IR 05000346/1986019
| ML20214X049 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 12/03/1986 |
| From: | Danielson D, Fair J, Yin I NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20214X001 | List: |
| References | |
| 50-346-86-19, CAL-85-13, IEB-78-12, IEB-79-07, IEB-79-7, NUDOCS 8612100518 | |
| Download: ML20214X049 (24) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Report No. 50-346/86019(DRS)
Docket No. 50-346 License No. NPF-3 Licensee: Toledo Edison Company
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Edison Plaza
, 300 Madison Avenue Toledo, OH 43652 Facility Name: Davis-Besse Nuclear Power Station, Unit 1 Inspection At: Davis-Besse Site, Oak Harbor, Ohio Bechtel Power Corporation, Gaithersburg, Maryland (Bechtel)
Teledyne Engineering Services, Waltham, Massachusetts (TES)
MPR Associates, Inc., Washington, D.C. (MPR)
Inspection Conducted: July 15-18, August 27-28, September 23-24, '
October 15-16, and 29-30, 1986 at the site
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July 30, and October 22-23, 1986 at Bechtel September 9-11, 1986 at TES ctober 23, 1986 at MPR
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Inspectors: . T. Yin ih 6 Ulite
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b J. R. Fair I O (July 30, 1986 only) Date M '
Approved By: i. Danielson, Chief Materials and Processes Section Date Inspection Summary Inspection on July 15 through October 30, 1986 (Report No. 50-346/86019(DRS))
Areas Inspected: Special, announced inspection of auxiliary feedwater pump turbine steam supply (AFPTSS) piping tests (92701); the Facility Change Request (FCR) system (92701); implementation of Region III (RIII) Confirmatory Action Letter (CAL) 85-13 actions required prior to plant restart (92703); TED actions on Licensee Event Reports (92700); TED actions on IE Bulletins (92703);
TED project responses to the QA design control audit findings (92701); and review of pressurizer PORV piping (92701).
8612100S18 861203 PDR ADOCK 05000346 G pop
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Results: Of the areas inspected three violations were identified (TED failed to ensure that design was adequately controlled by engineering contractors -
Paragraphs 10 and 12; Bechtel procedures did not contain requirements for combining support loads and the Bechtel staff failed to follow design verification procedures - Paragraph 9; TED failed to document nonconformances in PCAQRs - Paragraph 14).
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DETAILS Persons Contacted
' Toledo Edison Company (TED)
J. Williams, Jr. , Senior Vice President, Nuclear
- P. C. Hildebrandt, General Director, Nuclear Facilities Engineering Division (NFED)
- S. J. Smith, Assistant Plant Manger
- J. Wood, Systems Engineering Director, NFED +
- P. H. Straube, Senior Engineer D. J. Harris, QA Engineer
- T. S. Swim, Civil / Structural Engineering Manager
- B. J. Carrick, Design Engineering Director, NFED T. J. Bloom, Senior Licensing Specialist
- J. C. Sturdavant, Licensing Specialist C. A. Rainey, Structural Engineer
- A. G. Weedman, Manager, Engineering Assurance
- J. E. Moyers, Quality Verification Manager W. T. O' Conner, Assistant Plant Manager, Operations <
M. L. Murphy, Acting Mechanical Engineering Manager J. C. Buck, Supervisor, Quality Verification L. P. Zarkesh, Engineering Support Coordinator
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J. Dunne, Nuclear Safety Analysis Supervisor i S. C. Jain, Nuclear Engineering Director J. E. Soares, Management Staff R. Gardomski, Associate Nuclear Engineer E. M. Salowitz, General Superintendent, Outage and Program Management
- B. Klein, I&C Engineer
D. E. Kazimir, I&C Engineer T. J. Zunk, Technical Engineer
- Johnson, Design Engineering Manager E. Bain, QA Auditor
- L. O. Ramsett, QA Director
- R. F. Peters, Licensing Manager
- L. Storz, Plant Manager
- D. B. Amerine, Assistant Vice President, Nuclear H. Stevens, Systems Engineer Bechtel Power Corporation (Bechtel)
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W. C. Lowery, Project QA Engineer S. A. Bernsen, Manage of QA D. Gill, Project Quality Engineer P. Carrato, Civil Staff M. Franzen, Assistant Civil Group Supervisor M. S. Wasserman, Mechanical Supervisor
- N. Kalyanam, Assistant Project Engineer N. Tolani, Senior Engineer C. H. Abutaa, Senior Engineer
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V. Marathe, Project Engineer C. Querry, Engineer J. M. Ogle, Civil Supervisor J. Hook, Civil / Structural Engineer B. Mohip, Civil / Structural Engineer Teledyne Engineering Services (TES)
G. Moy, Principal Engineer L. B. Semprucci, Principal Engineer D. F. Landers, President MPR Associates, Inc. (MPR)
R. M. Weiner, Supervisor Engineer P. Kasik, Engineer H. W. McCurdy, Engineer U.S. Nuclear Regulatory Commission, Region III (RIII)
A. B. Davis, Deputy Regional Administrator C. J. Paperiello, Director, Division of Reactor Safety D. H. Danielson, Chief, Materials and Processes Section R. W. DeFayette, Projects Section Chief
- N. Jackiw, Projects Section Chief
- P. M. Byron, Senior Resident Inspector
- D. Kosloff, Resident Inspector
- L. Kelly, Project Engineer (NRC-NRR)
- I. T. Yin, Senior Mechanical Engineer Denotes those attending the management meeting at the site on October 30, 198 * Denotes those attending the management exit meetings on September 24, October 29 and 30, 198 . Licensee Action on Previous Inspection Findings (Closed) Unresolved Item (346/83017-02): During the NRC inspector's previous review of small bore (S/B) piping suspension systems, unusual piping arrangements were identified. Consequently the licensee committed to evaluate the generic effects of interactions between snubbers, loose guides, tight guides, rigid restraints, and rigid supports. On August 28, 1986, the NRC inspector performed piping walkdown inspections in congested S/B areas inside the containment "0" ring locations including steam generator No.1-1 on Elevation 565'-0", reactor coolant pump No.1-1-1 on Evaluation 565'-0', and top of the pressurizer on Elevation 623'-0".
