ML20128B855
| ML20128B855 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 05/10/1985 |
| From: | Farrar D COMMONWEALTH EDISON CO. |
| To: | James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| Shared Package | |
| ML20128B780 | List: |
| References | |
| 0065K, 65K, NUDOCS 8507030366 | |
| Download: ML20128B855 (19) | |
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Commonwealth Edison One First National PI 2a. Chicago. Ilknois Address Reply to: Post Office Box 767 Chicago, Illinois 60690 May 10, 1985 Mr. James G. Keppler Regional Administrator U.S. Nuclear Regulatory Commission Region III 799 Roosevelt Road Glen Ellyn, IL 60137
Subject:
Byron Station Units 1 and 2 Braidwood Station Units 1 and 2 IE Inspection Report Nos.
50-454/84-51 and 50-455/84-35 NRC Docket Nos. 50-454; 50-455; 50-456; 50-457 Reference (a):
April 3, 1985 letter from R.L. Spessard to Cordell Reed
Dear Mr. Keppler:
Reference (a) provided the results of inspections by Mr.
I.T. Yin from August 13, 1984 through April 2, 1985.
During these insoections certain activities were found to be not in compliance with NRC req,uirements.
Attachment A to this letter contains Commonwealth Edison's response to the Notice of Violation appended to reference (a).
On April 10, 1985, Commonwealth Edison was informed that a response to requirements 1 and 2 referenced from 10 CFR 2.201 is not necessary for each violation, as these requirements were inadvertently included in the Notice of Violation.
On May 2, 1985, Commonwealth Edison was granted a seven day extension on the due date for the response to the Notice of Violation.
The response in Attachment A also includes a description of our actions to assure the same concerns do not develop at Braidwood Station.
Attachment B to this letter contains Commonwealth Edison's response to the additional concerns stated in reference (a).
Please direct any questions regarding this matter to this office.
Very.truly yours, d-8507030366 850618 PDR ADOCK 05000454 4
G PDR D. L. Farrar Director of Nuclear Licensing cc:
Byron Resident Inspector Braidwood Resident Inspector St385 l
Attachments 0065K i
ATTACHMENT A VIOLATION 1 10 CFR 50, Appendix B, Criterion V, as implemented by CECO Topical Report CE-1-A, " Quality Assurance Program for Nuclear Generating Stations", and CECO Corporate Quality Assurance Manual, Nuclear Generating Stations, Quality Requirements, requires the licensee to accomplish safety-related activities in accordance with documented procedures.
NPS Procedure 3.0.9, Revision D, Paragraph 5.3 stated that "As-Built dimensions will be obtained by actual field measurement for all dimensions shown on the NPS-H-1000 series drawings."
Contrary to the above, NPS personnel only obtained by actual field measurements those dimensions which were checked to be questionable against the design and inspection documents.
Some cognizant licensee personnel misunderstood the provisions of the procedure.
RESPONSE
Nuclear Power Services (NPS) Procedure 3.0.9 Revision D, Paragraph 5.3 stated that "As-Built dimensions will be obtained by actual field measurement for all dimensions shown on the NPS-H-1000 series drawings."
The actual field measurements for as-built piping and support chain dimensions were taken by one of the two following methods:
1.
As-built measurements were obtained by Hunter Q.C.
personnel as part of the initial installation inspection and recorded on isometric drawings as " boxed in" (as-built) dimensions.
2.
As-built measurements were obtained by a team of NPS and Hunter Q.C. personnel as part of the preparation of as-built piping subsystem packages and recorded on isometric drawings as " boxed in" (as-built) dimensions.
The combination of these two methods provided an as-built subsystem package with dimensions that had been obtained by actual field measurement.
CORRECTIVE ACTION TAKEN AND THE RESULTS ACHIEVED We believe the intended provisions of the NPS Procedure 3.0.9 (Paragraphs 5.2 and 5.3 collectively) were met.
However, in order to clarify the intent of Paragraph 5.3 to be consistent with existing field practices, Procedure 3.0.9 was revised on December 13, 1984 (Revision E) to indicate that "as-built dimensions not previously obtained by Hunter Q.C. will be obtained by actual field measurement" as part of the preparation of as-built subsystem packages.
As-built information previously taken by Hunter Q.C.
personnel is not required to be remeasured.
