IR 05000346/1986004

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Insp Rept 50-346/86-04 on 860114-0402.Violations Noted: Failure of Bechtel to Rept Noncompliance W/Fsar Conditions Re Stress Calculations to Util & Insp & Evaluation of Piping & Supports Not Conducted
ML20210N950
Person / Time
Site: Davis Besse Cleveland Electric icon.png
Issue date: 04/30/1986
From: Danielson D, Fair J, Yin I
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20210N927 List:
References
50-346-86-04, 50-346-86-4, CAL-85-13, IEB-79-14, NUDOCS 8605050334
Preceding documents:
Download: ML20210N950 (18)
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U.S. NUCLEAR REGULATORY COMMISSION

REGION III

Report No. 50-346/86004(DRS)

Docket No. 50-346 License No. NPF-3 Licensee: Toledo Edison Company Edison Plaza 300 Madison Avenue Toledo, OH 43652 Facility Name: Davis-Besse Nuclear Power Station, Unit 1 Inspection At: Dasis-Besse Site, Oak Harbor, OH Bechtel Power Corporation, Gaithersburg, MD (Bechtel)

Inspection Conducted: January 14-16, February 5-6 and 19-20, 1986 at the site January 29-30 and April 1-2, 1986 at Bechtel Inspectors- . T. Yin oh W

Da~te (April 1, 1986 only)

duin Date Approved By:

e-D. H. Danielson, Chief d[3o[/t Materials and Processes Section Date Inspection Summary Inspection on January 14 through April 2, 1986 (Report No. 50-356/86004(DRS))

Areas Inspected: Special, announced inspection of the auxiliary feedwater pump turbine steam supply (AFPTSS) piping modifications; the Facility Change Request (FCR) system; the implementation of Region III (RIII) Confirmatory Action Letter (CAL) 85-13 actions; actions on Licensee Event Reports (LER); the status of completion of IE Bulletin (IEB) 79-14; the Bechtel control of High Energy Line Break (HELB) analyses; and followup on previous inspection finding Results: Of the areas inspected, two violations were identified; (failure of the Bechtel staff to follow procedures and failure of TED to follow site procedures - Paragraphs 4.b and 8.b(1); failure of the licensee to take adequate corrective action on identified problems - Paragraph 8.b(2)).

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DETAILS 1. Persons Contacted Toledo Edison Company (TED)

  • +*T. J. Bloom, Senior Licensing Specialist
  • T. Chowdharm, Manager, Engineering Services Department
  • +*P. H. Straube, Senior Engineer J. Dunne, Senior Engineer T. J. Myers, Nuclear Safety and Licensing Director
  • L. Ramset, Quality Assurance Director F. R. Miller, Staff Engineer P. W. Jacobsen, Senior Engineer J. F. Helle, Nuclear Facility Engineering Director
  • S. J. Osting, Senior Assistant Engineer D. R. Wyokko, Regulatory Affairs Supervisor
  • D. Kies, Manager, Mechanical / Structural Engineering
  • H. Brinkmann, Director, Nuclear Facility Engineering C. Merkbel, Civil and Structural Systems Engineer Bechtel Associates Professional Corporation, Ohio (Bechtel)

J. W. Brothers, Chief, Quality Engineering N. Tolani, Senior Engineer

+ S. Wasserman, Mechanical Engineer Supervisor

  • M. L. Murphy, Senior Engineer
  • W. C. Lowery, Project QA Engineer A. T. Vieira, Engineering Technical Specialist
  • +D. C. Kansal, Deputy Division QA Manager '

J. M. Ogle, Civil Engineer Supervisor

  • +D. L. Gill, Project Quality Engineer

+E. J. Ray, Project Engineer C. H. Abutaa, Senior Engineer R. Lee, Engineer Supervisor

+ I. Gillespie, QA Manager, Projects

+ R. Kalavar, QA Manager, Audit S. A. Bernsen, Division Manager of QA J. B. Wallis, Senior Engineer

  • V. R. Marathe, Assistant Project Engineer

, K. I. Patel, Engineering Supervisor U.S. Nuclear Regulatory Commission, Region Ill (RIII)

  • Rogers, Senior Resident Inspector
*D. Kosloff, Resident Inspector

+ Denotes those attending the management exit meeting on January 30, 1986

. at Bechte * Denotes those attending the management exit meeting on February 20, 1986

at the site.

