, for MCC No. 2 2P.
For the above, the i
setpoints with the additional MTE uncertainty are within their respective Technical Specification limit.
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(3)
RFM No. 88-0254. " Isolation for 4160 VAC C1 Bus" Circuit ICAC103A was not protected from a fire in the control room or cable spreading room. This modification provides for the isolation of the 4160 VAC Cl bus
undervoltage trip bypass circuit from the control room by the installation of a local-remote control switch and d
redundant fusing.
Loss of de control power as a result of a j
fire would prevent Breaker Nos. AC110 and ABDCl from i
tripping.
Failure to trip these breakers would prevent the emergency diesel generator automatic starting circuit to
energize as designed.
The documents reviewed included the safety evaluation, finalized design, design change documents, design reviews,
-J seismic adequacy of the installation, four (4)
implementation work orders, testing, and observation of the
actual installation.
The team concluded the following:. the design was adequate; the modification was being adequately
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controlled; and the modification was being adequately implemented and tested.
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D.
Instrumentation and Control Modifications
(1)
FCR No.79-196. " Add Flow Transmitter and Comparator on Stator Coolino Water System" The addition of the flow loop was at the recommendation of Ceneral Electric (GE) Technical Information Letter (Til-881). Water cooled generators of GE manufacture are
supplied with protective sensors and circuitry to protect the generator stator winding upon failure of the cooling water system.
Certain system failures may go undetected by both the low pressure and high temperature sensors.
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complete system blockage downstream of the pressure sensor would go undetected. A low flow sensor would detect this situation.
The documents reviewed included the safety evaluation, finalized design, design change documents, system
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S procedures, and alarm procedures. One alarm response procedure (DB-0P 02016) needs the low flow setpoint changed from 419 gpm to 420 gpm. This has no effect on the operator's response to this alarm. The licensee was correcting this error. The team concluded the following:
the design was adequate; and.the modification was being adequately controlled.
(2)
FCR No. 85-0083. "Two Out of Three Trio Scheme for Main Steam Reheater (MSR) Hioh level Trio Convert the original design one out_ of one MSR high level tri). logic to a two out of three logic. The original main turaine protection system used one level switch to trip the turbine. This modification adds two additional level switches and logics to increase the reliability of the turbine trip system and to help eliminate unnecessary turbine trips due to spurious spikes in MSR drain level.
The documents reviewed included the safety evaluation, finalized design, design change documents, design reviews, five (5) implementation work orders, cold and hot electrical checks, post modification testing, and observations of the actual installation.
The team reviewed Test Procedure TP 870.17, "MSR High Level Trip Acceptance Test." The procedure contained the necessary elements of a purpose, references, precautions and limitations, prerequisites, easy to follow procedure steps, restoration, and acceptance criteria that makes for a good test procedure. The test also verified each logic combination and provided test overlap into unmodified portions of the circuit. The team also noted that the implementing work requests contained elements similar to the test procedure and the work request activities log had a sign-off that the scope of work activities had been discussed with the craft personnel.
That the packages contained good work elements and the nature of the work was communicated to personnel perft.. ming the work is considered a strength by the team.
The team concluded the following:
the design was adequate; the modification was being adequately controlled; and the modification was being adequately implemented and tested.
(3)
FCR No.1208. " Fuel Oil Storaae Tank level Indication" Provides a high level local alarm to prevent tank overfilling. This modification replaces the existing Fuel Oil Storage Tank (FOST) low level switches with new icvel probes.
The new ) robes will still provide low level annunciation in tie control room.
in addition, local level
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indication as well' as high/ low level indicating lights will be provided. The low level alarm will be set at 38,000 gallons (same as original setpoint). This is well above the
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Technical Specification limit of 32,000 gallons.
The documents reviewed included the safety evaluation, finalized design, design change documents, six (6)
implementation work orders, draft test procedure, probe vendor manual, and observation of the actual installation.
Operations is currently reading the fuel oil level the same way as in the past (using a measuring stick) as the level indicators have not been powered up. The team reviewed draft Test Procedure No. DB MI 04303, " String Check of 26 B-ISL4891, Emergency Diesel Generator Fuel Oil. Storage Tank 1".
The procedure was to perform a vendor recommended two (2) point calibration. However, the calibration points chosen by the licensee encompassed the high and low level-setpoints and not a calibration point near 32,000 gallons (Technical Specification limit). The range of the level indicator is 30,000 to 40,000 gallons. The level probe is a capacitance ty)e arobe that was custom designed by the vendor to matti t1e non-linear tank curve.
The operators will be using the indicator to determine Technical S)ecification complicity. Therefore,-it is the opinion of t1e team that the probe calibration should encomaass.the 32,000 gallon indication point. The licensee ac(nowledged the team comments and the team has no further questions on this item at this time. The team concluded the following:
the design was adequate; the modification was being adequately controlled; and the modification was being adequately implemented.