CORRECTIVE ACTION TAKEN TO AVOID FURTHER NONCOMPLIANCE The Byron Unit 2 piping and support as-built dimensions are currently being obtained in accordance with NPS Procedure 3.0.9, Revision E.
At Braidwood Station, as-built verification will not be performed by NPS.
As-built verification will be performed utilizing approved contractor procedures.
DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED Full compliance was achieved on December 13, 1984 when Revision E to NPS Procedure 3.0.9 was issued.
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, VIOLATION 2a 10 CFR 50, Appendix B, Criterion III, as implemented by CECO Topical Report CE-1-A, " Quality Assurance Program for Nuclear Generating Stations", and CECO Corporate Quality Assurance Manual, Nuclear Generating Stations, Quality Requirements, requires measures be established to assure that applicable design bases for systems and components are correctly translated into specifications, drawings, procedures, and instructions.
Contrary to the above, the bases for acceptance criteria for piping interaction (clearance) had not been formally established prior to implementation of the inspection procedures.
RESPONSE
As indicated in the Inspection Report attached to reference (a), "S&L engineers showed tne inspector previous safety-related piping system thermal and seismic movement studies."
These studies were performed in the development of the procedures and criteria to be used for clearance walkdowns and disposition.
This information was judged by Sargent & Lundy to be sufficient to demonstrate the rationale for using 3" clearance as an evaluation point.
The 3" evaluation clearance identified in 1978 was judged adequate based on typical 1/2" seismic movements with a 2" margin.
Actual analysis of piping movements were evaluated in January, 1983 to verify the Sargent & Lundy judgement that 3" was a reasonable guideline for clearance identification.
CORRECTIVE ACTION TAKEN AND THE RESULTS ACHIEVED To address specific detailed questions raised by the inspector, it was decided to generate a formal report that could be utilized as a documented rationale for the 3" clearance threshold for evaluation purposes.
As indicated in the Inspection Report attached to reference (a), a formal report EMD-050661 " Statistical Study for Seismic Interaction" was generated.
CORRECTIVE ACTION WHICH WILL BE TAKEN TO AVOID FURTHER NONCOMPLIANCE The bases for the 3" clearance included engineering judgements drawn from previous studies.
The inspection determined that these judgements were not adequately documented.
A procedure is now in place at Sargent & Lundy that addresses the formal documentation requirements for certain engineering judgement.
Noted as an unresolved item, the Inspection Report attached to reference (a) indicates "The report did not consider situations that could invalidate inspection acceptance when hard calcium silicate pipe insulation was installed after piping clearance inspection and measurement."
The piping clearance procedures for the remaining Byron and Braidwood units will be modified to make provisions for considering instances where hard calcium silicate pipe insulation is used.
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_4 DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED The formal Sargent & Lundy report EMD-050661 was issued on November 16, 1984.
The revision to the Byron and Braidwood piping clearance procedures to address instances where hard calcium silicate pipe insulation is used will be completed by July 1, 1985.
PROPOSED REVISION TO IE BULLETIN 79-14 IMPLEMENTATION l
The inspector provided several suggestions that would J
improve and simplify the items that must be reviewed to address I&E Bulletin 79-14.
It is our intent to review our piping clearance procedures and revise them where appropriate to limit the clearance reviews to check for seismic interferences only on piping systems that operate below 2500F.
The areas to be examined on these piping systems would only be the high stress locations (e.g. branch lines, equipment nozzles, break locations).
Piping systems that
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operate at temperatures creater than 2500F will be examined for proper clearances during the Thermal Expansion Preoperational Test.
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. VIOLATION 2b 10 CFR 50, Appendix B, Criterion III, as implemented by CECO-Topical Report CE-1-A, " Quality Assurance Program for Nuclear Generating Stations", and CECO Corporate Quality Assurance Manual, Nuclear Generating Stations, Quality Requirements, requires measures be established to assure that applicable design bases for systems and components are correctly translated into specifications, drawings, procedures, and instructions.
Contrary to the above, possible structural deformation due to rubbing within an energy absorbing material (EAM) enclosure had not been considered in the pipe whip restraint (WR) design.
RESPONSE
The potential for structural deformation due to rubbing interference within energy absorbing material (EAM) enclosures was not considered in the pipe whip restraint design, although the effects of such interference are well accommodated within the design margir and performance of the affected pipe whip restraints.
Rubbing within the affected EAM enclosures was subsequently judged not to inhibit the energy absorbing action.