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  • Denotes those attending the management exit meeting on April 2,1986 at Bechte . Licensee Action on Previous Inspection Findings (Closed) Unresolved Item (346/83-17-04): Several new vintage Grinnell Corporation hydraulic snubbers with Miller cylinders, including PSP-1-H4 and PSP-1-H6 installed on the Pressurizer Spray Piping System, were observed installed without fluid reservoir breather and filter units. The NRC inspector reviewed the site Temporary Modification Request, dated January 10, 1986, and Section 8.2.5 of Procedure MP1410.02.04, " Maintenance of Hydraulic Snubbers," and considered the licensee's measures for reinstalling the filter units to be acceptable. A purchase order procuring 50 new filter units was issued on January 10, 198 (Closed) Violation (346/85013-01): The licensee failed to document nonconformances in accordance with procedure requirements. The NRC inspector reviewed Item IV.A.2 of the TED response letter (Serial No.1-604) to the NRC, dated January 27, 1986, and considered it acceptable. TED corrective actions are documented in RIII Inspection Reports No. 50-346/85013, Paragraph 8; No. 50-346/85033, Paragraphs 2 through 5; No. 50-346/85035, Paragraph 4.b; and Paragraph 4.a of this repor (Closed) Violation (346/85013-02): After the AFPTSS problems were identified, the TED evaluations did not investigate the cause of the problem and consequently measures to prevent recurrence were not developed. The NRC inspector reviewed Item IV.A.3 of the TED response letter (Serial No. 1-604) to the NRC, dated January 27, 1986, and considered it acceptable. TED corrective actions are documented in RIII Inspection Reports No. 50-346/85013, Paragraph 10; No. 50-346/85035, Paragraphs 4.a and 6; and Paragraph 6 of this repor (Closed) Violation (346/85013-03): Inadequate piping suspension system QC inspection and ineffective implementation of the IEB 79-14

! walkdown inspection program. The NRC inspector reviewed Item IV.A.1 l of the TED response letter (Serial No. 1-604) to the NRC, dated

{ January 27, 1986, and considered it acceptable. TED corrective i actions are documented in RIII Inspection Reports No. 50-346/85033, i_ Paragraphs 2 to 7; No. 50-346/85035, Paragraphs 4.b and 5; and L Paragraph 4 of this report.

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! (Closed) Violation (346/85013-06): The licensee failed to report l AFPTSS component deficiencies in accordance with 10 CFR 50.73 l requirements. The NRC inspector reviewed Item IV.B of the TED l' response letter (Serial No. 1-604) to the NRC, dated January 27, 1986, and considered it acceptable. TED corrective actions are documented in RIII Inspection Reports No. 50-346/85033, Paragraph 2; l No. 50-346/85035, Paragraph 4.b; and Paragraph 4.a of this report.

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f. (Closed) Violation (346/85031-01): Lack of a design interface procedure between Bechtel and Grinnell for evaluating pipe hangers in accordance with IEB 79-02 and IEB 79-14. The NRC inspector reviewed TED response letters, Serial No. 1-593, dated November 25, 1985, and Serial No. 1-616, dated February 27, 1986, and considered the matter resolved. The need to review Grinnell calculations is discussed in Paragraph 10 of this repo: g. (Closed) Violation (346/85031-02): TED did not effectively implement its FCR system in that a number of safety-related supports were not restored to their FSAR condition in a timely manner. The NRC inspector reviewed TED response letters, Serial No. 1-593, dated November 25, 1985, and Serial No. 1-616, dated February 27, 1986, and considered the licensee actions to be acceptable. The near term support modification work to assure FSAR conditions were met, was conducted in accordance with RIII CAL 85-13, Item 1.a(4)

requirements (see Paragraph'4.d of this report). The licensee's long term upgrade of the FCR program are being reviewed by RIII and NRC Headquarters personnel. See Paragraph 2.k of this report for details concerning piping design and support modification h. (Closed)Viofation(346/85035-02): The licensee failed to use the appropriate allowable stresses specified in Bechtel Evaluation Procedure MGP-04 for evaluating stresses at weld attachments to the piping pressure boundary. The NRC inspector reviewed the TED response letter, Serial No.1-314, and Bechtel Procedure CGP-04,

