ML20148R153

Responds to NRC Re Violations Noted in Insp Repts 50-454/87-07 & 50-455/87-24.Corrective Actions:Snubber Reduction Program Performed by Nutech Engineers,Including Identification of Snubbers to Be Deleted or Replaced

ML20148R153
Person / Time
Site:	Byron
Issue date:	01/26/1988
From:	Ainger K COMMONWEALTH EDISON CO.
To:	Davis A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
References
4144K, NUDOCS 8802020111
Download: ML20148R153 (7)

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( v ~7 A33rMiply to P6siO+Ee BoTi67-( / Chcago.1:hnos 60690 0767 January 26, 1988 Mr. A. Bert Davis Regional Administrator U.S. Nuclear Regulatory Commission Region III 199 Roosevelt Road Glen Ellyn, IL 60137

Subject:

Byron Station Units 1 and 2 NRC Inspection Report Nos.

50-454/87007 and 50-455/87024 NRC Docket Nos 50-454 and 50-455 Reference (a): September 16, 1987 letter from J.J.

Harrison to Cordell Reed.

Dear Mr. Davis:

Reference (a) provided the NRC's reply to Commonwealth Edison's response to a Notice of Violation concerning the snubber reduction program at Byron Station. That response requented the NRC to reconsider whEther some of the items in the Notice of Violation were appropriately classi#ied as violations of 10 CFR 50, Appendix B. In reference (a), the NRC concluded the violatior.s were valid and requested Commonwealth Edison to submit a revised response in accordance with 10 CFR 2.201 for vi.olation examples la, Ib, 2a and 2c. Attachment A of this letter contains Commonwealth Edison's revised rasponse to those violations.

Very truly yours, K. A. Ainger Nuclear Licensing Administrator 1m Attachment cc: Byron Resident Inspector 8802020111 080126 PDR ADOCK 05000454 G PDR 4144K j'M :1 -

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l AI3ACHMENT A RESPONSES TO VIOLATIONS VIOLATICN la i

10CFR50, Appendix B, Criterion V, as implemented by CECO Topical Report CE-1 A, "Quality Assurance Program for Nuclear Generating i Stations," and CECO Cornorate Quality Assurance Manual, Nuclear Generating Stations, "Quality Requirements," requires that activities affecting quality shall be prescribed by documented instructions of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions.

Instructionu shall include appropriate quantitative or qualitative acceptance criteria for determining that activities affecting quality have been satisfactorily accomplished. ,

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Contrary to the above, during the course of reviewing safety l related design efforts performed by NUTECH Engineers, the NRC 1 inapectors observed that activities affecting quality were not  !

prescribed by documented instructions in that In.tructions did not address seismic ef fects for all portions of the piping subsystems. Paragraph 4.1.1 of Instruction BYR 19, Revision 2. states that seismic interaction walkdown shall be specified for portions of subsystems where seismic displacement' exceed the original design seismic >

displacement.. by more than one inch.

Correctivp_ Action Taken and Results Achieved This issue was addressed in two distinct phases of the snubber reduction project. First, NUTECH identified snubbers to be deleted or replaced on individual subsysteus. They perf ormed ,

stress analyses and pipe support qualification calculations to l assure that the revised configuration met ASME Code allowables. '

Pipe displacements at all locations were computed as part of this process. This phase culminat90 in the issuance of design  ;

packages, referred to as PFCNs, to the piping contactor. +

Upon completion of the piping modifications, field walkdowns (the second phase) were performed to address the seismic interaction 4 concern. These walkdowns were perf ormed by NUTECH field l engineers under the direction of Commonwealth Edison personnel. i The walkdown program was premised on the fact that much of the l

piping operates well below allowable stresses and with vety small displacements during seismic events. Inspections were focused on l areas having either potential high stress (low margin to j allowable stress limits) or displacements which were either large or significantly increased over the original dispiacement. A key source of information was the displacement dasa obtained f rom the analysis phase.

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The procedure which controlled the scope of the unikdownn was COM-PI-BYR21. This procedure had instructions for identifying portions of piping to be inspected as wel. as methods to resolve potential interferences. The following portions of piping were l covered:  !