This judgement was based on the combined effect of the break force (500,000 to 1,000,000 pounds moving at approximately 20 feet per second) and the available restraint derormation mechanism of hinge formation.
We believe that the restraints would have performed as intended without l
modification.
CORRECTIVE ACTION TAKEN AND THE RESULTS ACHIEVED Due to startup schedule considerations, the side plates on the Byron Unit 1 pipe whip restraints MS-Rll, MS-R33 and MS-R48 were cut to remove the potential rubbing interferences.
As stated in the Inspection Report attached to reference (a), the inspector considered the modifications to be acceptable.
CORRECTIVE ACTION TO AVOID FURTHER NONCOMPLIANCE Subsequent to the above action, these three restraints were L
eliminated from the remaining Byron and Braidwood Units as a result
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of recent NRC approval to eliminate arbitrary intermediate pipe i
breaks.
DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED j
Full compliance was achieved on Byron Unit 1 when the modifications were made to the pipe whip restraints on November 21, 1984.
Byron Unit 2 achieved full compliance when the design changes were made to eliminate the three affected restraints on February 8, 1985.
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. VIOLATION 2c 10 CFR 50, Appendix B, Criterion III, as implemented by CECO Topical Report CE-1-A, " Quality Assurance Program for Nuclear Generating Stations", and CECO Corporate Quality Assurance Manual, Nuclear Generating Stations, Quality Requirements, requires measures be established to assure that applicable design bases for systems and components are correctly translated into specifications, drawings, procedures, and instructions.
Contrary to the above, EAM strength reduction due to excessive height to width ratio and due to stacking were not determined during the initial design stages.
RESPONSE
EAM crush strength reduction due to excessive height to width ratio and due to stacking were not evident during the initial design stages.
Therefore, design criteria to explicitly account for such strength reduction were not in place.
The test data initially received from the EAM manufacturer did not indicate strength reduction due to height to width (H/W) ratios of two or less, nor due to stacking.
Additional confirmatory tests were performed in October and November, 1984 which showed an approximate 10% strength reduction for H/W ratio of 2.1 and 15% strength reduction due to stacking.
Based on the available design margin of the pipe whip restraints, we believe further analysis would have shown the acceptability of such reduction in EAM crush strength.
CORRECTIVE ACTION TAKEN AND THE RESULTS ACHIEVED On Byron Unit 1, active EAM has been replaced.
The replacement EAM is not stacked.
The identified strength reduction of the EAM due to H/W ratio is small.
Therefore, the design is considered adequate.
CORRECTIVE ACTION TAKEN TO AVOID FURTHER NONCOMPLIANCE For the other three units, the active EAM will be replaced.
Stacking of the replacement EAM is not anticipated.
However, if stacking becomes necessary, the corresponding strength reduction will be accounted for.
With respect to strength reduction from H/W ratio, the design is considered adequate as stated above.
For future design of pipe whip restraints that use EAM, Sargent & Lundy has revised their design standard SDS-E29 to take into account reduced EAM crush strength due to H/W ratios and stacking.
. DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED Full compliance.has been achieved.
Sargent & Lundy design standard SDS-E29 was revised on January 31, 1985.
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. VIOLATION 2d 10 CFR 50, Appendix B, Criterion III, as implemented by CECO Topical Report CE-1-A, " Quality Assurance Program for Nuclear Generating Stations", and CECO Corporate Quality Assurance Manual, Nuclear Generating Stations, Quality Requirements, requires measures be established to assure that applicable design bases for systems and components are correctly translated into specifications, drawings, procedures and instructions.
Contrary to the above, possible interference between the EAM retaining box that could reduce the EAM crush strength during EAM deformation had not been evaluated for acceptance or tested for validation.
RESPONSE
The actual configuration utilizing EAM supported by a retaining box was not tested to validate whether possible interference could affect EAM crush strength during EAM deformation.
However, in our judgement, the EAM retaining box does not affect the EAM crush strength during EAM deformation because we believe the retainer box has been provided with sufficient clearance.
In addition, the design of the retainer box height meets the 50% strain criteria for EAM indicated in the FSAR.
CORRECTIVE ACTION TAKEN AND THE RESULTS ACHIEVED In lieu of additional EAM validation testing and due to startup schedule considerations, the retainer boxes for the Byron-Unit 1 restraints 1RY-3 and 1RY-5 were cut to provide EAM heights outside the retainer boxes equal to twice the design deformation of the EAM.