" Procedure for Evaluating Nonconformance Reports Related to Pipe Supports, Pipe Anchors, and Seismic Restraints at Davis Besse Nuclear Power Station, Unit 1," Revision 1, dated January 28, 198 The NRC-inspector noted that Procedure CGP-04 allowed higher allowable stresses for the SSE load combination than Bechtel Procedure MGP-0 Bechtel reanalyzed all affected piping using the revised allowable stresses. The results were documented in a Bechtel letter to TED, BT-16555, " Procedure GCP-04; Faulted Condition," dated April 2, 1986. Procedure CGP-04 was subsequently revised to reflect the lower allowable stresses for the SSE load combination on April 4, 1986 as Revision i. (0 pen) Unresolved Item (346/85035-03): -The NCR evaluations for weld deficiencies designed to the AISC specification do not require meeting the specification minimum weld sizes which correspond to the base material thicknesses. The NRC inspector reviewed the licensee's response contained in an intracompany memorandura (File 0093, T-0294)

dated January 24, 1986. The NRC inspector will discuss this matter with NRC-NRR to determine if it represents a position acceptable to the NRC staff.

! J. (Closed) Unresolved Item (346/85035-04): Bechtel exhibited

questionable design control for conducting HELB analyses and whip restraint designs. See Paragraph 7 for details of the followup

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. (0 pen) Open Item (346/85035-05): The TED MWO and FCR systems require :

further evaluation and improvement. A temporary organization named Engineering Services Department (ESD) was formed to focus the company's attention on closing out open FCRs and to establish better ways of handling future FCRs. ESD is presently manned by a full time technical staff of ten and occasionally by engineers of various disciplines depending on specific needs. The NRC inspector met with the Manager of ESD and reviewed the ESD organization chart and the

" Project Proposal for Closeout Backlog Evaluation" to evaluate the scope and provisions of the project and had no adverse comment A more detailed review of ESD will be conducted by the NRC inspector '

to assess the effectiveness of the new system. In addition to the above effort, an audit was conducted by Stone and Webster Engineering Corporation of the FCR system to identify system deficiencies and to recommend improvement ' Licensee Action on Licensee Event Reports (LERs) (Closed) LER (346/85019-LL): "PORV Discharge Line Overstressed Due to Inadequate Heat Trace," reported on November 6, 1985. See Paragraph 8 for inspection details, (Closed) LER (346/85023-LL): " Error in the High Energy Line Break Analysis in the Auxiliary Building," reported on December 28, 198 During a TED review of the environmental qualification (EQ) and single failure analysis for a proposed modification to the AFPTSS piping, TED discovered that portions of the system upstream of the MS admission valves 106, 106A, 107 and 107A would not be isolated during a postulated high energy line break event. A break in these pipe

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sections would affect rooms 500, 501 and other connected rooms. The licensee's corrective actions are included in RIII Inspection Report No. 50-346/85035, Paragraph The NRC inspector noted that similar situations could exist in other safety-related high energy piping systems. TED stated that efforts to expand the scope of the HELB review had been initiated, and that deficiencies had been discovered in the main feedwater lines. TED also indicated that nonconforming conditions will be documented in either an amendment to LER 85023 or in a new LE . Implementation of RIII CAL 85-13_[ction Items As a result of a meeting con @;cte at the site on October 9, 1985 (RIII Inspection Report No. 50 Af ' 150 , Paragraph 4) RIII CAL 85-13 was issued on October 17, 19t5 The licensee's implementation of the actions set forth in the CAL was reviewed by the NRC inspector. The status of CAL Item 1 (action items prior to plant restart) is as follows:

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, ] Item 1.a(1)-(Closed)

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s Reinspections per_ CAL 85-13 Item . ,

  • All 2365 hangers have been inspected; 921 require evaluation and

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... ,3; prior to restar IStatus of Engineering Evaluations as of March 25, 1986

  • 858 NCRs', were writteE ,af ter evaluation of the 921 inspected hangers s .
  • ,Of the 858 NCRs, 656 were dispositioned to "Use-As-Is"
  • Of the 858 NCRs,202 required corrective actions; rework has

been completed for 180 of the NCRs

,- s Item 1.a(2) (Closed)

i The following piping stress analyses were rerun by Bechtel to determine system operability:

Systems Affected Bechtel Calculation No '