No similar piping arrangement problems were observed. No violations or deviations were identified. This matter is considered resolve '
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s (0 pen)}0 pen Item (346/85035-01): TED committed to conduct confirmatory tests on the AFPTSS riping sy: tem follow the completion of modification and prior to Mode 3 operation. The NRC inspector reviewed the following TED test and inspection procedures:
- TP 850.99, "AFW-MS Supply Line Vibration Test, FCR 85-143 and FCR 85-163," March 22, 198 * IP-M-006, " Thermal Expansion and Dynamic Testing of Main Steam Supply to the Auxiliary Feedwater Pump Turbines," March 7,198 The NRC inspector had no adverse comments relative to the scope and control measures; however, he emphasized to the TED staff that additional surveillance / inspection should be placed at specific segments of AFPTSS line where high thermal stress is expected (NRC Inspection Report No. 50-346/86004, Paragrapii 6.b). The system test is presently scheduled for November 1986. The NRC plans to observe portions of the above tes (Closed) Unresolved Item (346/85035-03): The present TED nonconformance evaluation for weld deficiencier dasigned to AISC specification (7th Edition) did not include AWS D1.1 eiteria for minimum weld sizes corresponding to base material thicknesses. The NRC inspector reviewed the TED response contained in an intercompany memorandum, File 0093, T-0294, January 24, 1986. Similar findings were identified during a NRC Construction Appraisal Team inspection at Washington Public Power Supply System WNP The TED proposed resolution for Davis-Besse is similar to the resolution of the issue at WNP The NRC inspector considered the matter resolved based on the WNP 2 resolution being accepted by the NRC staff in WNP 2, SSER 5,
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Section 1 (Closed) Open Item (346/85035-05): The TED FCR and MWO systems require further evaluation and improvemen The open item covered three issues:
(1) Procedures at the time of the NRC inspection required that QC
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inspections including witness and hold points be determined by the QC/ Code reviewe The NRC inspector reviewed TED QC Instructions, QCI 3103,
" Maintenance," Revision 13, May 22, 1986, which was recently revised to require that all safety-related MWO work including installation of new hangers and WRs and modification, repair, and readjusting of any exiting supports should require mandatory QC inspection / verification.
. m ,(2) Procedures.at the time of the NRC inspection required that QC
. inspections for hanger configuration including location, dimension, and orientation not be initiated until all work within the MWO has been complete _ _ _ _ _ - . _
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The NRC inspector reviewed the Intracompany Memorandum from TED QA Manager to TED Licensing Manager, HA 86-0384, "LCTS Open Itam 1893, Hanger Inspections," March 31, 1986, which acceptably ensures that inspections for individual supports are conducted in a timely manner, as soon as the installation or rework has been completed by the craf ,
I (3) At the time of the NRC inspection, the FCR system in use was !
cumbersome and needed revision to improve its effectivenes !
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(a) To ensure effective interim corrective measure prior to plant restart, the NRC inspector performed followup review I of the TED closecut of some of FCRs to satisfy actions set l forth in the RIII CAL Item 1.a(4). See Paragraph 5.6 for detail (b) The NRC inspector reviewed the present TED improved control measures for streamlining the FCR process. See Paragraph 6 for details.
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e. (Closed) Violation (346/86004-01A): The Bechtel staff did not report to TED, as required by procedure, that several piping systems and supports did not meet FSAR conditions. The NRC inspector reviewed
, the TED response letter, Serial No. 1-639, June 9, 1986, and l considered it acceptable. During a followup inspection at Bechtel on July 30, 1986, the NRC inspector reviewed the following documents:
l * EDPI-4.61-11, "Nonconformance Reports and Supplier Deviation Reports," Revision 4, July 10,198 * Bechtel interoffice memorandum, " Generic Corrective Action, NRC Inspection Report No. 50-346/86004-01a," May 23, 198 * Bechtel personnel training record * Closeout records of transmittal letters from Bechtel Project Engineer to TE No deficiencies were identified during the above revie This matter was close f. (Closed) Violation (346/86004-018): The number of times the PORV lifted exceeded the TED procedure limitation. The TED staff informed the NRC inspector that the plant operators had been tracking the number of PORV lifts, and counted a total number of 17 lifts. Based on a subsequent review this was revised to 21 lift The TED review determined that the lack of procedure requirements to account for PORV lifts having dynamic transient loading effects on the piping system was a deficiency within the operation program. Corrective
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actions including component modification and evaluation of the operational restraints due to cyclic loading were conducted by TE The matter is close . . . . . _ , .. . .
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g. (0 pen) Unresolved Item (346/86004-02): Abandoned unfilled anchor i bolt holes and voids were identified during the TED repair of
! restraints pulled away from wall. The NRC requested TED to assure n that this is not a generic problem. TED responded in a letter to E
RIII, Serial No. 1-644, June 8, 1986, stating sufficient design margin exists in the design of Davis-Besse anchor bolts and support
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baseplate anchor systems to ensure adequate pullout strength even F for extremely adverse postulated unfilled anchor bolt locations; therefore, there was no need to inspect for additional cases of r abandoned bolt holes behind baseplates. On August 28, 1986, the NRC i inspector met with TED NFED Group Director and Engineering Director
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at the site, and discussed the NRC's technical basis for concluding additional inspections should be conducted. TED committed 9 to perform additional sample inspections. Both the NRC and TED have concluded that the existing installations meet piping system
, operability criteria (based on the results of extensive piping
, suspension system reinspection and the evaluation of identified m deficiencies), and that the sample inspections can be completed
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subsequent to plant restart but prior to the completion of next refueling outage.