CORRECTIVE ACTION TAKEN TO AVOID FURTHER NONCOMPLIANCE In order to maintain design consistency, and to avoid further testing and analysis, the same design change as above was issued for t'!e remaining Byron and Braidwood Units.
As indicated in the Inspection Report attached to reference (a), the remaining pipe whip restraints with retaining boxes were considered acceptable.
DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED Full compliance has been achieved.
As indicated in the Inspection Report attached to reference (a), the modification to the l
Byron Unit 1 pipe whip restraints was completed on January 26, l
1985.
The design change to implement the same modification on the other three units was issued on April 19, 1985.
. VIOLATION 3a 10 CFR 50, Appendix B, Criterion V, as implemented by CECO Topical Report CE-1-A, " Quality Assurance Program for Nuclear Generating Stations", and CECO Corporate Quality Assurance Manual, Nuclear Generating Stations, Quality Requirements, requires safety-related activities be prescribed by documented procedures which include appropriate quantitative or qualitative acceptance criteria.
Contrary to the above, the IE Bulletin 79-14 system walkdown procedure did not require measurements or estimates be made to establish the clearance between pipe and unsealed penetrations.
RESPONSE
As indicated in the Inspection Report attached to reference (a), measures were in place to assure that sufficient penetration gaps existed to allow unrestricted thermal and seismic pipe movements for penetrations that were sealed.
The vast majority of penetrations were sealed and penetration gap considerations were part of the seal installation program.
However, no specific program for documenting that sufficient gaps existed to allow unrestricted thermal and seismic movements for unsealed penetrations was in place.
CORRECTIVE ACTION TAKEN AND THE RESULTS ACHIEVED During this NRC inspection, 67 unsealed sleeves were field measured to determine the gap between the pipe and sleeve.
The
'i measured gaps were evaluated and found acceptable.
Only one optional modification was made to prevent the deformation of metallic insulation in the vicinity of penetration 1RB47e A 1" clearance was provided by trimming the insulation.
CORRECTIVE ACTION TO AVOID FURTHER NONCOMPLIANCE j
For the unsealed penetrations at Byron Unit 2, gap measurements will be performed as part of the pipe and support as-built program as detailed on NPS drawing H-1000.
Discrepancies will be dispositioned as appropriate.
At Braidwood, as-built verification is not being performed by NPS.
However, the applicable contractor procedures will include provisions for performing gap measurements on unsealed penetrations.
DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED Full compliance has been achieved.
On April 26, 1985, a detail was added to NPS drawing H-1000 to require measurement of the as-built pipe-to-sleeve gap on unsealed penetrations.
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. VIOLATION 3b 10 CFR 50, Appendix B, Criterion V, as implemented by CECO Topical Report CE-1-A, " Quality Assurance Program for Nuclear Generating Stations", and CECO Corporate Quality Assurance Manual, Nuclear Generating Stations, Quality Requirements, requires safety-related activities be prescribed by documented procedures which include appropriate quantitative or qualitative acceptance criteria.
Contrary to the above, retesting of steam generator snubbers using Revision 1 of Procedure SP$-8471-7 was conducted prior to approval of Revision 1.
RESPONSE
Paul Munroe Hydraulic (PMH) 1300 KIP steam generator snubbers were selected as possible replacements for the deficient Byron Unit 1 Boeing steam generator snubbers.
The PMH snubbers met the requirements for Byron Unit 1 based on the Westinghouse specification, the initial testing performed by PMH, and the snubber configuration.
However, Commonwealth Edison decided to perform a verification test on a single snubber to confirm that the PMH l
snubbers met the requirements of the Westinghouse specification.
The testing was conducted by ITT-Grinnell and the first cycle of the testing was witnessed by NRC Region III, Sargent & Lundy, and Commonwealth Edison.
Analysis of the test results revealed a problem in the test instrumentation setup which required further test development and retesting.
The retesting followed the same basic procedure used in the first test cycle, except the instrumentation setup was altered.
A test change was made to reflect the alteration of the instrumentation without the procedure being formally revised and approved prior to execution of the retest.
The retest results were consistent with those of the initial PMH testing.
In addition, subsequent testing performed by PMH in California during maintenance of the snubbers supported the results obtained in the.ITT-Grinnell retesting.