LPI 188, 180, 80A, and T-010A LPI/ Core Flood T-008 l / HPI 56D, 56F, and T-009B

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l Containment Spray 22F

.. Containment Sump to ECCS 32E

' Hydrogen Dilution 119L q MS-Exhaust from AFPT 161 The NRC inspector selected the following calculations for review at Bechtel and concluded that the evaluations were technically sound and conservative:

  • No. 56F, " Davis-Besse High Pressure Injection System,"

Revision C2, dated December-20, 198 . .:-

  • No. 188, " Davis-Besse Low Pressure Injection System,"

Revision C3, dated January 3, 198 In Calculation No. 18B there were two restraints that did not meet FSAR commitment They were not reported to TED in accordance with Bechtel Procedure MGP-04, " Procedure for Control of Interim /Short-Term Allowable Stress Criteria for Seismic Category I Piping Systems at Davis-Besse Nuclear Power Station Unit 1," Revision 1, dated September 27, 198 Procedure MGP-04, Paragraph 6.1.a and 6. states:

"For each nonconformance determined by calculation to require the use of interim stress criteria per Section 4.0, TED Facility Engineering shall be notified as follows:

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(1) Bechtel shall notify by telephone the Director, Nuclear Facility Engineering, or General Supervisor, Facility Engineerin (2) The notification by telephone shall be followed up by a written confirmation. The written confirmation shall include the NCR number, specific deviation evaluated, and the specific analyzed results. Modifications required to meet design /SAR requirements will be included. The written confirmation shall also include a comparison of the interim versus SAR allowables involved with the design and the available safety margin."

During the inspection, the NRC inspector noted that the following systems and pipe restraints required the use of interim stress criteria (exceeded FSAR allowables):

System Stress Calculation N Restraint N Containment Spray 22F(C2) -

LPI T-010A(C6) -

LPI 18B (C3) * GCB-1-H13

Prior to the conclusion of the NRC inspection, the licensee initiated a number of corrective actions which are included in the following documents:

  • TED letter to Bechtel, "NCR Resolutions," advanced copy dated-February 11, 1986; formal letter dated February 20, 1986. The letter requested a listing of all pipe support evaluations where FSAR commitments were exceeded but interim requirements were me * Bechtel letter to TED, BT-16335, dated February 14, 1986 provided the list requested by TE * Bechtel Interoffice Memorandum from Project Engineer to the Group Supervisors, "NRC Violation - NCRs Interim /Short-Term Allowables," dated March 15, 1986 provided additional instructions and training for the performing of evaluations, c. Item 1.a(3) (Closed)

See RIII Inspection Report No. 50-346/85035, Paragraph 4.b.

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. Item 1.a(4) (0 pen)

The FCRs that could impact safety-related piping system operability are listed in RIII Inspection Report No. 50-346/85035, Paragraph An update of this list (as of January 15, 1986) is as follows:

(1) FCR No.77-213, 77-398,80-221, and 80-276: No Maintenance Work Orders (MW0s) were issued for these FCRs. Since related modification work will not affect system operability, the MW0s will be developed after restart.

(2) FCR 78-360: This FCR was voide (3) FCRs78-126, 79-308,83-151, 85-086,85-010, 85-126,85-163, 85-176, and 85-224 were close (4) Status of the remaining FCRs:

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FCR N (No. of Supports Involved) Status79-421 AFW pump turbine modification (15) Work completed, in process of being closed out.

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83-136~ Replace FW pump governor (2) Work in progress

,.83-138 Change of 14 valves (4) Work in progress- 85-025 Motor driven FW pump (103) Work in progress

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85-143 Relocate steam admission Work in progress valves (6)85-160 PORV loop seal drain (8) Work in progress Item 1.a(5) (0 pen)

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The documentation in RIII Inspection Report No. 50-346/85035,

! regarding Paragraph 4.b remains unchange . Status of Completion of IE Bulletin 79-14 RIII Inspection Report No. 50-346/85031, Paragraph 6.a, documents that TED's ineffective utilization of the FCR system resulted in some support

{. component rework not being completed as stated in a TED letter to RII '

In response to the RIII findings, TED reported in a letter to RIII (Serial No. 1-598, dated December 20, 1985) that several FCRs which were originally issued as a result of the I&E Bulletin walkdown were identified as still being open, and for several of the work items which were originally identified as closed, the work was not yet fully completed.

