IR 05000219/1985014

Insp Rept 50-219/85-14 on 850514-17.Violation Noted:Failure to Perform Activities Re IE Bulletins 79-02 & 79-14 W/Required Procedures or W/Controlled Documents.Deviation Noted:Documentation Re as-built Piping Unavailable

ML20205B692
Person / Time
Site:	Oyster Creek
Issue date:	07/05/1985
From:	Bettenhausen L, Kamal Manoly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20205B654	List: IR 05000219/1985014 ML20205B648 ML20205B664 ML20205B675
References
50-219-85-14, IEB-79-02, IEB-79-04, IEB-79-14, IEB-79-2, IEB-79-4, NUDOCS 8509120199
Download: ML20205B692 (21)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Report N /85-14 Docket N License N DPR-16 Category C Licensee: GPU Nuclear Corporation P. O. Box 388 Forked River, NJ- 08731 Facility Name: Oyster Creek Nuclear Generating Station

, Inspection At: GPU Nuclear Corporation Office (Parsippany, NJ) and Forked River, NJ j Inspection Conducted: May 14-17, 1985 Inspectors: /b MMW k28/8T K. A. Manoly '

date j Lead Reactor Engineer M. E. Nitzel Senior Engineer (EG&G, Idaho, Inc.)

S. L. Morton Engineer (EG&G, Idaho, Inc.)

Approved by: c:tev~

L. Bettenhausen, Chief 7N 85 date

, Operations Branch, DRS Inspection Summary: Inspection on May 14-17, 1985, (Report No. 50-219/85-14)

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' Areas Inspected: Special, announced inspection by a region based-inspector and two NRC contractors from EG&G, at the licensee's corporate office and at the plant, of licensee actions in response to the NRC/IE Bulletins 79-02, Pipe

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Support Base Plate Designs Using Expansion Anchor Bolts, 79-14, Seismic

Analyses for As-Built Safety Related Piping Systems, 79-04, Incorrect Weights-

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For Swing' Check Valves Manufactured by VELAN Engineering Corporation, and i

79-07, Seismic Stress Analysis of Safety Related Piping. This included verifi-

. cation of actions taken and work performed .in modifications precipitated by these bulletin Results: One violations, two deviations and five unresolved items resul'ted from this inspectio DR ADOCK O 2y9

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, DETAILS Persons Contacted G.P.U. Nuclear Corporation

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R. Keaton, Director Engineering Projects D. Croneberger, Director Engineering & Design J. Thorpe, Director Licensing & Reg. Af K. Jasani, Mechanical Analysis Manager R. Ashby, Jr. , Manager, Oyster Creek Projects B. Bader, Corporate QA Program Development & Audit Manger Y. Nagai, Licensing Engineer J. Rogers, Licensing Engineer

,' J. Chardos, Manager, Oyster Creek Planning J. Piazza, Engineer, Engineering Mechanics I_nspection Purpose and Scope The purpose of this inspection was to review with cognizant and

responsible licensee representatives at the corporate office and at the l plant the completeness of their responses to NRC/IE Bulletins 79-02, Pipe Support Base Plate Designs Using Expansion Anchor Bolts; 79-14, Seismic Analysis for As-Built Safety Related Piping Systems; 79-04. Incorrect Weights For Swing Check Valves Manufactured by VELAN Engineering Corpora-tion; and 79-07, Seismic Stress Analysis of Safety Related Piping. The scope of the inspection included a review of engineering design and quality assurance documentation relating to inspection, testing and modifications satisfying requirements and licensee commitments with respect to the bulletins. A walkdown inspection was conducted by the staff to verify piping and support installations affected by IEB 79-02 and 79-14 related activitie The latest revision of the subject bulletins was used to define required actions by the utilit In addition, Temporary Instructions (TI) 2515/28 and 2515/29 were used to further define inspection requirements relative to IEB 79-02 and 79-14, respectively. Applicable sections of the Code of

Federal Regulations (10 CFR 50),were used to provide guidance regarding legal requirements where possible violations may have occurre . Review of Licensee Responses The inspection team reviewed bulletin responses available from NRC file Two transmittals were forwarded to the licensee at approximately two and three weeks prior to conducting the inspection. They included generic and specific questions relating to the' licensee responses in addition to a request for availing documents and personnel for the inspection tea These transmittals are included in attachments I and II to this repor Any apparent items of noncompliance or those requiring further discussion were noted as items to be addressed while at the corporate office or plant site.