E.
lemocrary Modifications (1)
General Ths licensee had adequate control over the temporary modification (TM) program. As of the time of this inspection, there were approximately 30 TMs open, including a number which were directly related to the outage. Nearly two-thirds of these TMs were less than a year old.
The licensee stated that they had a program underway to reduce the number of TMs still further. Their stated goal was to have no more than fifteen TMs by the end of the outage, with only five of these on safety related (or "Q") systems.
The team reviewed the safety evaluations, or basis for not performing a safety evaluation, for all the open temporary modifications.
In all cases, the safety evaluations were acceptable. For those TMs where no safety evaluation was performed, the team agreed with the justification as to why
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a safety evaluation was not required.
(2)
Temporary Modification 89-0051:
"Reclate Service Water Valve 5068 with a Laraer Size Valve" This particular TF was written to replace the one-inch valve on the Service Water 2nlet to the Hydrogen Dilution System with a two inch valve. This was done since the installed valve was found to have a bent stem, such that it was inoperable, and there was no readily available replacement. The replacement valve was removed from the discharge of the Service Water cooling line from the room cooler for the Decay Heat Coolers. The valves were procurred to the same requirements and were identical type valves, except that the replacement valve was larger. The licensee performed calculations to determine the need for new supports, and the effect on the valve operator, and took these into account in the installation.
The team reviewed the calculations for the supports and the settings for the valve o>erator, and had no problems with them.-
The team also reviewed tie safety evaluation and found it to acceptably document the consequences of installing the larger valve in the hydrogen dilution line.
However, the inspector questioned how the valve was obtained that was used in the TM. The TM package stated that the valve had been removed under another TM (89 0050), however the licensee confirmed that this TM had not been issued. The licensee stated that the valve was removed from the Service Water System under a maintenance work order (MWO 2145). The licensee further stated that the room cooler (C3103) was presently inoperable, and had been for a "long time". Since the room coolers are safety-related components which are required to work following a LOCA (in order to ensure that the ECCS components continue to function), the team questioned the acceptability of leaving this cooler inoperable.
The licensee provided several calculations to show that temperaturer in the room would remain below the qualification temperatures following an accident with both the 31-03 cooler inoperable and one of the two room coolers in the adjoining pump room inoperable.
The calculations had not been performed considering two of the three coolers inoperable; however it was possible to extrapolate to show that the plant was in a safe condition.
Furthermore, the licensee stated that during the outage they had run both Decay Heat Coolers with the room cooler inoperable, and the room temperature had remained below ~105 degrees.
Based on this, the team's concerns in this area were resolved.
4.
Potential Condition Adverse to Ouality Report (PCA01 Review SDER
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The licensee's quality program utilizies a PCAQ report to document potential quality issues and to initiate coriective action.
The team reviewed select PCAQ reports in an attempt to evaluate the-licensee's
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technical support effort in this area.
(a)
90 0287. " Incorrect KVAR Meter Model Installed on C3715 for EDG 1-
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1" The licensee identified a discrepancy in the indicated KVARs and what the actual KVARs should be for a replacement KVAR meter recently installed on EDG l-1.
The KVAR meter is locally mounted along with other meters that can be used to determine generator load and is used during local tending of the EDG.
The operators monitor normal EDG loading by a KW meter in the control room. Tha meter calibration was verified and found acceptable at the time of installation. The meter was calibrated over its entire range using a single phase source e.o the calibration method was similar to the supplier's method.
During troubleshooting, the licensee injected three phase power into the current and potential transformer secondary circuits to simulate actual operating conditions.
The actual VARs clid not match the indicated VARs.
The licensee determined that the supplier had sent the wrong KVAR model. The team reviewed the
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bill of material and concluded the licensee had ordered +he correct KVAR meter (Model No. 103321AXGJ1JAZ).
The. material issue ticket verified that meter Serial No. 43192 had been issued for the re)lacement KVAR meter, this was the meter that was received from tie su) plier.
However, the meter model number had been covered up )y the supplier's stock label.
The licensee peeled
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back the supplier's label and discovered Model No. 103282AMGG7EAA had been sent by mistake.
This model contained an internal phase shifting transformer where the plant design uses an external-phase shifting transformer.
The^ single phase calibration will indicate the same for either type of meter. The KVAR meter on EDG 1 1 was replaced with the correct model and the licensee verified the correct meter'was installed on EDG l-2.
The licensee indicated they may revise their calibration method
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for these types of meters to ensure they are responding correctly l
to three (3) phase parameters. The operability of EDG l-1 was not affected by the wrong KVAR nieter and the team has no further
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questions on this item.
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(b)
90 0200 " Fault Protection for Containment Electrical Penetration Assemblies (EPA)"
The licensee identified that four (4) Class IE circuits and twenty-one (P.1) non-Class lE circuits that lassed through
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containment EPA's did not have adequate baccup fault protection.