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Due to NRC questions concerning whether or not TED had made misleading statements to the NRC, the NRC inspector discussed the matters with the TED licensing and QA directors. Through their cooperation, the NRC inspector obtained the following information to aid in his assessment of the matter: TED Letters to RIII Documenting the Status of IEB 79-14 Implementation Scope or Status Serial No. (Date) Connitment Date of Work 1-137 (6/16/80) 3/1/81 Approximate 210 supports for accessible areas 1-177 (12/20/80) 12/31/81 207 supports for accessible areas 3/82 outage 15 supports for inaccessible areas (only 60% analyses completed)

1-187 (2/13/81) 3/82 outage Modification will be made for inaccessible areas 1-201 (5/22/81) 3/82 outage 45 supports for ipaccessible areas 1-223 (11/13/81) 12/82 133 supports for inaccessible areas 1982 outage Remaining supports inside containment 1-289 (8/19/82) Next outage 11 support modifications inside containment 1983 Remaining supports outside containment 1-429 (5/31/84) 12/31/83 All supports " mechanically completed" Causes of Support Modifications not Being Completed As Stated in TED Letter 1-429 Actual Completion Document N Support N Date Causes FCR 80-87 FSK-M-CCB-8-25-H 12/84 Bechtel failed to provide TED with modification package

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Actual Completion Document N Support N Date Causes FCR 80-87 FSK-M-HCD-15-18-H 9/85 Same as above FCR 80-91 A 399/A 402 4/10/84 TED Licensing (NCR393-79) (instead of oversight Feb.'84)

NCR 396-79 EBD-12-19 4/30/84 TED Licensing and EBD-12-20 (instead of oversight Feb.'84)

FCR 80-125 SR 33, 35, and 5/1/84 TED FMD 41 oversight FCR 80-125 SR37 3/30/84 TED FMD oversight FCR 80-125 A171 1/16/84 TED FMD oversight NCR 524 A8 11/85 TED FMD and QC oversight As a result of the review, the NRC inspector concluded that (1) from 1980 to 1982 TED reported the status of work as requested in IEB 79-14, (2) it appears that substantial funds were spent to implement the actions set forth in IEB 79-14 and IEB 79-02, and (3) despite the identified deficiencies, less than 4% of the components were not modified as reported in the TED letters to RIII. No further action is planned by Region III at this tim . AFPTSS System Modification Actions taken by TED to modify the AFPTSS system are discussed in RIII Inspection Report No. 50-346/85035, Paragraph 6. During the site inspection of January 14, 1986, the NRC inspector performed a waltdown of the AFPTSS crossover leg piping connecting steam generator No. 1-2 to auxiliary feedwater pump turbine No.1-1. This crossover leg, with its many piping expansion loops, is located in the " Fan Alley" (area having a large amount of HVAC equipment) on floor elevation 623'-0" inside the auxiliary buildin Inspection of Piping Restraints An inspection of piping restraints and anchor modifications, to evaluate their functionability, was conducted by the NRC inspecto The piping section observed was between valve HV 106 and valve HV 106A. No deficiencies were identifie . .

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. Questions Concerning High Thermal Lockup Stress During the walkdown, the NRC inspector observed excessive restraining within the region bounded by restraints 3A-EBD-19-H 104 and 3A-EBD-19-H 109. This region involved four directional changes in less than 22 feet of 6" diameter pipe. The restraints are:

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X and Y restraints H 105 -

Z restraint H 106 -

X restraint H 107 -

Y restraint H 108 -

Z restraint H 109 -

X and Y restraints The primary reason for all the thermal expansion loops installed in the crossover leg is to minimize piping thermal stresse The placement of so many rigid restraints within the pipe loop is a contradiction of this design inten In January 1986 the NRC inspector reviewed Bechtel piping stress analysis Problem 40A, " Main Steam," Revision D2, dated September 25, 1985, and observed: (1) a high ANSI B31.1 secondary stress (RIII Inspection Report No. 50-346/85035, Paragraph 6.d.(2)(a) reported a maximum ASME Section III secondary stress only), and (2) a piping dimensional deviation at a critical area between rigid restraint H 105 (Data Point 122) and pipe elbow (Data Point 124). The Bechtel analysis showed 25", the Bechtel Hanger Location Drawing HL-203J showed 15", and the field remeasurement taken during the inspection was 14". During the NRC inspection the piping stress analysis was rerun in a simplified configuration based on corrected dimension The results were:

Data Point Thermal and Existing /New Code allowable Existing /New SAM Stress (psi) % Change (psi) reference 124/72 33,942/33,298 -1.9% 37,500 128/80 8,858/12,690 +43.3% 37,500 132/90 15,410/24,043 +56% 37,500 Subsequent to the review, the NRC inspector noted:

~(1) The present design met code requirements but system relief should be provide (2) The present TED Procedure IP-M-002, "The Piping Support Inspec-tion and Verification Program: Verification of Support / Component Location and Quantity," Revision 1, dated November 7, 1985 . states

" Denote all dimensional differences on the walkdown HL drawings."

