IR 05000324/1993045

Insp Repts 50-324/93-45 & 50-325/93-45 on 930920-24.No Violations Noted.Major Areas Inspected:Structural Steel Verification Program,Repairs to Unit 1 Drywell Liner,Control of Maint Trouble Tickets & Structural Integrity Items

ML20059D931
Person / Time
Site:	Brunswick
Issue date:	10/21/1993
From:	Blake J, Chou R, Lenahan J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20059D924	List: IR 05000324/1993045 ML20059D920
References
50-324-93-45, 50-325-93-45, NUDOCS 9311030040
Download: ML20059D931 (15)

Text

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UNITED STATES

. / a pra%,'i g4 J NUCLEAR REGULATORY COMMISSION

'g . REGION li '

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-; ) 101 MARIETTA STREET, N.W., SUITE 2900

ej ATLANTA, GEORGIA 303234199

s o;, y*****

Report Nos.:- 50-325/93-45 and 50-324/93-45 Licensee: Carolina Power and Light Company P. O. Box 1551 Raleigh, NC 27602 Docket Nos.: 50-325 and 50-324 License Nos.: DPR-71 and DPR-62 Facility Name: Brunswick I and 2 Inspection Conducted: September 20 - 24, 1993 Inspectors: ,,

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J. J. Lenahan / Date Signed 1 '

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'R. C. Chou '

Date Signed :

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Approved by: '

f _ h /3 J. J. Blake Ch/ef Date Signed ,

Materials d processes Section Engineerin Branch ,

Division of Reactor Safety l

SUMMARY ,

Scope:

-l This special announced inspection was conducted in the areas of the miscellaneous structural steel verification program, repairs to the Unit I drywell liner, control of maintenance trouble tickets, licensee action on previous inspection findings, and short term structural integrity item Results: ,

In the areas inspected, violations or deviations were not identifie '

No strengths were identifie A weakness was identified in the licensee's maintenance program regarding  ;

control of trouble tag ;

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REPORT DETAILS Persons Contacted

• R. Anderson, Vic2-President,-Brunswick Nuclear Plant ,

• H. Beane, Quality Control Manager

• M. Brown, Unit 1 Plant Manager T. Eason, Quality Control Supervisor

• R. Grazio, Site Manager, Nuclear Engineering Department (NED) *

L. Grezeck, Project Engineer, Hiscellaneous Steel (NED)

• J. Harness, Supervisor, Nuclear Assessment Department R. Knott, Principal Engineer, NED

• W. Levis, Manager, Regulatory Affairs

• J. Purkis, Manager, Projects '

• G. Thearling, Senior Specialist, Regulatory Compliance

• R. Tripp, Supervisory Civil Engineer, NED

• S. Vann, Brunswick Civil Unit Manager, NED

• C. Warren, Unit 2 Plant Manager

• E. Willet, Manager, Project Management K. Williamson, Project Engineering Manager, NED Other licensee employees contacted during this inspection included engineers, technicians, and administrative personne Other Organizations

• Bizeck, Civil / Structural Consultant, Prism A. Marrow, Manager, DTOP, Bechtel NRC Resident Inspectors R. Prevatte, Senior Resident Inspector P. Byron, Resident Inspector

• M. Janus, Resident Inspector

• Attended exit interview , Review of Licensee Corrective Actions to Resolve Structural Deficiencies In a letter dated July 23, 1992, Serial: NLS 92-160, Subject: Reply to Inspection Report Nos. 50-325/92-12 and 50-324/92-12, the licensee '

connitted to completion of short term corrective actions listed in Enclosure 3 of the letter prior to startup. The inspector examined the short term corrective actions listed belo .

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2.1 Item A-2, Walkdown Inspection to Examine STSI Items Affected by Corrosion Licensee Connitment

" Perform a third-party walkdown of non-pipe support, short term structural integrity (STSI) items and pipe supports in areas with high ;

corrosion potential to validate design assumptions. Address any identified deficiencies in accordance with the methodology in Enclosure 2."

i Discussion Short term structural integrity (STSI) items are those identified by licensee personnel, which, after evaluation by the Nuclear Engineering Department (NED) are determined to be operable, although they do not meet design criteria established by the FSAR. The operability reviews are performed in accordance with Design Guide II.20, " Civil / Structural Operability Review". Design Guide (DG) 11.20, was reviewed and approved by the NRC Office of Nuclear Reactor Regulation (NRR).