The Davis-Besse Updated Safety Analysis Report (USAR - Section 8.3.1.2.20) states, in part, "... the electrical system design
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)rovides both' primary and' backup fault protection for the circuits (RG)g fed through all electrical penetrations." Regulatory Guida
)ein No. 1.63-1973, " Electric Penetration Assemblies in Containment Structures for Water Cooled Nuclear Power Plants" (which endorses IEEE Standard No. 317-1972), defined the NRC's position on how to demonstrate compliance with General Design Criterion (GDC) No. 50, " Containment Design Bases".
The RG stated
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the following in Paragraph Cl:
"... The electric penetration assembly should be designed to withstand, without loss of mechanical integrity, the maximum possible fault current vs. time conditions (which could occur because of single random failures of circuit overload protection devices) within the two leads of any one single phase circuit or the three leads of any one three-phase circuit.
Incorporating
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adequate self-fusing characteristics within the penetration conductors themselves constitutes an acceptable design ap) roach.
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Where self+ fusing characteristics are not incorporated, tie circuit overload protection system should conform to the criteria of IEEE Standard 279-1971... "
i The licensee committed in the safety analysis report to the following exceptions to RG 1.63:
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IEEE 317-1971 is used instead of IEEE 317 1972.
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Paragraph C.1 is not couplied with, as the penetrations do not have self fusing characteristics but are designed to withstand the short circuit conditions. Also, the overload protections of non Class lE systems do not comply with IEEE 279-1971.
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Further, Final Safety Analysis Report (FSAR) revisions have consistently stated that non Class lE circuits are exempted from meeting the criteria of IEEE 279 1971.
The licensee has reduced the list of non-Class IE penetrations that lack backup protection and are energized during normal plant operation to the following:
Containment-Sump Pump 1 1
RCP Primary A.C. Oil lift Pump 1-1
RCP Primary A.C. Oil Lift Pump 2-2
RCP Cooling Water Return Valve (1-1)
RCP Cooling Water Return Valve (2 2)
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Containment Sump Pump 1 2
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RCP Primary A.C. Oil Lift Pump 1-2 8)
RCP Primary A.C. Oil Lift Pump 1-2
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RCP Cooling Water Return Valve (1-2)
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RCP Cooling Water Return Valve (2 1)
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The licensee performed Calculation No. C-NSA-059.01-012, " Effects
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of Fault Current on Electrical Penetration Assembly (EPA)
Leakage". The calculation purpose was to assess the potential for phase to ground faults to damage the exposed insulating material
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in No. 8 AWG wire size Amphenol EPA modules which are installed on
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No. 12 AWG wire size circuits. The calculation determined the No.
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12 wire would exhibit "self fusing" characteristics and open prior j
to the EPA module sustaining any damage. Thus, containment
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integrity would be maintained and backuo protection was not
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required.
j 1he licensee is environmentally qualifying the equipment connected to the four (4) Class IE penetrations. The. reclassification is to
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be completed prior to restart from refueling cycle six (6).
By meeting the environmental qualification (EQ) requirements, the licensee is taking credit that accident conditions will not induce j
electrical faults in Class IE equipment. As a result, specific backup fault protection devices are not required.
The licensee provided the team a package of supporting documentation, Electrical Penetration Assembly Integrity
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Protection Report and Calculation No. C-NSA-059.01-012..This
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information has been provided to the Office of Nuclear Reactor
Regulation (NRR) for further review. This is considered an i
unresolved item (346/90010-04) for the NRC to determine if environmentally qualifying the IE circuits and taking credit for the "self-fusing" characteristics of the wire is acceptable to
assure the mechanical integrity of the EPA modules under fault current conditions, a
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5.
Procurement and Dedication of Commercial Grade Components The team performed a brief overview of the licensee's procurement program with respect to the dedication of commercial grade items for
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nuclear applications.
The licensee has implemented a procurement program w1ich meets the guidelines of EPRI NP-5652 " Guideline for the Utilization of Commercial Grade items in Nuclear Safety-Related i
Applications (NCIG 07)".
This program was initiated in March 1989, and
was still in the development process at the time of the inspection.
In accordance with the guidance found in NRC letter SECY 90 076 " Inspection and Enforcement Initiatives for Commercial Grade Procurement and Dedication Programs", dated 4/11/90, no programmatic inspection in ti,e procurement area was performed during this inspection.
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The team attended a presentation by the licensee in regard to their program status and reviewed a " Procurement Assessment Report" prepared by the licensee's Quality Assurance function in December of 1989. The
team also reviewed procedures EN DP-01023 "Commercirl Grade Dedication" and EN-0P-00070 " Procurement".
The licensee appeared to have a good
awareness of the issues involved in the procurement and dedication of commercial grade items.
In regard to the QA review of the procurement process, the team found
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that the QA organization had performed a thorough audit and identified areas where additional improvements could be made.
6.
'Mnresolved Itemi
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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.b, 3.c, and 4.
7.
Exit Interview
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The team met with licensee representatives denoted in Paragraph I during and at the conclusion of the inspection on May 18, 1990. The team summarized the scope and results of the inspection and discussed the
likely content of this inspection report.. The licensee acknowledged the information and did not indicate that any of the information disclosed
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during the inspection could be considered proprietary in nature.
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