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h. (Closed) Violation (346/86004-03): TED did not incorporate the '!
latest safety / relief valve design requirement into the pressurizer
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relief system operating procedure. Since the PORV loop seal was '
romoved during a recent system modification, the specific concern p no longer exists. To determine if similar problems exist in other
- systems, TED initiated a document review of piping stress analyses
[ performed by Bechtel and its contractors. The review is documented I in a Bechtel letter to TED, BT-16968, July 23, 1986. The NRC inspector reviewed the letter including the attachment to the letter,
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, at Bechtel en July 30, 1986, and commented that the operational
[ restraints had not been compared with plant operational procedures and records. During the followup inspection at the site on August 27, E 1986, the NRC inspector reviewed the design and operation comparison
[ and had no adverse comments. See Paragraphs 10 and 11 for details
'! on questions identified relative to the reactor coolant makeup and purification system operatio V
[ i. (Closed) Unresolved Item (346/86004-04): The NRC inspector reviewed d the TES reports on the effects of the PORV lift The following are RIII concerns and TES responses: '
i (1) The maximum transient load locations where piping restraint i damage was observed, did not correlate with the TES analytical f prediction.
b The difference was probably due to support installation
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deficiencies, repeated dynamic transients during early PORV e tests, and component as-built deviations in some localized g areas (Paragraph 13.b).
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(2) Review of the report reference lists revealed what appeared to be an analytical basis that was either preliminary or interim in natur The NRC inspector reviewed the key elements in the PORV test data contained in EPRI Report No. NP-2628-SR, "PWR Safety and Relief Valve Test Program - Safety and Relief Valve Test Report,"
December 1982, and found it closely matched the EPRI Interim Report, "Research Project V102," April 1982, which was used by TES to form its analytical basi i (3) Several support design loads, based on the present dynamic transient without loop seal (at 400 F subcooled water condition),
were considerably higher than the original design with loop sea The present support loads should have been bounded by the original desig The higher support loads were attributed to the actual PORV nozzle area being potentially larger than design. However, the absence of a water hammer analysis (400 F subcooled water) had resulted in significantly underestimating support loads upstream of the PORV and justified support load increases between the
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PORV and the quench tan (4) The support design load combination for rigid restraints should be the larger of (Thermal + Weight + SSE) or (Thermal + Weight +
Blowdown). The snubber design load should not consider the thermal and weight loads. Some of the TES support design loads could not be as verified using the above criteri See Paragraphs 12 and 13 for NRC inspection followup.
l (5) The PORV discharge modes of operation include:
(a) PORV opens on 2450 psig saturated stea (b) PORV opens on 2450 psig saturated steam followed by a transition to subcooled wate (c) PORV opens on 2450 psig 640 F subcooled wate (d) PORV opens on 2450 psig 400 F subcooled wate The present design is based on mode (d). Based on the comment stated in (3) above, it is not clear that the mode (d) support loads will bound all modes of operation. See Paragraph 13.c(2)
for NRC inspection followu j. (Closed) Unresolved Item (346/86004-05): Issues relative to the lack of a formal design interface control between TED, Bechtel, and Grinnel _
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(1) The TED staff informed the NRC inspector that all Grinnell original hanger calculations had been shipped.to the site. TED will develop a program to review these calculations to assure compliance with design procedure RIII followup will be conducted-after plant restart, but prior to the next refueling outag This is an unresolved item (346/86019-01).
(2) The NRC inspector reviewed the TED audit report on design control / interface. See Paragraphs 7 and 8 for detail (Closed) Violation Item (346/86019-06AB): Inadequate design control and reconciliation to the as-built and as-operated conditions for the purification letdown and pressurizer relief piping system See Paragraphs 11 and 13 for NRC inspection followu * Licensee Action on NRC IE Bulletins (IEBs)
, (Closed) IEBs 78-12,78-12A, and 78-128 (346/78012-BB, 346/78012-1B, 346/78012-28): IEB 78-12B, " Atypical Weld Material in Reactor Pressure Vessel Welds," March 19, 1979. The NRC inspector reviewed TED respons3 letter to RIII, Serial.No. 1-74, May 31, 1979, and
'had no adverse comment The NRC inspector also discussed the matter with the TED responsible engineer at the site on September 24, 1986. The TED investigation concluded that B&W's shop practice virtually eliminated the possibility that a fabricated vessel might contain off-chemistry weld material (Closed) IEB 79-07-(346/79007-BB): IEB 79-07, " Seismic Stress Analysis of Safety-Related Piping,". April 14, 1979. The NRC
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inspector reviewed TED letter to RIII, Serial No. 1-62, April 24, 1979, and had no adverse comments. The results of investigations conducted by the A-Es and the subcontracted engineering organizations, indicated that piping analysis computer programs did not use
- algebraic summation of the seismic loads in eitner horizontal or vertical direction ' Licensee Action on Licensee Even Reports (LERs) (Closed) LER (346/86013-LL): "Two Piping System Stress Problems Exceed USAR Allowable Stresses," reported on April 14, 1986. The
NRC inspector review of the problem was documented in NRC Inspection Report No. 50-346/86004, Paragraph 4.b. TED actions taken to resolve the issue were considered acceptabl "Small Break LOCA Analysis Assumptions
' (Closed) LER (346/86028-LL):
4 Potentially Non-Conservative," reported on August 8, 1986. The NRC
inspector generally discussed the issue with TED Director of Nuclear Engineering, and considered the planned corrective measures adequat During a followup inspection at the site on October 15, 1986, TED