The methodology in Enclosure 2 is the licenses's program for evaluating deficiencies to determine if they need to be corrected prior to restart or if corrective action can be delayed until a future outage. CP&L Procedure PN-30, " Integrated Recovery Methodology", specifies the requirements for the operability review The Unit 2 STSI items affected by corrosion were inspected during the inspection documented in NRC Inspection Report Number 93-15. Short term corrective action Item A-2 for Unit 2 was closed during this inspectio During the current inspection the inspector reviewed the open STSI items for Unit 1. With the exception of the STSI items relating to the ,

service water pumps and diesel generator cooling water piping, the licensee plans to complete corrective action on all Unit 1 STSI items prior to restart of Unit 1. There were approximately 12 Unit 1 STSI items still remaining open as of the current inspection. None were affected by corrosion. Therefore Item A-2 can be closed for Unit Conclu11gn Short term corrective action Item A-2 is completed and acceptable for restart of Unit 1.

2.2 Item C-1, Hiscellaneous Structural Steel Verification Program Licensee Commitment

" Complete Unit I and Unit 2, Drywell Phase 2, miscellaneous steel walkdowns. Complete preliminary bounding load studies. Address repairs, as required in accordance with the methodology in Enclosure 2."

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Discussion  :

Miscellaneous structural steel consists of platforms and other beams or ;

columns, which provide personnel access and support for piping, electrical raceways and conduits, HVAC ducts, instrumentation, and other equipment not supported from the main building structures. Numerous deficiencies in miscellaneous steel had been identified by either the licensee or NRC, including lack of design calculations, lack of as-built drawings, missing bolts and welds, incorrect size members, undersized welds, missing members, and other construction deficiencies. The licensee retained Bechtel Power Corporation to perform walkdown ,

inspections, prepare as-built drawings, and perform design calculations to qualify the miscellaneous stee The Bechtel structural steel verification program, which is called the <

Miscellaneous Steel Verification Program (MSVP), is a two phase project with the purpose of establishing a high confidence that the miscellaneous steel is adequate for operation. The Phase I program was a walkdown inspection to identify and evaluate any irregularities which could affect the integrity of the structures. The Phase II program involved obtaining detailed field measurements to update design documents, prepare as-built drawings, performance of a detailed structurai analysis, and preparation of a load tracking program to identify the magnitude and location of load The Phase 11 program has been completed in the Unit I drywell. The Phase I program was completed for the miscellaneous steel in the Unit I reactor building. The inspector previously inspected the' Unit 1 MSVP during inspections documented in NRC Inspection Report numbers 50-325,324/93-25 and 93-31. The Unit 2 MSVP, item C-1, was closed out in NRC Inspection Report number 50-325,324/93-15. During the current inspection, the inspectors reviewed Unit 1, Phase I and Phase II walkdown documentation, reviewed design calculations, and' examined repairs to structural steel in the Unit I reactor buildin Details of the inspection follows below:

2.2.1 Review of Walkdown Documentation The inspector randomly selected and reviewed field inspection data packages which document the results of Phase I and Phase II walkdown inspections. Acceptance criteria utilized by the inspector were Bechtel Procedures WDP-001, " Phase I Engineering Walkdown Procedure for Reactor Building Miscellaneous Steel," Revision 2, and WDP-002, " Phase II Walkdown Procedure for Reactor Building Miscellaneous Steel and Drywell Platform Steel," Revision Phase 11 packages reviewed were those which document walkdown inspections performed on the elevations 17 and 80 drywell platform steel. Phase I packages reviewed were those which document the Phase I

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walkdown performed in the North RHR room. The inspector randomly selected four Phase I walkdown packages, walked down the structural steel, and compared the data in the walkdown packages with actual field conditions. No discrepancies were identifie The inspector reviewed the results of an audit completed by Bechtel Quality Assurance engineers, Audit 21963, conducted January 11 - 27, 199 This was an audit performed on 64 randomly selected components from five Phase II packages. A few minor errors were identified, two-thirds of which were conservative error The inspector also reviewed the results of independent walkdown inspections by CP&L engineers. These walkdowns were performed to assess the adequacy of the Bechtel walkdown inspections and inspection documentatio The results of the walkdown inspections performed by licensee engineers showed the number of errors in the Bechtel data was lo An independent consultant reviewed the results of the Bechtel QA audits, the results of independent walkdown inspections performed by licensee engineers, and the independent field verification walkdown inspection performed by Bechtel field engineers. Bechtel field engineers performed independent field walkdowns on approximately 11 percent of the Unit I walkdown packages to verify in the field the walkdown data. The consultant's findings regarding the results of the various independent walkdown verifications were the same as those found for the Unit 2 walkdown dat These were as follows:

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Given the large number of components and the corresponding large number of inspection attributes for each component, some error in the field inspection data is inevitable,in spite of the use of a detailed checking proces The weld measurement errors were biased on the conservative sid There were only slight differences in weld measurements between various inspector The overall error rate was low, and will not affect the final walkdown result The inspector concurred with these conclusion Based on the operability reviews performed for the irregularities identified on Units 1 and 2, the inspector concluded that random errors in the field inspections data would have a negligible effect on the results of the MSV . . . __ ._ ..- _ _ . __ .

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.2.2.2 Review of Evaluation and Modification Calculations for Unit 1  :

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The process of the irregularity evaluation for the Unit 1. MSVP is to review the irregularities against the standard calculations which - :

evaluated the common irregularities and set forth the acceptance  :

criteria based on the Unit 2 experience. If the irregularities did not '

meet the acceptance criteria for the stendard calculations, the irregularity was evaluated individually. The final disposition for each-irregularity was either use as is, rework, or modify. Each walkdown package has an evaluation calculation to evaluate irregularities and determine final disposition. If a irregularity required a modification,

a modification calculation was generated to qualify the proposed modification and a Design Change Notice.(DCN) was issued for the ,

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modification. A Field Design Change Notice (FDCN) was issued if the final modification deviated from DC The modification calculations were initially issued as Revision A when a '

DCN was issued. The subsequent revisions for the modification .

calculation will be Revisions B, C, etc. during construction. After the field modification was completed, a set of final as-built drawings were attached to the modification calculations and the calculations were' ,

reviewed by design engineers. The calculations were then reissued as Revision A statement that the QC verified modification sketches for the calculation were reviewed and verified to be consistent with the-design criteria was noted in the final closed ~out calculation packag The inspectors. reviewed the calculations listed belo Review of Evaluation and Modification Calculation ,

Calculation Evaluation or

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Revision Discrepancies /

Number Modification Comments IRB1-ll48 0 Evaluation -

1RB1-ll50 0 Evaluation

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IRB1-ll56 0 Evaluation 1RB1-ll59 0 Evaluation  ;

1RB1-1181 0 Evaluation IRB1-ll82 0 Evaluation IRB1-ll83 0 Evaluation IRB1-ll87 0 Evaluation

1RB1-1191 0 Evaluation 1-l 1RB1-1208 0 Evaluation

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1RB1-1129 0 Modification The attachment B for walkdown irregularities was not referenced in this calculatio IRB1-1130 0 Modification The evaluation calculation was .

not reference IRB1-1209 0 Modification Same as above IRB1-1217 0 Modification IRB1-1233 0 Modification A 5/16" fillet weld size was used for a t" thickness of base meta Also evaluation calculation was not reference RB1-1240 0 Modification The above calculations were reviewed for completeness, accuracy, adherence to design criteria and procedural requirements, acceptability of calculation methods with American Institute of Steel Construction

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(AISC) code criteria, and good engineering practices. All the evaluations and modifications calculations reviewed were determined to be acceptable although the inspectors. identified some minor discrepancies. These discrepancies did not affect operability for

, restar All the evaluation calculations and standard calculations to evaluate the irregularities were based on the short term requirersents for the operability criteria which had been approved by the NRC Office of Nuclear Reactor Regulation. The short term operability criteria uses higher allowables than those specified in the Final Safety Analysis Report (FSAR). The licensee committed to complete the MSVP after restart, using the FSAR as acceptance criteria. This will include re-evaluating the structural steel for Operational basis Earthquake (0BE)

conditions in addition to the Design Basis Earthquake (DBE) criteria specified in the MSVP design criteria. The licensee will establish a' -

schedule for completion of the restart portion of the MSVP in December 1993. The inspector will follow up on completion of the restart portion of the MSVP in future inspections. This will be tracked as Inspection Followup Item 325,324/93-45-01, Post-Restart MSVP completion