CORRECTIVE ACTION TAKEN AND THE RESULTS ACHIEVED Revision 1 to ITT-Grinnell Procedure SPS-8471-7 was reviewed and approved by ITT-Grinnell on September 27, 1984 and by Sargent & Lundy on October 17, 1984.
i-I. CORRECTIVE ACTION TAKEN TO AVOID FURTHER NONCOMPLIANCE Although testing of this nature is now complete, Commonwealth Edison re-emphasized to ITT-Grinnell and Sargent &
Lundy the importance of approving revisions to test procedures prior to execution of a test.
For prototype testing, ITT-Grinnell instituted a procedural control for test experimentation and reconfiguration of test apparatus.
i DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED Full compliance was achieved on October 17, 1984 when Revision 1 to ITT-Grinnell Procedure SPS-8471-7 was reviewed and approved.
The prototype testing procedural control was implemented by ITT-Grinnell on November 7, 1984.
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, VIOLATION 3c 10 CFR 50, Appendix B, Criterion V, as implemented by CECO Topical Report CE-1-A, " Quality Assurance Program for Nuclear Generating Stations", and CECO Corporate Quality Assurance Manual, Nuclear Generating Stations, Quality Requirements, requires safety-related activities be prescribed by dncumented procedures which include appropriate quantitative or qualitative acceptance criteria.
Contrary to the above, EAM was field cut without an approved procedure that included criteria for dimensional acceptance deviations and cautions not to remove the pre-crush.
CORRECTIVE ACTION TAKEN AND THE RESULTS ACHIEVED The cutting of EAM in the field was performed in accordance with approved FCR's and process sheets, but not to approved procedures that included dimensional acceptance criteria and cautions not to remove the pre-crush.
As a result, Hunter Site Work Instruction 10, Revision 0,
" Field Cutting of Energy Absorbing Material for Pipe Whip Restraints" was written to provide standard provisions and precautions for cutting EAM.
Due to unresolved material variability issues associated with the EAM and Byron Unit 1 startup schedule considerations, the active EAM on the Byron Unit 1 pipe whip restraints was subsequently replaced.
CORRECTIVE ACTION TO AVOID FURTHER NONCOMPLIANCE Hunter Site Work Instruction No. 10 is currently being used to control future cutting of EAM in the field at Byron Station.
In addition, as-built dimensions are being taken so that replacement EAM for the active Unit 2 pipe whip restraints can be shop fabricated to the proper size.
This should eliminate the need for field cutting EAM.
At Braidwood Station, field cutting of EAM will be avoided if possible.
If field cutting of EAM becomes necessary, the proper procedures will be followed.
DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED Full compliance has been achieved.
Hunter Site Work Instruction No. 10 was approved on December 28, 1984.
, VIOLATION 4 10 CFR 50, Appendix B, Criterion X, as implemented by CECO Topical Report CE-1-A " Quality Assurance Program for Nuclear Generating Stations", and CECO Corporate Quality Assurance Manual, Nuclear Generating Stations, Quality Requirements, requires a program be established and implemented for inspection of activities affecting quality.
Contrary to the above, piping as-built dimension inspections failed to identify instances where the tolerances established by the Architect Engineer had been exceeded.
Also, the efforts of walkdown personnel who had not been properly qualified prior to February 15, 1983, were not formally evaluated.
CORRECTIVE ACTION TAKEN AND THE RESULTS ACHIEVED During this NRC inspection, several piping dimensions were found to be outside the tolerance for recording of as-built measurements.
In addition, prior to February 15, 1983, as-built measurements were taken by Hunter personnel who were not certified Q.C. inspectors.
In order to assure that the dimensional discrepancies were isolated occurrences and not design significant, and to verify the accuracy of measurements taken prior to February 15, 1983, the following two reviews were performed:
1.
Hunter Q.C. inspectors remeasured the piping and support chain dimensions on 44 isometric drawings.
Only 21 dimensions out of the 1185 reinspected were outside the tolerance for as-built measurements.
These 21 discrepancies were evaluated by the appropriate architect-engineer and none were found to have an impact on piping analysis.
2.
The chain dimensions at 10 high stress locations on subsystems with operating temperatures greater than 2500F were reinspected.
None of the 61 dimensions remeasured were found to be outside the as-built tolerance.
Based on the results of these reviews, it was concluded that the accuracy of the as-built piping and support dimensions was acceptable.
As-built dimensions were determined to be reconcilable with analysis.