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presented the results of the specific evaluation, and concluded that B&W's incorrect design assumptions had not caused unsafe system i
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operation. TED also performed a broader review of all B&W topical report assumptions, including: (1) Small break and large break LOCA, (2) Fuel and core design characteristics, and (3) LOCA effects on reactor vessel internals. Preliminary results indicated no operability concerns. TED's Independent Safety Engineering Group will finalized the studies prior to the next plant refueling outag This LER is considered close (Closed) LER (346/86037-LL): " Design Assumption Not Incorporated in Plant Operations," reported on September 12, 1986. This is a violation (346/86019-02A) item discussed in Paragraph 10 of this report. TED performed corrective actions during the NRC inspectio See '-3ragraph 11 for closeout detail (Closed) LER (346/86038-LL): " Potential Design Deficiencies in Auxiliary Feedwater Pump Turbine Governor," reported on October 1, 1986. The NRC inspector reviewed the previous repair and maintenance records for the device. The new Woodward PGG governor supplied by Terry Company and installed on AFPT 1-2 on January 1985 and installed on AFPT 1-1 during the present outage represented design and quality improvements. The speed settling bushing in the governor was replaced to meet the Technical Specification pump response time. The speed change motor was improved and tested to prevent oil mists from getting into the motor internal . Implementation of RIII CAL 85-13 Action Items As a result of meeting conducted at the site on October 9, 1985 (NRC Inspection Report No. 50-346/85033, Paragraph 4), RIII CAL 85-13 was issued on October 17, 198 The licensee's implementation of the actions set forth in the CAL was reviewed by the NRC inspecto The status of CAL Item 1 (action items prior to plant restart) was as follows: Item 1.a(1), 1.a(2), and 1.a(3) (Closed)
See NRC Inspection Report No. 50-346/86004, Paragraph Item 1.a(4) (Closed)
The FCRs that could impact safety-related piping system operability are listed in NRC Inspection Report No. 50-346/85035, Paragraph An update of this list is as follows:
(1) FCR Nos.77-213 and 77-398: No Maintenance Work Orders (MW0s)
were issued for these FCR Since related modification work will not affect system operability, the MW0s will be developed after restar (2) FCR 78-360 was voide .- __ _ _ _
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(3) FCRs78-126,~79-308,80-221, 80-276,83-136, 83-138,83-151, 85-010,85-025, 85-086',85-126,.85-143,'85-163, 85-224, and 86-160 were close As of the date of the NRC inspection, FCR 85-025 still has some minor tubing work to complete, but this will not. impact the piping
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and pipe support design requirements. Some FCR closeout design verification deficiencies were identified by TED and by the NRC inspector (Paragraphs 7.c, 8, and 9). The piping system and component
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operability assessment required in the CAL Item No. 1, was evaluated by TED. Any identified component deficiencies will be upgraded or
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. repaired to the FSAR commitments. This action was confirmed by the NRC inspecto The remaining FCR issues were considered long term items and will be resolved under the CAL Item 2 action . Item 1.a(5) (Closed)
The NRC inspector reviewed the summary report prepared by TED to document: (1) reinspection findings and subsequent evaluations, (2) resolution of adverse findings which required rework, and (3)
the basis for not performing inspections /walkdowns and resolution
of the resulting nonconformance reports for all safety-related piping system supports not listed in Item 1.b, and had no adverse comments. A formal TED letter will be forwarded to RIII to document the TED actions and evaluation "
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] Review of FCR Program Upgrade-Since the later part of 1985, TED t,as been actively. attempting to reduce
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the FCR closeout backlo The Engineering Service Department was established in October 1985 to handle this work and was replaced by the Information Management Department on January 1986. Since August 1986, the Engineering Support Department was formed within NFED to provide coordination and inter-organization interface to ensure timely FCR
- processing and closecut. The NRC inspector reviewed the following TED procedures which will be implemented in the near future: ,
- NMP-NE-301, " Nuclear Mission Procedure - Plant Modification,"
August 10, 198 * NEP-010, " Nuclear Engineering Procedure - Processing FCRs,"
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- NEP.-010.2, ',' Nuclear Engineering Instruction - FCR Closecut,"
Revision 2, draf Based on this review, the NRC inspector had the following comments: The revised TED FCR program adequately addressed the recommendations made in the Stone and Wester Engineering Corporation report, " Toledo Edison Company Davis-Besse Nuclear Power Station Facility Change
Request Study Report," January 16, 1986.
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4 There appears to be a lack of connection between the policy level document and the working level instructions. As an example, the present "one time only" hanger and piping reinspection procedures IP-M-001 and IP-M-002 had not been transformed into more permanent forms of inspection requirements and acceptance criteria that can be included in the Divisional program identifiable to the TED policy pape Furthermore, Paragraph 3.0 of NMP-NE-301 only referenced limited portions of the operation manuals. Specifically, only Sections 2, 3, and 4 of the QA manual were liste TED stated that additional clarification and procedural revisions will be made. This is an unresolved item (346/86019-02). TED Audit on Design Control / Interface TED QA Department conducted a design control and interface audit of TED NFED and Bechtel on February 3 through March 11, 1986, and issued Audit Report No. 1511 on March 27, 1986. The audit was partially in response to NRC Inspection Report Item 346/86004-05. The audit report documented the scope and 12 Audit Finding Reports (AFRs). The following was the status of these AFRs: AFRs that had already been closed by QA
- No. 1, " Training of Personnel."
- No. 10, "Q Level Classification."
- No. 11, " Procedural Inadequacy (for Modification 103)."