The dimensional differences are then evaluated by Bechtel to determine if there will be any effect on the pipe stress analyse The NRC inspector stated his view that priority should be given to critical /high stressed area . .

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7. Bechtel Design Control for HELB Analysis A number of unresolved matters were raised during a previous NRC inspection conducted at Bechtel on December 4-6, 1985 (RIII Inspection Report No. 50-346/85035, Paragraph 6.e(4)). A followup review conducted at Bechtel on January 29-30 and April 2, 1986 revealed the following conditions: The HELB analysis for the AFPTSS modification was based on desk top design guides. Bechtel Procedure, MGP-05," Procedure for Interdiscipline Coordination for High Energy Line Break Evaluation at Davis-Besse Nuclear Power Station Unit No.1," Revision 1. was issued for use on March 14, 198 The desk top design guide, " Pipe Whip Details," was determined to be unacceptable by Bechtel upon reevaluation. Among the 21 postulated AFPTSS break locations, five indicated that the plastic hinges will form at the third elbow instead of the second elbow. However, new whip restraints were not required due to the location of the impact area The impact areas are as follows:

Break N Area of Impact 6 ceiling 21 wall penetration 36 wall 38 wall 69 no safety-related equipment A Bechtel QA Management Audit, No. 12501-05, issued QA Finding No. 1 on February 28, 1986, documenting a similar findin Design tables utilized in the AFPTSS HELB component design were not approved for the specific application These design tables also had not been evaluated for applicabilit Subsequent Bechtel evaluation determined that the tables presented in RIII Inspection Report No. 50-346/85035, Paragraph 6.e(4)(c), were conservative for the applicatio The remaining two unresolved matters documented in RIII Inspection Report No. 50-346/85035, Paragraphs 6.3(4)(d), and 6.e(4)(f) were adequately addressed in MGP-0 The NRC inspector questioned whether or not all the previous Bechtel HELB analyses were adequate. Bechtel's response was as follows:

  • The original design criteria were very conservativ The FSAR, Revision 13, dated June 1975, states in Paragraph 3.6.2.2.2,

" Pipe Restraint Design Criteria to Prevent Pipe Whip Impact Outside the Containment - The basic philosophy used to prevent high energy pipes from whipping is to provide restraints of

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sufficient capacity and with such spacing that pipe whipping cannot develop. These restraints are independent of operating and seismic supports. Consequently, they are designated for pipe rupture loads onl Pipe rupture of either guillotine or side-split type are postulated to occur in accordance with Table 3-4b for the 36 inch main steam and 18 inch main feedwater lines. Allowable pipe spans are calculated, assuming that the force developed during the accident experience is transferred to the pipe restrain The pipe restraints are then designed to withstand this forc The restraints are further designed to prevent the pipe from shearing off and generating missiles. Jet effects from ruptured pipe are

, considered in designing pipes, walls, and shiel The concurrent effects of jets and pressure differentials are also considered in designing walls and shields."

  • Bechtel reevaluated the Steam Generator Blowdown System modifica-tion conducted in 1982. Of the approximate 20 postulated pipe break locations, one was found having a plastic hinge location at the third elbow instead of the originally determined second elbo Field inspection observed no safety-related equipment in the force impact area The NRC inspector considers the Bechtel response acceptabl . Followup on Licensee Event Report (LER)

The inspector reviewed the TED LER 85019, "PORV Discharge Line Overstressed Due to Inadequate Heat Trace," dated November 6, 1985, and questioned the TED actions taken to identify and resolve the issu TED Inspection Findings During recent hanger reinspections the following damage and deficiencies were identified on the PORV discharge piping:

Hanger N NCR No. (date) Damages / Deficiencies 30 GCC-9-H10 0920 (10/3/85) Concrete cracked; plate separated from wal GCC-8-H6 0921 (10/3/85) Concrete cracked and fell; plate separated from wal GCC-8-H5 0914 (10/4/85) Concrete spallin GCC-8-H7 0822 (10/1/85) Pipe clamp slipped off and binde . .