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2.2.3 Examination of Repairs to Unit I Reactor Building Structural Steel q The inspector examined repairs to the Unit 1. reactor building structural steel on elevation 50 and in the north RHR room. The repairs involved correction of the deficiencies (irregularities) identified during the MSVP which were repaired under plant modification PM 92-076. The repairs completed under the PM involved installation of new members, new welds, and reinforcement of existing members. The inspector also reviewed records documenting repairs to structural steel. irregularities competed under work request 92-BITH-1. The repairs made under the work request included replacement of existing bolts, installation of missing ;

bolts, tightening existing bolts, repairs to overcuts in beams, ad other minor repair The inspector walked down the reactor building and examined repairs completed under PM 92-076. Acceptance criteria utilized by the inspector appear in CP&L Specification BSEP 248-107 and the design drawing (sketches) listed below. The repairs examined by the inspector and the controlling installation drawings were as follows:

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Elevation 50 platform steel - sketch numbers Sk-92076-C- 1085, sheets 1 and 2 of 2, Sk-92076-C-1088,, sheets 1 and 2 of 2, Sk-92076-C-1097, sheets 1 and 2 of 2, and Sk-92076-C-1105, sheets 1-3 ,

of North RHR room - sketch numbers Sk-92076-C-1023, sheet 1 of 1, Sk-92076-C-1032, sheets 1-3 of 3, Sk-92076-C-1033, sheets 1-2 of 2 and Sk-92076-C-Il03, sheets 1-4 of During the walkdown, the inspector verified the modifications were completed in accordance with design requirement No discrepancies were identifie The records reviewed by the inspector documenting implementation of repairs completed under Work Request 92-BITH 1 included documentation of j installation of new fasteners, tightening ~of nuts on new and existing i bolts (by turn of nut method), repairs to beam overcuts, and visual weld ;

inspection data. No discrepancies were identifie '

2.2.4 Review of Corrective Actions to Disposition MSVP Irregularities !

The inspector randomly selected MSVP walkdown inspection packages and reviewed the licensee's actions to resolve irregularities. The )

inspector reviewed the disposition of the irregularities to verify the irregularities were addressed in accordance with the licensee's design ,

criteria under the MSVP guidelines. Irregularity disposition was l examined for the following walkdown packages I

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1RBB El (0'9") (P-S/2R-4R)

1RBB El (-4' 1") (S-T/2R-3R)

1RBB El (10' 6") (P-S/2R-4R)_

1RBB El (4' 6") (S-T/2R-4R)

1RBD El (60' -1") (P-R/5R-8R)_

IRBB El (8' 6") (S-T/2R-3R)

The inspector verified that each of the irregularities identified in the above packages had been properly di.po:itioned. The dispositions included restoring the irregularity to original design conditions using a work request or plant modification, initiating a modification to comply with design criteria, performing additional evaluations to accept as is,or review by level 1 and 2 screeners to accept-as-is. The inspector concluded that the irregularities-in the above packages had been properly dispositioned for restar Conclusions The licensee has satisfactorily implemented the miscellaneous structural steel program for restart of Unit Short Term Corrective Action C-1 is completed and acceptable for restart of Unit 1. Post restart completion of the MSVP will be followed up as an inspector followup item ,

IFI 325,324/93-45-0 ;

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2.3 Item C-12, Short term structural integrity (STSI) program Licinsee Commitment l l

" Perform a third-party review of the short term structural integrity l program to address evaluation techniques, field validation of critical !

assumptions, and a review of communications from the Technical Support '

organization to the Engineering organization. Address any identified deficiencies in accordance with the methodology in Enclosure 2."

Discussion The licensee's program for third-party review of STSI items was inspected prior to restart of Unit 2 during the inspection documented in NRC Inspection Report 50-325,324/93-20. The licensee plans to complete corrective actions on all Unit 1 STSI, items, except for those relating to the service water pumps and the diesel generator cooling water l piping, prior to restart of Unit Some STSI items which are 1 classified as affecting common areas will also remain open. The !

licensee documented their position in a letter to NRC dated May 28, l 1993, Serial NLS-93-13 I

The inspector reviewed the licensee's program for classification of l deficiencies identified in Unit I during the current extended outag The inspector concluded that the classification of deficiencies was conservative. As stated above, Unit I deficiencies considered as STSI l items were either corrected or will be corrected prior to restar l-l

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I The licensee has identified some additional items in common area These items will be evaluated and may not be corrected prior to restar However, any new STSI item will be subjected to a cumulative effects evaluation per the letter from NRC to CP&L dated October 8,1992,

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Subject: Correction of seismic Qualification Deficiencies, Brunswick Units 1 and 2.