The NRC inspector reviewed the closeout documentation and had no adverse comments, AFRs Addressed by Corrective Action Requests (CAR)
TED issued a CAR No. 86-06 to NFED on August 21, 1986 stating NFED had not provided timely corrective actions for the AFRs 1151-4, 5, 6, 8, 9, and 1 AFRs Open as of September 23, 1986
- No. 2, " Design verification checklists were incomplete," issued on March 11, 1986. NFED responded on April 16, 1986, that they would perform a sample clerical review of FCR packages closed in 1985 and 1986, and modify the design procedures by May 31, 1986. NFED subsequently found that 30 of the 32 safety related FCR closeouts did not have design verification checklists properly completed. 18 of the 32 FCRs were without any design checklist at al NRC followup was documented in Paragraph __ - - _ _
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- No. 3, "It was revealed during a QA random review of 12 FCRs that six did not have design meetings in accordance with procedures," issued on March 11, 1986. NFED responded on April 16, 1986, that they would complete personnel training by June 30, 198 NFED's second response on July 14, 1986 stated that they would review FCRs, correct deficiencies, and conduct training, and requested a new completion date of September 1, 198 QA accepted the new completion date on August 7, 198 * No. 4, " Inadequate instructions for procurement of service,"
issued on March.11, 1986. NFED originally responded on April 16, 1986, that they would correct the procedural inadequacy by July 11, 1986. The second NFED respons.e on July 24, 1986, requested a new completion date of September 12, 1986. The third NFED response on August 26, 1986, requested a new completion date of October 1, 1986. The QA accepted the NFED's third response on September 19, 198 * No. 5, " Task / Work Authorization lacks procedure description of content and scope of procurement for engineering services,"
issued.on March 11, 1986. NFED responded on April 16, 1986, that they would change the procedure by July 11, 1986. The second NFED response on August 26, 1986, requested a new completion date of October 1, 198 * No. 6, " Inadequate NFED design interface control," issued on March 11, 198 NFED responded on April 16, 1986, that they
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would correct the procedural inadequacy by July 11, 198 The second NFED response on August 26, 1986, requested a new completion date of November 1, 198 <
- No. 7, " Inadequate Bechtel design interface control," issued on March 31, 198 Bechtel responded on May 6, 1986, that they
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would correct the procedures. July 27, 1986 Bechtel stated that the procedures had been corrected. TED QA letter to Bechtel on September 19, 1986 stated that Bechtel addressed only portions of the procedural deficiencies, and did not evaluate past design
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effort A new completion date of October 10, 1986 was propose * No. 8, " Purchase order documents (reports, letters, and etc.)
were not properly controlled," issued on March 11, 1986. NFED responded on April 16, 1986, that they would locate the referenced documents and forward them to Engineering Services, and conduct staff training by May 31, 1986. NFED's second response on August 26, 1986, requested a new completion of October 1, 198 ._ . _ - - _ . - . - _ _ _ _ _ . . _ . _ _ _ .
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- No. 9, "Except piping / pipe support, there was no Becntel procedures' addressing interim operability evaluations," issued on March 11, 1986. NFED responded on April 16, 1986, that they
- would review and revise the procedure by June 30, 1986, if required. NFED's second response on August 6, 1986, stated that there was no TED commitment to NRC for evaluation of electrical, I&C, HVAC, and other systems for interim operabilit QA responded to NFED's second response on August 22, 1986, that-TED QA Manual Section 5 requires procedures control be provided for all quality activitie ,
- - No. 12, "An unapproved specification was used by NFED to purchase hot leg level monitoring system hardware," issued on March 11, 1986. .NFED responded on April 16, 1986, that they would review and approve the specification by June 30, 1986. The second NFED response on August 26, 1986, requested
, a new completion date of November 1, 1986.
! During the week of July 7, 1986, QA returned copias of the Audit 1511
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open AFR Nos. 2, 3, 4, 5, 6, 8, 9, and 12 to NFED (QA Internal Audits
- Supervisor memorandum to Audit 1151 File, August 21, 1986). The lack
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of QA control of AFRs and the lack of NFED timely correction of program deficiencies will be reviewed further by the NRC inspector after the plant restart. This is an unresolved item (346/86019-03). TED FCR Design Verification The use of a detailed Design Verification Checklist (DVC) was made effective in TED Procedure NFE-012.02, " Design Verification,"
February 17, 198 The DVC was upgraded in TED Procedure NFED-090,
" Design Verification," Revision 3, May 10, 1985. In addition to the TED review of the 32 FCR closecuts discussed in Paragraph 7.c, TED NFED also reviewed the 317 safety-related FCRs closed by TED from February 1983 to i
May 1985. The review revealed the following: Among the 34 FCRs with completed DVCs,
-(1) 33 were found acceptable, and
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(2) one contained design deficiencies that requires followup action. Among the 46 FCRs with incomplete DVCs,
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(1) 37 contained minor clerical errors, ;
(2) one required no DVC (drawing change),
(3) two contained complete DVCs, and
)- (4) six had omissions that require corrections.
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. Among the 237 FCRs that required engineering review, (1) 17 had complete DVCs, (2) 131 required no DVCs, (3) 58 showed drawing changes, of these (a) 56 design verifications were done in nonconformance reports, (b) one DVC required improvement, and
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(c) one involved more than drawing change, contained potential deficiencies and evaluations were underway, and eight FCRs were found removed from the file, and would be evaluated in the near futur Among the total 317 FCRs, only one was determined to be a CAL 85-13 Item 1 restart required item. Specifically, FCR 85-10, " Modify AFW Support 6C-EBB-4-H11," Supplement 0, closed on April 23, 1986. No deficiencies were found in this FCR. The NRC review of the TED control of FCRs/DVCs will be a part of CAL 85-13 Item 2.c followup action This is an Open Item (346/86019-04). Bechtel FCR Design Verification Unlike the TED FCR design modification closeout, Bechtel does not utilize a DVC. Two Bechtel FCR closecuts were reviewed at Bechtel, and both had design control proble FCR 79-308, Supplement 3, February 25, 1983 The FCR involved a change of pipe routing for the high pressure injection recirculation line. The reanalysis of the revised piping configuration is documented in Bechtel calculation No. T-009 B, Revision D1, February 24, 1983, and No. 54, Revision D1, October 8,
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1984. The affected CCB-12 line showed a design temperature of 260 F in the Bechtel Plant Design Standard M-602, " Piping Class Summary,"
Revision 17, March 17, 1986. Calculation No. 54 used 150 F instead of 260*F. M-602 was subsequently revised to reflect the 150 F temperature utilized in the desig FCR 85-176, Supplement 2, September 25, 1985 This FCR involved design review and component modification for non-standard small bore (S/B) piping supports. Responding to an NRC finding, Bechtel reviewed approximately 800 non-standard (can not be designed by cook book method) S/B supports and found 52 required
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hardware modification.to meet FSAR requirements;-however, none of the 52 exceeded the IEB 79-14 interim stress allowables. The NRC inspector selected Bechtel Calculation Nos. 255, 256, 257, and 258 (emergency diesel generator starting air. supply) for review and found no evidence of design verification documentation and sign-off In addition, the justification for thermal and dead weight load reduction.from the original calculations dated in January 1976 were not documented. -After a lengthy evaluation, the weight reduction was determined to be.due to the absence of pipe insulation, and the thermal. reduction was due to a design. temperature of 264*F being used instead of 400*F for line HBC 59 and 60 as shown in Bechtel Plant
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Design Standard M-60 Subsequently, the Bechtel evaluation of the 52 calculations found that 14 were performed in.accordance with design ,
procedure, and 38 were not (the four reviewed by the NRC inspector ;
are within this 38 group).