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1. Previously identified locations with radial clonrance of 3" or less (referred to as "rattle points") where the ,

new displacements exceeded both the clearance '

requirement and the original displacement computed in  !

the initial Hostinghouse pipe analysis.

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2. Areas of potentially high stress including elbows, branch connections, equipment nozzles, and postulated break locations. This included enough adjacent piping to account for all directions of movement.

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3. Locations at which newly computed displacements exceeded the original displacements by more than one inch. This '

was added to account for localized increasen in ,

displacements due to snubber removal, even though the i overall subsystem motion was not significantly changed. Procedure COM-PI-BYR19 (mentioned in the .

violation) simply defined the methods for determining  !

these locations.

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4. Locations at which new total computed displacement i I

' (thermal + seismic) exceeded the limits of the original l design criteria (3 inches). The three inch criteria was arbitrarily selected early in the plant design process (

based upon preliminary assessments of congestion in the '

auxiliary building and containments. Three inch  :

separation was a goal for both the original design and I installation. In the limited cases where 3 inch separation could not be achieved during snubber  !

i reduction design, the PEclis identified these segmento l l for preinstallation inspection which led to field change  !

l requests and subsequent walkdowns and resolutions. The l piping installers also identified all other points where j the three inch separation could not be achieved so those

• j areas could also be analyzed.

Per procedure Coll-PI-BYR21, all potential seismic interactions i identified during these walkdowns were resolved either by i analysis or modification.

l l Cort;ective Action To Be Taken To Avoid Furtiler VLqlM,_lon The governing 11UTECl! procedures have been revised and training held to include the scope of walkdowns described above. All l

future snubber reduction activities at the Byron Station will utilize this walkdown criteria.

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Date When Full Compliance Will Be Achieved f Corrective Action associated with this item is complete and full r compliance has been achieved.

VIOLATIO!! lb l 10CFR50, Appendix D, Criterion V, as implemented by Ceco Topica] l Report CC-1-A, "Quality Assurance Program for !!uclear Generating i Stations," and CECO Corporate Quality Assurance Manual, lluclear (

, Generating Stations, "Quality Requirements," requires that '

l activities affecting quality shall be prescribed by documented instructions of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions.

Instructions shall include appropriate quantitative or  !

qualitative acceptance criteria for determining that activities affecting quality have been satisfactorily accomplished.

l Contrary to the above, during the course of reviewing safety f related design efforts performed by llUTECll Engineers, the 11RC f

inspectors observed that activities affecting quality were not ,

prescribed by documented instructions in that l

Instructions did not include seismic anchor movement (S AM) ,

effects during seismic walkdowns.  !

Corrective Action Taken and Results Acitieved i

llOTECil performed a detailed review of SAM displacements for all  !

subsystems within the scope of the snubber reduction project. [

This review showed that all piping except those connected to l reactor coolant pumps and decoupled lines had a SAM displacement l of less than 1/8 inch. For the few instances where the SAM  ;

displacements exceeded 1/8 inch, the potential impacts were l satisfactorily resolved without modifications. Specifically, '

five (5) additional rattle points were identified and added to l the previously identified 113 rattle points. All were roeolved ,

through analysis.  ;

Corrective Action Taken To Avoid Further Violation l The governing flUTECl! procedures have been revised end training held to ensuro SAM displacements are considered during seismic '

walkdowns.

Opte Nhen Full Compliartge Will Be Ag[Lieved Corrective Action associated with this item is complete and full compliance has been achieved.

VIOLATION 2q  !

10CFR50, Appendix B, Criterion III, as implemented by Ceco l Topical Report CE-1-A, "Quality Assurance Program for Nuclear i Generating Stations," and CECO Corporate Quality Assurance Manual, Nuclear Generating Stations, "Quality Requirements,"

requires that measures shall be established to assure that '

J regulatory requirements and the design basis for those structures l

and systems are correctively translated into specifications, drawings and instructions. These measures shall includo i provisions to assure that appropriate quality standntda are >

l specified and included in design documents and deviations from

.' such standards are controlled.