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Upon removal of the H10 and H6 baseplates for repair, the following abandoned drilled anchor bolt holes were discovere Hanger N Abandoned Holes H10 Three holes were outside the three bolt diameter limit (will not affect strength of installed bolts)

One hole was at the 2.25 diameter limi Evaluation determined the spacing to be acceptabl H6 One hole was outside three bolt diameter limi Chip void in concrete. Evaluation determined this void to be acceptabl Since two out of two baseplates that were removed for repair revealed abandoned anchor bolt holes, the NRC inspector indicated to the licensee that additional inspections of the areas behind other baseplates in the vicinity of H10 and H6, including Hanger 6C-EBB-4-H12 (see Paragraph 9 for justification), should be considered. Depending on the hole / void configuration and relative distance to the affected anchor bolts, these holes / voids could mean reduction in support strength or fracturing of concrete at a loading much lower than full load capacit The TED engineering department did not share the NRC inspector's view. TED's justification for not inspecting for possible abandoned drilled holes existing in other baseplates will be discussed further during a future inspectio This is ar. unresolved item (346/86004-02).

b. TED Correr_t13 e Actions As a resul' ;f the TMI event, the NRC issued NUREG-0737 in 1980 requesting vtilities to reevaluate the pressurizer safety and relief valve operations. The Teledyne Engineering Services (TES) issued Technical Report (TR), TR-5639-2, " Davis-Besse Analysis and Evaluation of the Safety / Relief Valve Discharge System per NRC NUREG-0737," Revision 0, dated January 1983, to address this issu This report indicated that the PORV loop seal temperature should be maintained at 500 F. As a result of the damaged supports observed to date, the NRC inspector reviewed the operation records and procedures and concluded that the TED action to ensure design implementation was ineffective. The bases for the determination were:

(1) From 1976 to 1979 there was a total of 82 PORV lifts where the loop seal temperature was less than 500 F. Of these, 46 were assumed by TED to be 130 F. The 46 lifts with the temperature less than 400 F is in violation of Davis-Besse Periodic Test

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Procedure, PT 5164.03, " Pressurizer Relief Valve Heat Trace Test," Revision 5, dated August 20, 1982, which indicates that the test acceptance criteria for heat trace (T772) must be above 400 F, to allow PORV (RC 2A) to be lifted 650 times and 25 times for a heat trace below 400 This is a violation of 10 CFR 50, Appendix B, Criterion V (346/86004-01B).

(2) Contrary to the conclusion stated in TES TR-5639-2, PT 5164.03 was not revised to reflect the latest loop seal temperature of 500 F and no attempt was made to inspect and evaluate the condition of the piping and support This is a violation of 10 CFR 50, Appendix B, Criterion XVI (346/86004-03).

c. TES Design Control The NRC inspector reviewed the following TES reports docun.enting their evaluations of effects on PORV inlet and discharge piping and supports:

TR-5639-2, " Analysis and Evaluation of the Safety / Relief Valve Discharge System per NUREG-0737," dated January 198 *

TR-6388-1, " Analysis of Davis-Besse, Unit 1 Pressurizer Relief Line 400 F Loop Seal Blowdown," dated September 26, 198 *

TR-6388-2, " Davis-Besse Nuclear Power Station Reconciliation of ASME Section III Evaluation of Class 1 Pressurizer Relief Piping," dated November 15, 198 *

TR-6388-3, " Davis-Besse Nuclear Power Station Evaluation of Class 3 Pressurizer Relief Piping," dated January 21, 198 Subsequent to the review, the NRC inspector had the following concents:

(1) The maximum transient dynamic loading locations where piping restraint danage was observed, differed from the TES analytical predictio (2) Review of the reference lists revealed what appeared to be an analytical bases that were either preliminary or interim in natur (3) Several support design loads, based on the present dynamic transient without loop seal (at 400 F subcooled water condition),

were many magnitudes higher than the original design with loop seal. The present support loads should have been bounded by the original desig _ _ _ _ _ __ __ _ _- _ _ _ , _