7 While reviewing S/B support seismic design, the NRC inspector found reduction factors were not applied on SSE loads and was puzzled by the way applicable primary loadings were combined to come up with the' hardware design loads. In an effort to understand the design *
method,.the NRC inspector requested to review the controlling procedure. The Bechtel staff indicated there was no procedure for Davis-Besse Unit 1, but the method was contained in Procedure SUP-03,
" Support Design Loads,'! Revision 0, June 1,1978, as part of an uncontrolled piping stress analysis procedure manual planned for Davis-Besse Units 2 and 3. Although there was no controlled procedure to perform'the work, the mathematics were verified to be correc Bechtel's failure to follow design procedure EDPI 4.37-11, " Design Calculation," Revision 5, January 29, 1986, and its failure to have a 4 support design load combination procedure are violations of 10 CFR 50,
Appendix B, Criterion V (346/86019-05).
-1 Piping Analysis for Reactor Coolant Purification Letdown (PL) System During followup of Violttion No. 346/86004-03, the NRC inspector reviewed
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TES report TR-3831-10, "ASME Section III Analysis Class 1 Based on IE Bulletin No. 79-14 Program, R. C. Drain and Purification, R. C. Letdown,
- Problems T-001(b) and T-002(b)," November 8, 1982. TR-3831-10 states,
"TES used Babcock and Wilcox Reactor Coolant System Functional
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Specification, CS (F)-3-92/NSS 14, to develop the thermal transients
, for this system. For the thermal transient evaluation of this system '
TES assumed that there is continuous flow in both legs to the letdown 2 -coolers. If a cooler is taken out of service, TES assumes that flow
will not be reinitiated in the leg during normal plant operation and the hot standby condition," The TES assumption was not verified by Bechtel, who contracted TES to perform this analysis, and resulted in
, a memorandum from TED Director, Nuclear Plant Systems to TED Nuclear j Licensing Manager, NES 86-0128, "NRC Violation Item 50-346/86004-03,
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Operational Restrairits Imposed by System Design of the Makeup and Purification System," August 12, 1986, stating, " Plant procedures do not address the Teledyne Analysis assumption as an operational restrain Consequently, the following procedures open MU2B during normal or hot standby conditions:
PROCEDURE CONDITION PP 1101.01 Placing an idle letdown cooler in servic PP 1102.03 For trip recovery, returning letdown to servic PP 1102.10 Plant shutdown and cooldown, strokes MU2B every 100 F to avoid separation of the wedge from the ste SP 1104.02 Makeup and Purification System, high letdown line temperature >135 F and high component cooling water header pressure >135 psig open MU2B if the reason for initiation is know Although the fatigue analysis is required for Class 1 piping, the validity of the assumption and/or impact on the piping fatigue limits should be evaluated. These evaluations should consider appropriate procedure modifications, operational changes, possible physical changes and/or testing (NDE)."
7.e failure of TED to reconcile the PL piping stress analysis design udsis assumptions (design basis) and the actual plant system operation condition is a violation of 10 CFR 50, Appendix B, Criterion III (346/86019-06A).
11. NRC Followup on TED Evaluation of PL System In conjunction with Paragraph 10, the NRC inspector reviewed the following documents:
- Bechtel letter to TED BT-17197, "NRC Potential Violation for Purification Letdown Piping," October 1, 198 * Bechtel Calculation No. M-29, " Pipe Stress Analysis - Reactor Coolant Letdown Piping Fatigue Evaluation," Revision 1, October 21, 198 The analysis was based on the worst case design thermal shock of 580*F to 70 F. The results showed that the 1" CCA-18 branch connection to the 2 1/2" CCA-18 header had an allowable cycle of 1 * TED PCAQR No. 86-0319, " Inappropriate Design Assumption on Plant Operating Characteristics for Piping System Analysis Done by Teledyne," initial review approved on September 2, 1986, and proposed disposition approved on October 23, 198 .
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TED " Safety Evaluation for PCAQR 86-0319," approved on October 28, 198 The NRC inspector concluded that the Bechtel generic evaluation for all contractor submitted analyses was acceptable. The present Bechtel measures which have contractor stress reports go through a Mechanical Supervisor will ensure that system operation will be reviewed. The TED engineers viewed the present PL 15 cycle thermal shock restraint as being overly conservative. However, the station's operating procedures will be revised to include a record of each letdown isolation for engineering evaluatio This will be required until further stress analysis can prove that removal of such an operating restraint will not affect the systems safet Violation Item 346/86019-06A and LER 86-037 are considered close . TES Design of Pressurizer Relief Piping The NRC inspector reviewed the TES piping stress analyses contained in TES Reports TR-5639-2, TR-6388-1, TR-6388-2, and TR-6388-3 (refer to NRC Inspection Report No. 50-346/86004, Paragraph 8.c) using the following Bechtel drawings as the review and verification basis:
- M-230A, " Piping Isometric, Pressurizer Relief System, Containment Building," Revision 13, September 25, 1985. This is the as-built piping configuration drawin * HL-230A, " Hanger Location Drawing, Pressurizer Relief System, Containment Building," Revision 2, February 1, 1984. This is the IEB 79-14 walkdewn as-built hanger location drawin The following piping stress analysis computer input model discrepancies were identified:
- PORV outlet nozzle flange weight and dimension were not modele * Hanger No. 30 CCA-8 H6 was modeled 6" from the as-installed locatio * Hanger No. 30 GCC-8 H5 was modeled 12" from the upstream pipe elbow, was shown 5" from the elbow on HL-230A, and was measured 15" from the elbow during a TED reinspectio * Hanger No. M-1140 H1 was modeled l'-1" from the upstream pipe elbow, was shown 3'-3" from the elbow on HL-230A, and was measured 3'-4 1/8" from the elbow during TED reinspectio * Velan 2 1/2" gate valve No. HV RC-11 with Limitorque valve operator :
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Center of Gravity-
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Weight (lbs) Height Axial offset Lateral offset TES analysis 443 1.916 ft 1.042 ft 0 Velan data 425 i 10% 1.46 ft 1.29 in 1.6 in e Hanger No. 30 GCC-8H8 was modeled as a rigid strut, when in fact, it was a snubbe The lack of design control to ensure correct stress analysis models, and for reconciliation of as-built deviations is a violation of 10 CFR 50, Appendix B, Criterion III (346/86019-06B).