Contrary to the above, during the course of reviewing safety-l related design activities performed by HUTECil Engineers, the NRC '

inspectors observed that the above criterion was not met in the following areast Zero period acceleration effects were not considered in the 1

evaluation of safety-related piping systems. As a result,  ;

s 106 supports and nine valves required requalification and one support required modification.

Corrective Action Taken and Results Achieved i l NUTECit performed a study of all 67 subsystems analy=ed during the  :

J snubber reduction project. This study involved a uniform static i i acceleration analysis in each of the three spatial directions. l

) The acceleration amplitudes correspond to the response spectra l acceleration values at the modal cut-off frequency (33 !! or l l

greater). Responses from the three static acceleration cases I were then combined to obtain a resultant response due to the ZPA 1

analysis by the square-root-sum-of-the-square (S RSS) technique.

For valves, accelerations obtained from the response spectra l analysis were combined with the ZPA values by SRSS. In all cases i

! (approximately 210 valves), the combined accelerations were [

1 determined to be within acceptable limits. For supports, the

] reactions obtained from the ZPA analysis were compared to those l l from the response spectra analysis. The maximum reaction of these two cases were checked against the design load for the I i

supports. Of the 1700 supports on these subsystems, only 113 l 1 supports had to be requalified and only 1 of the supports, a l strut, required replacement with a larger si:o. Although this  ;

j strut change was not needed to meet the design basia j requirements, the change was made to provide extra margin.

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J E Corrective Action To Be Taken To Avoid Further Violations l

The OPTPIPE computer program has been modified so that ZPA effects will be calculated by the program and included in the -

response spectrum results for future analyses. Applicable ,

1 procedures have been revised accordingly. Also, an internal memo l has been issued to all NUTECH personnel involved in performing l piping analysis informing them of this issue and defining i corrective action taken to resolve it.

il Date When Full Comoliance Will Be Achieved I

j Corrective action associated with this item is complete and full I compliance has been achieved. ,

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] yIOLATION 2c I

10CFR50, Appendix B, Criterion III, as implemented by CECO Topical Report CE-1-A, "Quality Assurance Progrrm for Nuclear

., Generating Stations," and CECO Corporate Quality Assurance

Manual, Nuclear Generating Stations, "Quality Requirements,"

requires that measures shall be established to assure that j regulatory requirements and the design basis for those structures and systems ..re correctly translated into specifications, drawings and .nstructions. These measures shall include i provision to assure that appropriate quality standards are j specified and included in design documents and deviations from j such standards are controlled. '

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i Contrary to the above, during the course of reviewing safety-related design activities performed by NUTECH Engineers, the NRC ,

inspectors observed that the above criterion was not met in the following areas:

Other than engineering judgements, there were not technical  ;

bases for the 90% or insulation criteria used in the rattle ,

point walkdowns.

Corrective Action Taken and Results Achieved l ,

2 The 90% stress critoria and insulation criteria were used as a screening process during only the initial analysis and design phase of the project. These criteria did not substitute for any 1 walkdowns. These criteria allowed NUTECH to identify the rattle i points which would not likelf exceed design criteria after l 1

anubbers were removed from the lines containing those rattle l points. In all cases, after the snubbers were removed, each l rattle point was inspected for potential interference.  !

The screening criteria allowed for resolution of rattle points as

) long as the pipe stresses were less than 90%. That is, though

] the pipe deflection may have exceeded,the rattia point dimension, i l the fact that the pipe stresses were less than 90% of code i

) allowables enabled a PECH to be issued. The possibility of a 1

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major modification during the walkdown was considered improbable. The walkdown criteria defined boundaries of seismic interaction walkdowns to include potentially high stress locations, to include all piping where stresses exceed 90%.

However, final resolution of all potential impacts within 3 inches of the piping (e.g., rattle point) was accomplished through rigorous analysis or modification. There were statomonts in procedure COM-PI-BYR21 which could be viewed as using the 90%

criteria as a means of resolving impac.ts after the walkdowns.

However, these statements were clarified during the inspection.

Corrective Action To Be Taken To Avoid Further Violation All future anubber reduction activities at the Byron Station will utilize seismic interaction effec,ts evaluation criteria similar to that currently documented in the project instructions for this project.

Date When Full Compliance Will Be Achieved Corrective action associated with this item is complete and full compliance has been achieved.

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