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Conclusion Short term corrective action item C-12 is completed and acceptable for I restart of Unit 1. However the inspector will review any new I uncorrected STSI items identified by the licensee and verify a l cumulative effects evaluation has been completed. This will be tracked as Inspection Follow Up Item IFI 325,324/93-45-02, Cumulative Effects of STSI Items in Common Areas, j 3.0 Repairs to the Unit 1 Drywell Liner NRC Inspection Report 50-325,324/93-02 identified a problem with i corrosion of the drywell liner plate at the intersection of the liner l with the elevation 4'-6" concrete floor, around the entire circumference -)

of the drywell. A violation, item No. 325,324/93-02-01, was identified i regarding failure of the licensee to measure and evaluate the corrosio l After the corrosion problem was identified, the licensee performed I extensive inspections and repairs of the corrosion in the Unit 2 drywell . This work was examined by the inspector during the inspection documented in NRC Inspection Reports 50-325,324/93-02 and 93-15. The Unit 2 corrosion repairs were found to be acceptable for restart of Unit The inspector examined the repairs to the Unit I drywell liner corrosion during inspections documented in NRC Inspection Reports 50-325,324/93-25 and 93-31. A violation item, Number 325/93-31-03, was identified by the inspector, regarding inadequate corrective actions to repair the Unit I corrosion. The inadequate corrective actions involved failure of the

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licensee to repair several areas of the liner plate which were severely ;

corroded. After this violation was identified, the licensee wrote ;

Adverse Condition Report B93-251 to document the problem, and initiated i corrective actions to repair the remaining areas of the liner plate which were corrode The inspector and resident inspectors examined various stages of the corrosion repairs to the Unit drywell liner after Violation 325/93-31-03 j was identified. This included cleaning (sandblasting) of the liner 1 plate, measurement of the remaining plate thickness, additional weld repairs to the liner plate, surface preparation for application of ,

coatings, and coating applications.

l During the current inspection, the inspector examined the completed

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coating of the liner, the new grout which had been placed to replace i

portions of the concrete floor which had been removed to facilitate

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liner repairs, and the elastomer seal which had been applied to cover

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the expansion joint between the liner and concrete-floor. No deficiencies were identified with the completed coating. A few mino defects were identified with the seal, but these will not' affect.its serviceabilit The inspector reviewed records documenting.the. completed repairs to the-liner plate. There records included the following: a QC assessment of coatings application-to the drywell liner, ultrasonic test data which_-

determined thickness of the drywell liner plate, measurement of depth o corrosion (pitting) on the drywell liner plate, weld repair data, QC inspection records for coatings, and records _for grout testing and placement. The inspector also reviewed. work request WR J0.93-AFNW1, Engineering Evaluation Report (EER) .93-0173, which established the acceptance criteria for minimum drywell liner plate thickness, and EER 93-0417 which documented the liner inspection and corrective action Based on review of the quality control records, and inspection of the completed repairs to the drywell liner, the inspector concluded that' the Unit 1 liner corrosion problem had been repaired and was acceptable for restart of Unit Violations or deviations were not identifie .0 Control of Maintenance Trouble Tickets (62702)

During a walkdown inspection in the Unit I reactor building, the inspector noticed a conduit support which had both end. mounting straps missing. The inspector noted that Trouble Tag number 5780'was. attached.-

to the suppo-t and documented' the missing mounting straps. . The trouble tag had been written during the summer of 1992. The inspecto questioned licensee engineers regar' ding the status of this trouble tag and whether the required repairs had been scheduled. As a result of th inspector's questions, licensee engineers stated that apparently a work request (WR/J0) had never been written to initiate the repair wor Licensee engineers wrote a WR/J0, number WR/JO 93-AZPL 1, to install the end mounting strap Further discussions with licensee personnel and review of licensee procedures disclosed that the trouble tag system apparently ~is not well-controlled. The use of trouble tags is covered by CP&L Administrative Instruction, AI-79, " Administration of the Automated Maintenance Management Manual System (AMMS)." Instruction AI-79 recommends attaching a trouble tag to identify a deficiency in the plant. One purpose of the trouble tag is to identify to other. plant personnel that the deficiency had been identified by someone else to' avoid duplicate work orders. After the trouble tag is written, the individual who "

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identified the deficiency is responsible to initiate a WR/J0, or to notify the maintenance group regarding the deficiency (nonconforming  :

condition).