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The inspection team reviewed additional material provided by the licensee during the inspection. The material relating to IEB 79-02 consisted of additional procedures governing inspection, testing, maintenance and modification of piping supports, base plates and concrete anchor bolt Sample calculations of concrete anchor bolt loads were reviewed and samples requiring modification were chosen for detailed field inspection and QA/QC documentation follow-up. The additional material relating to IEB 79-14 consisted of documents governing the field walkdown of piping systems and current piping system isometric drawings. Samples of engineering evaluations of nonconformances were also reviewed. No addi-tional information, specifically dealing with IEB 79-04 or 79-07, was examined at the corporate offices. The pertinent documents described above for IEB 79-02 and 79-14 are listed in the following table , . _ __ _ . _ _ _

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TABLE 1. ENGIliEERING DOCUMENTATION REVIEWED

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DOCUMENT DESCRIPTION

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PP-002 Jersey Central Power & Light (JCP&L)

procedure for the implementation-of IEB 79-0 OCNGS IEB 79-02 JCP&L letter to the USNRC, dated 12-7-79, regarding IEB 79-02 action items 2, 4, 5, 6, and Bergen-Patterson Pipe Support Co.

j letter to JCP&L, dated 4/25/79, regarding original support desig i .

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Transmittal of seismic analysis information for the emergency service i water (ESW) system from Burns & Roe, Inc. to JCP&L dated 1/8/80.

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Field walkdown checklist for implementation of IEB 79-14

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IEB 79-02 inspection and test reports for ESW piping system supports SW-SN-5, SW-SN-45, SW-2-49, SW-2-H10, SW-2-H10A, SW-2-H13, SW-2-H15, SW-SN-3, SW-SN- , Rev.'l Procedure for the modification

, and repair of pipe support ~ Procedure for the. relocation of

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pipe supports.

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PP-001 Procedure for the implementation of IEB 79-1 .31.309 Thermal expansion analysis of-line 6-ND-10 performed in response to an IEB 79-14 discrepancy.

l 8.31.208 Containment spray system pipe support calculation performed in. response to an IEB 79-14 discrepanc Guide, anchor and rigid pipe support data sheets from restart certification

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TABLE 1. ENGINEERING DOCUMENTATION REVIEWED CON DOCUMENT DESCRIPTION

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Summary description of the Oyster's Creek pipe support inspection program for the 1984 restart certificatio Summary list of inspections, specifications, exceptions and deficiencies for the 1984 restart certification progra General description of 1984 restart certification progra Thermal analysis of emergency condenser line 10"-NE-S from the reactor to NE01-A, Special No. 79-31 Procedure for inspection test and installation procedure for concrete expansion anchor bolts in seismic piping system at OCNG Serial NO. 048 Letter addressing response to IEB 79-02 from Burns & Roe, Inc. and attachment including index of baseplates designed by Atlanta Engineering which were reviewed in compliance with IE Bulletin 79-0 Special No. 79-31 Status report on closed cooling water system testing of hangers attached to wall or ceiling by expansion anchor Special No. 79-31 Status report on emergency condenser system hangers attached to wall or ceiling by expansion anchor Special No. 79-31 Status report on poison piping system hangers attached to wall or ceiling by expansion anchor TR-3501-1 Technical Report by Teledyne on Summary (Vol. 1) response to NRC IEB 79-02 for baseplate /

concrete expansion anchor bolt TR-3501-1 Verification of baseplate and post (Vol. 2) processing programs for baseplate structural analysis system _.