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(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 verified using the above criteri (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 loading will bound all modes of operatio Further review of the subject matter is planne This is an unresolved item (346/86004-04). PORV Operability Without Loop Seal The TED decision to remove the PORV loop seal could affect PORV long term operability due to a continuous steam leak (small amount) and hydrogen attack of valve disc and seat TED letter (A85-30681) dated August 30, 1985, to Crosby Valve and Gage Company, the PORV designer and manufacturer, requested evaluation of this conditio The Crosby response was documented in a letter to TED, dated February 28, 1986. The NRC inspector reviewed the Crosby letter and observed the valve leak detection devices installed on the piping syste The measures taken to ensure system safe operation due to the design modification were determined to be acceptabl . Auxiliary Feedwater (AFW) Line Transient During the inspection conducted inside the containment on February 6, 1986, the NRC inspector observed concrete spalling and grout cracking at load intensified locations on the baseplate for AFW Train 1-2 hanger 6C-EBB-4-H12. These conditions are indications of excessive transient loads that may not been accounted for in the Bechtel design and analysi The NRC inspector's review of NCR 85-0198, issued on September 3, 1985, and evaluated by TED engineering on October 5, 1985, identified no indication that the causes and measures taken to correct and prevent recurrence had been evaluate During the inspection conducted on February 19-20, 1986, the NRC inspector further discussed his observations. The TED engineer stated that the concrete damage most likely occurred during construction. The NRC inspector re-entered the containment on February 20, 1986 for a closer examination of hanger H12, and observed the following additional adverse conditions:

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  • One of the anchor bolts appeared to be ben * One of the baseplates was partially separated from the wal * Markings that could have resulted from thermal displacement and possible loads in the direction the bolt appeared to be bent, were observed where the restraint contacted the pip During the NRC inspection conducted at Bechtel on April 1, 1986, TED presented the following documents to address the NRC inspector's concern:
  • Bechtel letter to TED, BT-16470, " Evaluation of Support 6C-EBB-4-H12 Auxiliary Feedwater System," dated March 17, 1986. This letter documented the Bechtel inspection and evaluation. Bechtel concluded that there was no evidence of transient loads affecting the suppor * Bechtel letter to TED, BT-16552, "NRC Inspection AFW System Dynamic Transients," dated April 2, 1986. This letter compared the steam condensation with the check valve leak rate and concluded that a severe dynamic transient will not occu The TED representative further committed to reinspect the piping system including Support SC-EBB-4-H12 during the next refueling outage after restar The NRC inspector reviewed the above documents and considered the TED actions to be acceptabl . TED Review of Grinnell Calculations Due to the lack of a formal design interface control etween TED, Bechtel, and Grinnell (now a Tyco company) and the fact that G 'nnell did not have final design responsibility for the adequacy of the or., nal 4,000 (estimate) safety-related support calculations and subsequeat IEB 79-02 and IEB 79-14 evaluations, the calculations performed by Grinnell were in question. Since completion of the IEB 79-02 and IEB 79-14 work, some of the Grinnell hanger calculations were replaced through the disposition of Bechtel NCRs and FCR The licensee is in the process of obtaining all hard copies of the Grinnell original calculations and IEB 79-02 and IEB 79-14 calculations so they can complete their review of this wor The NRC inspector discussed the matter with the licensee and indicated that the following areas warrant additional review: In January 1986 TED requested Grinnell transmit all Davis Besse 1 original hanger calculations and subsequent calculations for IEB 79-02 and IEB 79-14 to TED. No documents have been received to date. Further efforts are required to obtain these calculations if they are available, Upon receipt of Grinnell calculations, TED will develop a program to review these calculations to assure compliance with design procedure The NRC inspector concurred with TEDS recommendation that this work be completed prior to the next refueling outag _

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. The TED QA Department recently conducted an audit in the area of design interface of vendors who provide engineering service to TE The NRC inspector plans to revier the audit report during a subsequent inspectio This is an unresolved item (346/86004-05).

11. 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 8.a. 8.c, and 1 . Exit Interview The NRC inspector met with licensee representative (denoted in Paragraph 1)

at the conclusion of the inspection. The inspector summarized the scope and findings of the inspection. The inspector also discussed the likely informational content of the inspection report with regard to documents reviewed by the inspector during the inspectio The licensee representatives did not identify any such documents as proprietar

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