1 NRC Followup on TED Correction of Pressurizer Relief Piping Analyses Computer Reanalysis TED contracted Bechtel to rerun the subject analyses based on as-built data. The NRC inspector reviewed the following Bechtel calculations, and had no adverse findings:
- No. T-11, "PORV Line from Pressurizer to Quench Tank,"
Revision 1, October 11, 1986. This includes piping from the in-line pipe anchor to quench tan * No. T-12, same title, Revision 2, October 12, 1986. This includes piping from pressurizer nozzle to the in-line pipe ancho The reanalysis showed piping stresses were within the code allowable The NRC inspector further reviewed the present loadings for the four damaged hangers discussed in NRC Inspection Report No. 50-346/86004, Paragraph 8.a and had no adverse comments:
400 F Subcool Water Design Load (lbs)*
Hanger N Bechtel Cal Teledyne Cal GCC-8-H10(1) 1935 2405 30 GCC-8-H10(2) 1894 1650 30 GCC-8-H6 2147 2481 30 GCC-8-H5 3937 3558 30 GCC-8-H7 1267 1461
- Load combination was based on same blowdown forces, and different thermal, weight, and seismic loading . , . -. _ -. -- - . -
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, Hanger Failure Cause Determination The cause of the four hanger failure was attributed to installation deficiencies and the lower PORV loop seal temperature. A comparison of data for the later cause is as follows:
469 F Loop Seal 500 F Loop Seal Hanger N PORV Lift PORV Lift
- Blowdown Support Blowdown' Support Force (lbs) Design (its) Force (lbs) Design (1bs)
H 10(1) 966 1930 319 1253 H 10(2) 868 2087 232 1451 H6 1991 3684 1005 2698 H5 1431 4989 1026 4584 H7 606 1814 236 1444 In review of the above data, the NRC inspector determined that even though the blowdown forces at 469 F loop seal are much larger than at 500 F (design), the support design loads are sufficiently large such that the increase in blowdown forces could not have resulted in the hardware failur The most likely cause of failure determined by the NRC inspector, the TED engineers, and the MPR engineers, was that the RELAP 5 computer model was substantiated by PORV laboratory testing using only four changes of flow directions with four piping segment The Davis-Besse Unit 1 PORV piping system involved 15 flow direction changes and one branch "T" connectio The RELAP 5 computer program could have under calculated blowdown forces in far regions of the PORV discharge as in this case where the four failed supports are locate Since the removal of PORV loop seal, the support design loadings were decreased in the majority of the cases. The support design loads based on a subcooled water blowdown from a main steam or feedwater line break are much less likely to occur than loop seal blowdown loads. Based on the above two reasons, the NRC inspector concluded
the present design is adequate and acceptabl Design Related Issues (1) Recheck TES PORV Blowdown Calculation The Bechtel PORV computer reanalysis did not cover the fluid dynamic loading determined by the RELAP 5 MODICY14 computer program. The NRC inspector met with the TED and MPR staff to discuss these design concerns. TED agreed to recheck the subject matter to ensure design adequacies. The MPR review is documented in a letter to TED, "TES Analysis of Davis-Besse Pressurizer PORV Discharge Piping," October 28, 1986. The
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letter concluded that the MPR review of the discharge piping model, the RELAP input deck, and the force-time history plots revealed no significant errors or discrepancie , , _ . . . . _ _ , _ ~ _ _ . . -
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(2) Recheck Worst Design Loading Condition B&W Specification for Davis-Besse bounding PORV inlet conditions due to extended HPI. operation during FSAR event showed the following subcool water discharge conditions:
- FW line break, PORV opens on 2450 psig with subcool water temperature at 640 F maximum, and 602 F minimu * Steam line break, PORV opens on 2450 psig with subcool water temperature at 602 F maximum, and 400 F minimu The NRC inspector discussed the issue with MPR engineers, reviewed the related evaluation documentation, and concurred with the TES/MPR assessment that the 400 F subcool water discharge case is the worst design loading conditio (3) PORV Piping Water Weight The NRC inspector noted that the PORV piping water weight was not considered in the primary load Pipe Material Water In Pipe-4" Sch. 105 5.6 lbs/in 6.18 lbs/in 8" Sch. 20 22.4 lbs/in 22.50 lbs/in During the subcool water blowdown, partial water column (mixture of steam and condensate) weight could affect the design loading on the supports. After discussion with the MPR staff, the NRC inspector accepted the MPR analysis that the high pressure subcool water discharge into the low back pressure regions before the quench tank will create steam flashing. This effect will be substained well into the steady state blowdown conditio (4) Design Temperature and Pressure In review of Bechtel computer reanalysis for the thermal cases the NRC inspector observed the following conditions:
Run N Piping Reg 3on Temp. (*F) Pressure (psig)
1 Pressurizer to PORV 670 2750 PORV to Tank 450 400 2 Pressurizer to PORV 670 2750 PORV to Tank 120 0 The NRC inspector questioned Bechtel's basis for the run No. 1 PORV to tank temperatures and pressure. The TED engineer responded that a site test showed 400 psig to be conservativ The 450 F temperature corresponds to the 400 psig obtained from the steam table. The matter is considered resolve .