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TABLE 1. ENGINEERING DOCUMENTATION REVIEWED CON l DOCUMENT DESCRIPTION  !

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Record of 79-14 hanger repair for liquid poison system hangers No. NP-2-R1, NP-2-R2, NP-1-R4, and NP-1-R ,

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I TABLE 2. IEB 79-02 SUPPORTS REVIEWED FIELD CALCULATIONS SYSTEM SUPPORT NUMBER INSPECTED REVIEWED CRD NC-2-H21 No Yes

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ND-7-H2 No Yes

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CC-3-R4 No Yes

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411-R3 No Yes

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CH-2-SN-40 No Yes SDC NU-1-H7 Yes No ESW SW-SN-11 Yes No ESW SW-SN-10 Yes No ESW SW-SN-8 Yes No ESW SW-2-H5 Yes No ESW SW-2-H3 Yes No 1 ESW SW-2-H2 Yes No j

ESW SW-2-H6 Yes No

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i CRD = Control Rod Drive SDC = Shutdown Cooling

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ESW = Emergency Service Water

" Field Inspected" indicates that the support was verified for agreement with current drawings during the plant walkdown by the staff.-

" Calculations Reviewed" indicates that structural adequacy calculations,.

walkdown data sheets and other related documentation were examined.

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Findings: The review of the documentation described above resulted in the following inspection findings:

(1) Further discussions with the licensee's representatives answered questions regarding IEB 79-04. The response to this bulletin was considered adequat , (2) The licensee's response to Bulletin 79-07 was considered adequate and complete.

4 (3) The Oyster Creek Final Safety Analysis Report (FSAR) does not address the requirements for safety-related components and piping supports. Activities conducted in conjunction with IE Bulletin 79-02 and 79-14 were performed either with no documented engi-

neering procedures and specifications or with procedures and instructions which were not controlled in accordance with the Operational QA plan as stated belo Examples of nonconformance in the first category include

document No. 8.13.309 - thermal expansion analysis of line No. 6-ND-10; document No. 8.31.208 -

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containment spray system pipe support calculations; document for thermal analysis of emergency condenser line 10-NE-S from the reactor to NE01-A, B. Examples of nonconformance in the second j category include: Field walkdown checklist for implementation of IE Bulletin 79-14; 80-44 - procedure for the modification and repair of pipe supports; 80-45 procedure'for the relocation of pipe supports; PP-001 procedure for the implementation of IEB 79-14; Special No. 79.31 procedure for inspection test and installation procedure for concrete expansion anchor bolts in seismic piping systems at OCNGS.-

Section IV, Part 1, Subsection A, Revision 2 of the JCP&L/GPU Operational Quality Assurance Plan states in part, "the Manager-Generation Engineering is responsible for controlling design work and administering the design control activities related to modifications or addition of structures, components and systems".

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Section X, Part 1, Subsection F, Revision 2 of the JCP&L/GPU Operational QA manual also states, in part, that the Manager-Generation Engineering is responsible for the issuance and approval of specifications, drawings and installation require-ments associated with modifications and additions...".

Performing the above identified activities either without the required procedures or with uncontrolled documents is a viola-tion of JCP&L/GPU Operational QA plan and Criterion III of Appen-dix B to 10 CFR 50 which require the establishment of measures to assure that applicable regulatory requirements and design basis for important to safety structures and components are correctly transferred into specification, procedures and in-

structions. (219/85-14-01).

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(4) Numerous records related to Bulletins 79-02 and 79-14 were not provided in sufficient time for review dsring the inspectio Examples include all records of inspection by Quality Control personnel of activities related to IE Balletins 79-02 and 79-14; all marked-up walkdown drawings used to document as-built confi-gurations of piping systems, pipe supports, base plate and anchor bolts;- and IE Bulletin 79-02 evaluation records which identify the categorization of base plate and anchor bolt configurations according to geometrical similarities and provide the basis for selection of the governing case in each category for the purpose

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of detailed finite element analysis. Requests for the required information were provided to GPU two to three weeks prior to the beginning of the inspection (Attachments I & II to this report).