The WR/JO is controlled by CP&L Maintenance Management Manual Procedure OMMM-003, " Corrective Maintenance." Procedure OMMM-003 specifies

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requirements for referencing the trouble tag on the WR/J0, and recommendations for removing the tags after work is complete Discussion with licensee personnel and review of the procedures listed above disclosed several weakness in the use of trouble tags. These were as follows:

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Trouble Tags are not controlled and therefore there is no accountability regarding use of the tag These are no explicit requirements regarding attachment of trouble tags to identified nonconforming conditions. The wording in the procedures indicated attachment of trouble tags is optiona Trouble Tags are not uniquely numbered. They are numbered consecutively from 0001 to 9999. When one batch of tickets is used, a new batch is opened and distributed, starting over with number 000 No list of open trouble tags is maintained. Since the corresponding, WR/JO number (if any) is not listed on the trouble tag, it is very difficult to determine the status of work relating to a particular trouble ta Based on the inspector's observations, it is apparent that requirement for removal of trouble tag specified in OMMM-003 are not consistently implemented. The inspector noted several trouble tags still hanging in the plant for which work was complete r

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Unless a WR/JO is written, corrective action will not be initiated to correct nonconforming conditions. Specific responsibility for writing the WR/JO is not covered by the licensee's procedure The above problems regarding lack of controls for trouble tickets was identified to the licensee as a weakness in their maintenance progra Violations or deviations were not identified.

5.0 Licensee Action on Previous Inspection Findings (Closed) Violation item 325,324/92-10-01, Inadequate Corrective Actions (Civil Penalty) i The licensee responded to the Notice of Violation and proposed Civil Penalty in a letter dated March 12, 1993, Serial: NLS-93-075, Subject: 1 Reply to Notice of Violation and Proposed Imposition of Civil Penalt '

The licensee enclosed a check in the amount of two hundred twenty-five i thousand dollars with their response letter to pay the civil penalty.- -l l

The Notice of Violation civil penalty was issued because the licensee failed to identify and correct deficiencies promptly in construction of safety-related masonry wall in the diesel generator building. Although

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the problem was initially identified by licensee engineers ore than five years prior to Inspection 92-10, no actions were undertaken by the licensee to evaluate and correct the problems until they were prompted to do so by NRC inspectors. After it was determined that widespread deficiencies existed in numerous diesel generator building walls, both units were shutdown. The licensee corrective actions included a comprehensive inspection, design review, and repair program to correct the deficiencies in the masonry walls.- The repairs to the block walls were inspected during previous inspection The licensee also identified other issues relating to degraded plant equipment which they identified as short-term corrective actions in Enclosure 3 to their July 23, 1992 letter, referenced and in Paragraph 2, above. The short term corrective actions have been inspected by various Region II inspectors during the last sixteen month The licensee has also initiated several procedures upgrades and program changes to address corrective actinns. These are doomented in the licensee's Brunswick Three-Year Plan. The implementation of the three year plan was inspected during a special inspection documented in NRC Inspection Report 50-325,324/93-32. The resident inspectors and other Region 11 personnel have also inspected implementation of the three year plan.

5.2 (Closed) Inspector followup Item 324/93-15-01, Followup on Phase II of MSVP for Unit 2 Reactor Building Steel This item was updated to include the Unit I reactor building structural steel and will now be tracked as IFI 325,324/93-45-01, Post-Restart MSVP Completion.

5.3 (Closed) Unresolved Item 325,324/F -31-01, Possible Inadequate Weld Inspection. This unresolved item involved failure of a licensee QC inspector to identify some missing welds on instrument rack H21-P004-002. The licensee initiated ACR S-93-019 to document and correct this problem. The QC inspector had failed to note that a weld specified on the design drawing could not be made due to the installed configuration of the component. The problem was caused by a cluttered and confusing drawing which had been revised several times due to installation problems. Licensee design engineers reviewed the design calculations for the instrument rack and determined that the connections were acceptable in their as-built condition. The drawings were also revised to reflect the as-built condition. Licensee QC supervisors tcvlwed the weld inspections performed by the QC inspector who made the error and reinspected a sample of his work to determine if it was necessary to perform a 100 percent reinspection of his previous inspections. No other discrepancies were identifie Tae cause of this problem was attributed to human error. However, the (onfusing drawing was considered to be the primary cause of the erro Licensee QC supervisors Informed all QC inspeci. ion personnel of the