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The_NRC inspector concluded that the TED measures taken to correct and resolve the PORV issues were adequate. Violation No. 346/86019-06B is close . PORV Pipe Support Review In conjunction with the Bechtel and MRP re-evaluation of the TES PORV piping analysis, TED engineers reviewed all the PORV pipe supports originally designed by Grinnell Corporation. The review found the as-built supports adequate, with the following exceptions:
- TED Calculation No. C-CSS-64.05-003, " Support No. 30-CCA-8-H2,"
Revision 0, October 21, 1986, found that the support did not meet FSAR commitments and interim stress allowables. TED issued FCR No. 86-0350, " Add stiffener plates to Support No. 30-CCA-8-H2,"
October 20, 198 * TED Calculation No. C-CCS-64.05-001, " Support No. 30GCC-8-H16 and 30 CCA-8-H4, " Revision 0, October 21, 1986, found that the supports did not meet FSAR commitments, but was within the interim stress allowables. TED issued FCR No. 86-0352, " Add stiffener plates to Support No. 30-GCC-8-H16," October 20, 198 TED stated that hardware fixes will be completed prior to the plant restar The NRC inspector reviewed the TED calculations and had no-adverse comment Since TED could not address the underlying causes of the deficiencies, the NRC inspector requested to review the PCAQR (nonconformance report) on the subject matter, and was informed that none was issued. The failure to document nonconformances in PCAQRs is a violation of 10 CFR 50, Appendix B, Criterion XV (346/86019-07).
1 Piping Configuration Re-evaluation Piping Dimension and Support Location TED performed piping diinension and support location reinspection in accordance with IEB 7S-14 requirements. As-built piping dimensions and support locations deviated from the hanger location drawings and the piping stress analysis mathematical models. The Bechtel engineers reviewed all 56 piping calculations that could affect plant restart (CAL 85-13, Item 1). Of these, 34 were accepted by engineering judgment, and the remaining 22 were determined to meet FSAR requirements after computer reanalysis. The NRC inspector reviewed the three worst cases, and had no adverse comments:
- Bechtel Calculation No. 1C, " Auxiliary Feedwater System,"
Revision 8, October 20, 1986. Maximum dimensional deviation was 4'-2". The maximum pipe stress (Node Point 5) increased from 6654 psi to 6860 psi after reanalysis. This is within the ASME Level B piping allowable stress of 18000 psi. The affected pipe anchor No. 225 is still within design load limit . I i
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- Bechtel Calculation No. 2C, " Auxiliary Feedwater System,"
Revision D2, October 21, 1986. Maximum dimensional deviation l was 4'-2 ". The maximum pipe stress (Node Point 535) increased from 8936 psi to 9750 psi after reanalysi This is within the ASME' Level B piping allowable stress of 18000 ps The affected pipe anchor No. 225 is still within design load limit !
- Bechtel Calculation No. 120A, " Main Steam to Auxiliary Feedwater Pump Turbine," Revision 04, September 26, 198 The maximum pipe stress increased from 12429 psi at Node Point No. 61 to 14980 psi at Node Point 118. This is within the ASME Level 8 allowable piping stress of 18000 psi. The affected pipe anchor No. 240 is still within design load limit Motor Operated Valve (MOV) Center Of Gravity (CG)
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Incorrect computer model of MOV CG and dead weight was identified in TES reports (Paragraph 12). LED requested Bechtel to recheck all TES (no other A-E's were contracted by Bechtel to perform piping analysis for Davis-Besse project) submitted piping stress analyse The Bechtel engineers reviewed all 9 TES analyses besides the 2 PORV analyses that could affect plant restart (CAL 85-13, Item 1). Of these, 4 contained no MOVs, 1 had 1 globe valve, 3 contained MOVs without CG and weight discrepancies, and 1 contained MOV with discrepancies that required computer reanalysis. The reanalysis showed piping stress and support loads within the Code stress allowables. The NRC inspector reviewed Bechtel Problem No. T-010A,"
P-29 to Decay Heat Pump Suction Piping," Revision 01, October 29, 1986, and had no adverse comment Phase 2 Activities In conjunction with "a" and "b" above, after plant restart, TED will review approximately 230 Bechtel calculations to reconcile as-built dimensional deviations, and verify the MOV models for the remaining 13 TES piping analyses. These activities are considered to be a part of the CAL Item 2.b requirements. This is an Open Item (346/86019-08).
1 Management Meeting The NRC management met with the licensee management and senior staff members (denoted in Paragraph 1) at the site on October 16, 1986. The NRC management and inspection staff performed a walkdown inspection of the pressurizer relief valve inlet and discharge piping including components, AFTPSS crossover legs inside the " fan alley" areas, and the AFPTSS and AFP piping systems at the turbine and pump nozzle connection regions. After observing the hardware, a meeting was held to discuss:
(1) the status of RIII CAL 85-13 items that will be completed by TED prior to restart, (2) the corrective actions taken to resolve the PORV design
o deficiencies and RIII concerns, (3) the recent findings relating to the design reconciliation effort for the as-built data obtained from the TED re-walkdown of the systems, and (4) the TED effort to improve timely resolution and closecut of audit findings. The TED actions to correct the RCP 2-2 bearing (a non-safety related item) f ailure, and the measures taken to prevent recurrence were also discusse . 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. Three unresolved items disclosed during this inspection are discussed in Paragraphs 2.j(1), 6.b and . Open Items Open Items are matter which have been discussed witie the licensee, which will be reviewed further by the inspector, and which will involve some action oa the part of the NRC or licensee or bot Two open items disclosed during the inspection are discussed in Paragraphs 8 and 1 . Exit Interview The NRC inspector met with licensee representative (denoted in Paragraph 1)
at the conclusion of the inspectio The inspector summarized the scope and findings of the inspectio The inspector also discussed the likely informational content of the inspection report with regard to documents reviewed by the inspector during the inspection. The licensee representatives did not identify any such documents as proprietar _