Subsequently, during the meeting on June 25, 1985, the foregoing

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missing records were retrieved and made available for NRC review.

1 However, these records have not been fully reviewed. This-item l ,

is unresolved pending final N2C review. (219/85-14-02).

j (5) Bulletin 79-02 states.that the functionality of each Seismic Category I system be assured. Action Item No. 4 states that a

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test program to verify correct concrete anchor bolt design j attributes be established in the absence of adequate quality

control documentation. Information contained in Attachment 8 of I the licensee's document, " Review Of Conformance To USNRC IE Bulletins 79-02 And 70-14" indicates that only seven of twelve seismic Category I systems were tested. Furthermore, no assur-ance of functionality based on completed tests nor an adequate statistical confidence level was provided for the seven systems l

that were included in the testing program.

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Bulletin 79-02 also states that the effects of base plate flexi-bility be included in the calculation of concrete anchor bolt (CAB) loads. The licensee's approach to fulfilling this requirement was to_ assign each pipe' support using CAB's to a representative type and category which would.then be subjected to an analysis that was intended to envelope all supports in the

respective category. However, the licensee was unable to demonstrate the applicability of the categories used in the calculations to all. existing pipe support configurations. The licensee was also unable to demonstrate the correlation of the categories-to the individual supports. Thus,-there was no assurance that conservative bolt load criculations were per-formed or that individual supports were placed in the proper category.

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On the final day of the inspection, approximately one hour

before the exit meeting, a document was presented to the staff, which, according to the licensee, contained the missing informa-tion regarding the categorization of pipe support baseplates for performing the baseplate flexibility analysis. This document will be reviewed in a followup inspection to verify its adequacy.

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Nevertheless, the licensee's failure to meet the conditions of Action Item No. 4 of the bulletin regarding the initiation and conduct of the testing program of concrete expansion anchors i is a deviation from the commitment in the July 6,1979, letter from JCP&L/GPU to the USNRC (219/85-14-03).

l (6) IEB 79-14 addresses the accuracy of information used in the seismic analyses of safety-related piping systems. Specific-

ally, the design specifications, drawings, and other data used i for these analyses should accurately reflect the as-built i condition of the respective systems. Among the bulletin i conditions were the following

_ Identify the piping system parameters which were input into the seismic analyse Identify the specific design documents from which these values were obtained.

, Establish acceptance criteria which as-built values of these parameters must mee No information was provided during this NRC inspection that clearly identified the design documents applicable to seismic analysis of piping systems. Although the licensee did establish an >

!_ inspection checklist, no evidence was provided to.show that this checklist was applied in a controlled and consistent manner and that appropriate acceptance. criteria were established.- '

i The licensee's failure to meet the bulletin conditions as described above is a deviation from the commitment,in the  !

August 1, 1979, letter from JCP&L/GPU to the USNRC (219/85-14-04).f

Bulletin 79-14 also states that the safety-related piping systems be inspected to demonstrate conformance with the inspection ele- ,

ments established to show correlation with design. Contrary to this requirement no documentation of any marked up drawings, field inspection data sheets or other first generation-documents resulting from an inspection were provided for review. The documents were presented during the June 25, 1985, meeting

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discussed in paragraph. (4) " Findings", of this report. This

item is also unresolved and included with the previously

referenced item (219/85-14-02).

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(7) The adequacy of the baseplate and bolt evaluation for support

, No. NC-2-H21 (CRD system) could not be determined. A verifi-cation of the correctness of the analysis (i.e., baseplate dimen-sions, bolt size and locations, correct loading, correct struc-tural steel geometry, etc.) is needed before the su'pport adequacy can be determined. This' item is unresolved pending the licensee's

,! reev41 uation and NRC review (219/85-14-05).