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a problem, and cautioned them that increased attention to detail is l required to insure all QC inspections are performed correctly. The inspectors were told to contact their supervisors when cluttered or 1 confusing drawings are encountered while performing inspections, j Discussions with QC management disclosed that they will request that.NED- .)

revise cluttered drawing to clarify installation details when they feel H that inspection and/or installation errors could result from using the- )

cluttered drawin The inspector concluded that the problem regarding the. missing weld was an isolated occurrence which had no safety significance.- (0 pen) Unresolved Item 325/93-31-04, Pipe Supports Calculation' Design' l Questions for Unit 1 Service Water System  !

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This unresolved item involved questions on some assumptions or criteria '

contained in calculation No. PS-SW-763-910ll, Rev. 0A for Support Mark number PS-4350. The questions were as follows:

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A dimension of seven times the bolt diameter. was used for the minimum center to center distance for anchor bolt spacing. The seven times the bolt diameter is less that the minimum spacing = ;

t specified in Design Guide DG-II.6. There was no justification referenced in the calculation for this difference. ' Incorrect allowable tension values may have been used in some base' plate design A safety factor of 2 was used to qualify the anchor bolts. ' Design I'

Guide DG-II.6 requires a safety factor of-4 for wedge anchor bolts -l'

for non-safety-related piping requiring seismic support The inspectors discussed the above problems with licensee engineers and-reviewed written justification for the different anchor bolt spacin The licensee concluded that the concrete was still acceptable for this 7d (diameter) spacing at all embedment depths. The method is to divide '

the overlap areas equally based on the numbers of anchor bolts involved and to add to each the divided overlap area back to the unoverlap shear cone area to count as the total shear cone resisting area. Based on the above method, the licensee calculated the concrete tension capacity for 7d spacing, tabled them based on the anchor bolt types, diameters, .

embedded depths, and compared them to the anchor bolt tension allowable :

listed in Design Guide DG-II.1,' Baseplate Design Guide. All wedge anchor bolts including Redhead and Hilti were acceptable. However, all sleeves anchor bolts were unacceptable. Therefore, the licensee

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proposes to revise' DG II 6 to delete the footnote allowing a minimum center to center spacing of 7d. .The licensee will also develop a formula ta reduce the tension allowable for anchor bolts violating the )

minimum center to center spacing ~ requirements specified in Design Guide :

DG-II.1. Two additional problems were identified by the inspector in j calculation No. PS-SW-763-91011:  ;

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On pages 47 and 48 of the calculation, a bending ' allowable value was used without checking the unbraced angle length Page 57 of the calculation was a memorandum to Design Turnover Program (DTOP) staff dated May, 1991. This memo stated that non-Q, seismic supports in boundary condition which one included in safety related stress analysis will not be evaluated for Operational Basis Earthquake (0BE) loadings and Safe Shutdown Earthquake (SSE) loadings will be evaluated to faulted Short Term Steel Integrity (STSI) allowabl The use of a memo to provide design criteria is unacceptabl This problem will be reviewed as part of resolution of violation Item 325,324/93-25-02. This violation was for a similar issue, and the licensee's corrective action for that violation should this. This memo also states that supports in boundary conditions for safety related stress analysis are to be qualified for expansion anchor bolts by using a safety factor of The licensee's current design procedure designates the 8th edition of AISC Manual for the design of the pipe support structural members. The licensee's engineers were unable to tell the inspector which formula in the 8th edition of AISC Manual were used for the allowable reduction of the steel angle due to the extensively unbraced lengths. As a result of the inspector's questions regarding unbraced angle lengths, the licensee is

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considering revising their procedures and design criteria to include the steel angle design methods used in 9th Edition of the AISC Boo The licensee will continue to review the above problems. This item remains open pending further review by NR .0 Exit Interview The inspection scope and results were summarized on September 24, 1993, with those persons indicated in paragraph 1. The inspectors described the areas inspected and discussed in detail the inspection results listed below. Proprietary information is not contained in this repor IFI 325,324/93-45-01, Post Restart MSVP Completion, Paragraph IFI 325,324/93-45-02, Cumulative Effects of STSI Items in Common Areas, Paragraph 2.3.