(8) A verification of the engineering disposition for nonconformance report calculation No. 8.31.208 was not provided. This calcula-tion considered the additional weight on support NQ-2-H44 due to '

the identified change in pipe size from 10 inches to 14 inche However, the increase resulting from seismic inertia loads was ignored for the local pipe nozzles and pipe stresses. The seis-mic adequacy of the 14 inch pipe must be addressed before this nonconformance evaluation can be considered complete. This con-

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cern should also be addressed for similar type nonconformance calculaticns. This item is unresolved pending licensee response

and NRC review (219-/85-14-06).

(9) No documentation was available to verify the conservatism of the original seismic span tables. Even though some reanalysis work was performed in the IEB 79-14 effort, and a more comprehensive analysis was performed for the Mark I torus modifications, this does not assure the adequacy of remaining safety-related piping which was originally qualified by the span tables. Five piping

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system reanalyses were performed during the integrated assess-ment of the Systematic Evaluation Program (SEP) topic III-6 regarding Seismic Design Considerations. Records relating to these evaluations were not provided for review during the inspection. However, the adequacy of the-as-built drawings utilized for the SEP analyses is called into question as a result of this NRC inspection. The uncertainty concerning some i of these drawings was identified during the staff's walkdown verification of the Emergency Service Water System (ESW) piping

.' and supports. Numerous discrepancies were noted between the drawings and the actual installations (see the finding in Section 5 of this report). Unless these five analyses can be shown acceptable, additional information should be provided to

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ensure the seismic adequacy of the safety-related piping systems which are still qualified by the span tables. This item is unresolved pending the licensee response and NRC review (219/85-14-07).

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(10) Pipe support analyses of supports with frictional loads did not include these loads in the support evaluations. Frictional loads can potentially increase support loads to an unnacceptable i

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level in the support steel, baseplate, and concrete anchor bolts if not considered as part of the design of the support. Quali-j fication of pipe supports subject to frictional loading should be demonstrated. This item is unresolved pending license response and NRC review (219/85-14-08). '

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Licensee Response To Items Above: The licensee's representatives acknow-ledged the information leading to violations, deviations and unresolved

items described abov During the course of the inspection, several discussions were held to clarify the areas of concern. The licensee

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representatives indicted during the management meeting of May 20, 1985, that an evaluation of these findings is underway. Results of the licensee's evaluation are expected to be reported to the regional staff

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hetw9en two to three weeks from the above date.

. Verification Walkdown and Independent Measurements Inspection

A. physical inspection of portions of plant systems selected by the NRC inspection team was conducted. The purpose of this walkdown was to verify l samples of piping systems and supports for compliance to as-built condi-i tions as described in the licensee's documentation and to verify repairs

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or modifications to piping, pipe supports:and/or baseplates required to i conform to the subject bulletins. The verification effort was conducted using visual inspection in addition to performance of independent measurements of selected piping and support components. The piping i systems listed in Table 3 were examined during this inspection.

l i Findings: In addition to the the walkdown inspection of the piping

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systems noted in Table 3, a general inspection of several other areas of the plant was conducted. This part of the plant walkdown was done without

. detailed drawings since it was an effort to identify areas of concern i chosen on a random basis.

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TABLE PIPING SYSTEMS FIELD VERIFIED SYSTEM LINE DRAWING LOCATION a

ESW 14"-SW-2 H0102 Reactor Building North ESW 14"-SW-2 H0102 Reactor Building South S. "-NU-2 JCP-19434 Reactor Building Sht 3/3 "-NU-1 JCP-19434 Reactor Building Sht 1/3 Emergency Service Water Shutdown Cooling l

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i The following findings resulted from the plant walkdown: The emergency service water line 14"-SW-2 (North) was examined in detail between the containment spray heat exchangers (1-1 and 1-2) and support SW-SN-8. The discrepancies listed below were note *

Valve V-3-88 was not located as shown on the "as built" drawing (Burns & Roe H0102). The valve is shown mounted in a vertical run but was actually found mounted in a horizontal run. The actual location is approximately seven feet away from the location shown_on the drawing. Also, the valve operator (Limitorque) was found mounted upside down with oil leaking through the electrical junction bo *

The setting of spring hanger SW-2-H4 on the spring scale was not readable since it was painted over.

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Support SW-SN-11 had what is typically considered to be excessive gaps. This support is a box frame guide. Excessive gaps between the pipe guides and frame members would not allow proper function of the support in a seismic even In addition, a 7/16" to 1/2" gap area was identified in the vertical direction between the bottom of the pipe and the support member. This indicated that the support did not provide a dead weight restraint to'the piping as shown on the support drawing. The anchorage of this support is also questionable. The '

frame members transfer loads to the anchor bolts through angle sections. The vertical leg of these angle sections will be subject to weak axis bending when subjected to a lateral support load. It is questionable whether the stiffness of these angle legs was accurately accounted for in the piping analysis. An acceptable state of stress should also be demonstrated in the support analysis.

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Support SW-SN-10 had what is considered to be excessive gaps. This support is a box frame guide. A 1/2" gap was identified between the pipe and the support vertical member of the north sid In addition, a 3/16" gap was measured between the bottom of the pipe and the support member. This was an indication that the support does not provide a dead weight restraint to the piping as shown on the support i drawing. The location of this support was also found to be different j from that indicated on the drawing by 11 inche *

The location of support SW-2-H7 was found to be three feet seven

inches different than that indicated on the piping isometric drawing.

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14 The shutdown cooling line between penetration X-7 and pumps NUO2-A,B, and C was examine It was found that two concrete anchor bolts in the west baseplate of support NU-1-H7 had less than full thread engagement with their respective nuts. While the configuration of the piping matched that shown on drawing JCP-19434, Sheet 1/3, it did not match the config-

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uration seismically analyzed by Burns & Roe as shown in Attachment 12 to

, JCP&L/GPU document, " Review of Conformance to USNRC IE Bulletins 79-02 aad 79-14".

l ' Support NO-2-R1 was found to have excessive gaps between the pipe and

, frame member The support is a bilateral restraint of a vertical piping run (6" diameter) on the containment spray line. 1/2" and 11/16" gaps were measured in one horizontal direction while a 1/8" and 11/16" gaps were measured in the other directio A support structure was found at the 23 foot level near the south wall of the reactor building. This pipe support structure is intended to be a pipe anchor, however, it will not provide six degrees of restraint. The

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use of.this structure in the piping analysis should be verified and fts

! stiffness values demonstrated. There was no support mark available for i reference but the location is near valve V-21-13, ,

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j The stanchion on support NC-2-H60 was foun'd to be offset such tnat direct ,

loading of the spring was not being achieve '

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. The vertical piping from containment spray pumps A and B (located on l

elevation -19 feet) does-not appear to have adequate seismic restrain .

j The extension rod (two inch pipe) on support NQ-2-53 (East-West snubber at. elevation 15 ft.) is extremely long. The adequacy of this rod under '

buckling loads and reduced capacity of the snubber for this installation should be addresse An unmarked support near valves V-21-75 and V-21-76 appeared deformed.

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The W section comprising the main member of this support was attached to the baseplate in a rotated position. Also, the bottom flange of this W

.; section was deformed downward. The cause of the permanent deformation should be determine <

! Two concrete anchor bolts in the baseplate of support NQ-2-S14 (snubber 487-R5) located near valve V-21-5 had less than full thread engagement with their nut Information addressing the discrepancies noted above should be provided by the licensee. This item in unresolved pending the licensee's response and NRC l review (219/85-14-09).

License Response To Above Unresolved Items: The license's representatives acknowledged the unresolved items noted above. The licensee agreed to address these findings and report'the results of the. evaluation along with the other finding in Section 4.0 abov ,

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15 Review of Quality Assurance and Quality Control Activities

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During the time period of the bulletins, all QA/QC activities were con-ducted under the JCP/GPU Operational Quality Assurance Plan. The present licensee (GPUN) indicated that the involvement of QA/QC in bullet;n's

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related activities was very limited. Quality assurance personnel were only involved in the review of procedures and audit of contractors, whereas Quality Control's involvement had focused on implementation of hardware modification Very few records were available for review by the staff during this inspection. This further confirmed the absence of effective quality assurance and quality control' participation in bulletin activities.

It should be noted from Section 3. above that none of the QC inspection records were available for review during the inspection (Violation 219/85-14-02). However, the licensee indicated.in a follow-up telephone conversation that some of the missing QC records were retrieved after the inspection. The following records were reviewed:

• Procedure #2002: Generation Department Procedure Nonconformance and Corrective Action Report (NC. JAR).

0-CCM-82-06: General Public Utilities Nuclear Company audit report of Burns & Roe, Inc., Quality Assurance-Progra *

RFW-0117: GPU-Oyster Creek outage design modification control audit

report.

9 No violations were identified in these record . Definition of Unresolved Items Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable, violations or deviations Unresolved items identified during this inspection are discussed in Paragraphs 4 and 5 abov . Exit Interview The EG&G team participated in an exit interview conducted by the NRC Inspector and attended by licensee and A-E personnel (denoted in Paragraph 1).

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The NRC inspector summarized the inspection findings and the licensee acknowledged these comments. Notes pertaining to the inspection were

, provided to the NRC inspector by the EG&G team and were included in the exit interview; however, at no time during the inspection did the inspector or the EG&G team provide written material to licensee personnel.

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GENERAL QUESTIONS APPLICABLE TO IE BULLETIN RESPONSE WORK BY LICENSEES ,

i How was QA/QC involved in overall effort to satisfy bulletin requir,ements.

How was overall engineering effort organized, coordinated and directe >

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3. ' What AE/ Consultant firms (if any) were used/Io develop bulletin ,

responses. What was the extent of their _invo.lvement (scope of work).

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. 'What personnel qualificationsi and/or special training was provided (or ,

required) for personnel involved in bulletin wor '!

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, AE/ Consultants 5' Oescribe overall design control & QA procedures (i.e., how does QA interact with engineering).

i Describe overall plant modification procedures, Maintenance Work Design Changes *

j Major Modifications j ,

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. > Describe,L and provide for review, procedures. followed to account for

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increased building structure loads due to increased support loads result-ing from bulletin reanalyze / ;

> Describe procedure followed to incorporate new .loaps or other design changes in AE/ Consultant work in progress or completed (design control related to questions 3 and 5 above).

Provide Documentation of Corporate Office Audit of AE and/or Contractor

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Design Activitie . Provide Documentation of Corporate Office Audit / Surveillance of Plant Specific Activities.

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Attachment 1 -2-I&E USNRC REGION I BULLETIN AUDITS GUIDANCE FOR LICENSEES All aspects of work related to the audited bulletins will be subject to review at the discretion of the inspection team. Therefore, the licensee should have the following information readily available: Engineering procedures and instructions governing organization and direction of engineering effort This includes design control, verification and all QA aspect QA procedures for design and modification control plus any special QA procedures specific to the subject bulletin Specific engineering or test procedures relating to the subject bulletin Complete documentation covering all bulletin work. Examples are:

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Concrete Anchor bolts (CAB) torque test or load tests results

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Base plate flexibility criteria

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CAB verification walkdown results (size, embedment, etc.)

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IS0s or "walkdown sheets" for as-built verification of piping systems

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Maintenance orders or design change packages for modifications required by bulletins (systems selected by inspection team)

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Sample calculations and/or computer input for support / baseplate analyses and piping reanalyze . Cognizant management, engineering and QA personnel should be available as needed by the inspection team to answer questions regarding the documentation being reviewe . Responsible engineering and/or maintenance personnel should be available as requested by the inspection team or required by plant procedures to act as escorts during walkdown verificsif on . Additional information pertinent to the reviews may be requested by the inspection tea !

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USNRC REGION I e-BULLETIN FOLLOW-UP INSPECTION OYSTER CREEK GENERATING STATION i

1 The following requests for information are based on licensee submittals l contained in NRC Region I files.

i IEB 79-02 4 Referring to the July 6, 1979, submittal:

< What systems were subject to IEB 79-02 action? Define the category designations used in Tables II and III.

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• What method or procedure was used to control the categorization of 4 base plates? Provide governing documentatio [ Provide a tabulation showing system, s oport, type and category or other documentation used to account for each subject support during

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i the reanalyzes, 1

. Provide the analyses used for Baseplate / Bolt Qualificatio l What organization performed the reanalyzes? Provide the procedure (s) governing this wor Provide samples of corrective actions uadertaken for those supports with baseplates or concrete anchor bolts (CAB's) not meeting

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allowable criteria. Samples should be taken from systems such as:

Main Steam l Core Spray Service Water (Emergency)

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where possibl Provide the procedure (s) used to qualify small bore piping supports. Samples of applicable documentation (calculations,

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charts, etc.) should be provided.

) Provide samples of analyses of containment penetration supports as

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mentioned in 5.6 of the attachment.

Provide a copy of the procedure (s) used to perform CAB pull tests and inspections.

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Attachment 2 -2-i

Provide data sheets or other applicable documentation resulting from the pull tests and inspection Item "b/" under the response to action Item 4 states that sampling method "a" in the bulletin supplement was followed (i.e., test a minimum of one CAB in each baseplate). However, a later submittal states that some supports were not tested due to inaccessibility (see August 3,1979, Ross-Grier letter). Clarification should be provided as to the exact extent that testing was complete If any supports were not inspected due to inaccessibility, provide a listing of:

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• Each support designation

The applicable piping system

The basis for functional qualification of each affected system Was the follow-up or " Final" report referred to on Page 2 of the August 3,1979, Ross-Grier letter submitted? If so, provide a cop . Provide response to Action Items 5 and 6 of Revisien 2 of IEB 79-0 .

IEB 79-04 1 Referring to the May 1,1979, Finfrock-Grier letter:

' Were any reevaluation calculations performed using the actual valve weight? If so, provide a copy for review.

1 Was the subject line reanalyzed under IEB 79-14?

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IEB 79-14 Provide a list of all safety related systems subject to IEB 79-14 action. The list referred to in the August 1, 1979, Finfrock-Grier letter includes a generic category of " Class I piping < 10" in diameter." All systems should be specifie .

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Provide a copy of the information submitted on July 9, 1979, that is

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referred to in the August 1,1979, Finfrock-Grier lette . Provide a copy of the procedure (s) used to perform the plant walkdown . Provide a copy of the procedure (s) used to process nonconformances.

i Provide copies of processed nonconformances and the associated engineering justifications for the dispositions.

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Attachment 2 -3- Provide marked up isometrics, "Walkdown Sheets" or other documents resulting from the field walkdowns. Samples should be taken from systems such as:

Core Spray Emergency Service Water RHR (Shutdown Cooling)

Main Steam Feedwater Samples should include as much accessible piping as possible for the selected system . Orovide current isometric or the appropriate current drawings corresponding to the samples in 6. abov . Clarify exactly what is meant by "Does not affect system design requirements". Was the two foot support location tolerance applied to all pipe sizes?

10. Provide copies of the piping reanalyzes performed when require . Clarify the difference between the following nonconformance descriptions:

" Hanger not installed as per design documents".

"Not attached to piping as indicated in design documents".

12. Were any systems or portions of systems not inspected due to high radiation or other inaccessibility reason , Since numerous supports were not installed according to the design documents, what efforts have been made to assess fatigue due to possible high stresses and adverse operating conditions experienced by these systems?

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