CORRECTIVE ACTION TAKEN TO AVOID FURTHER NONCOMPLIANCE At Byron, Diping and support as-built dimensions are currently being taken by certified Q.C. personnel.
These inspections are reviewed by the unit concept and overinspection j
programs in addition to Quality Assurance audits and surveillances.
l At Braidwood Station, personnel performing as-built inspections are certified.
, DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED Full compliance has been achieved.
The two reviews mentioned above were completed in. December, 1984 and were subsequently found acceptable by the NRC as documented in the Inspection Report attached to reference (a).
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ATTACHMENT B This attachment addresses the concerns noted in reference (a) that were separate from the items contained in the Notice of Violation as follows:
1.
Functionability of steam generator snubbers 2.
Strength of whip restraint energy absorbing material 3.
Interferences of whip restraints and snubbers 4.
Assessment of-the need to enhance the rigor of our examination of data received from our approved vendors related to qualification and production testing of their products.
1.
FUNCTIONABILITY OF STEAM GENERATOR SNUBBERS For Byron Unit 1, the functional problems associated with the Boeing steam generator snubbers were rectified by replacing them with Paul Munroe Hydraulic snubbers.
Replacement snubbers of a different design were chosen because the deficient Boeing steam generator snubbers could not be redesigned, modified, and requalified in time to support the Byron Unit 1 fuel-load schedule.
For the remaining three units, modified Boeing steam generator snubbers will be used.
The ITT-Grinnell redesign of the Boeing snubbers was reviewed by the NRC, Sargent & Lundy and Commonwealth Edison, and was subsequently approved.
The materials from the original Boeing snubbers were ultrasonically tested by ITT-Grinnell and acceptable parts are being reused.
In addition, a stress report provided by ITT-Grinnell concerning their redesign has been approved.
ITT-Grinnell has performed qualification testing on the first modified Boeing snubber they assembled.
The results of this testing qualified their redesign for use.
The qualification testing results were reviewed by the NRC, Sargent & Lundy and Commonwealth Edison, and were found acceptable.
As each modified Boeing snubber is assembled, it will be functionally tested by ITT-Grinnell to confirm its proper operation.
These functional test results will be reviewed and approved by Sargent & Lundy and Commonwealth Edison.
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STRENGTH OF WHIP RESTRAINT ENERGY ABSORBING MATERIAL For Byron Unit 1, the results of EAM crush strength testing at Hexcel to address angularity issues and field cut issues were ultimately accepted by the NRC.
This satisfied the Byron Unit 1 low posar license conditions concerning these EAM issues.
Although this confirmatory testing resolved design issues associated with the strength of EAM, new issues arose which questioned the variability of EAM crush strength within each EAM core block.
Since the Hexcel investigation of the material variability problem was not. complete at the time, Commonwealth Edison decided to replace the Byron Unit 1 active EAM due to startup schedule considerations.
As a prudent measure, we also decided to replace the active EAM ca the remaining three units.
The EAM replacement program for the remaining three units will assure design strength requirements of EAM are met.
The program requires Hexcel to cut test specimens from each EAM core block at various horizontal and vertical locations to confirm crush strength uniformity and adherence to specification requirements concerning crush strength.
Further, Commonwealth Edison will conduct source inspection of the testing of the core blocks and fabrication of EAM, and has imposed hold points on EAM shipments to assure specification requirements are met.
Issues associated with EAM strength reduction due to stacking and height to width ratio (Violation 2c) have been addressed in Attachment A to this letter.
3.
INTERFERENCES OF PIPE WHIP RESTRAINTS AND SNUBBERS Throughout the normal piping system design and installation process, interferences or interactions associated with piping and other components may be created.
Because of this, installation inspections and walkdowns are performed to identify interferences that do occur and assure they are properly dispositioned.
Concerning pipe whip restraint interactions, walkdowns on each unit will assure that no safety significant interferences remain within the zone of tip deflection.
The Inspection Report attached to reference (a) noted an example of a support plate at WR FWR-1 next to a 2" steam generator blowdown line and stated a concern that buckling of the support plate could damage or break the 2" pipe.
The Inspection Report further noted that this condition had been previously identified by the Sargent & Lundy system
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walkdown and was evaluated.
This example reinforces our belief that system walkdowns are an effective means of identifying and dispositioning restraint interactions that may occur on the remaining units.
The other whip restraint interference issues concerning EAM enclosures (Violations 2b and 2d) have been addressed in Attachment A to this letter.
Concerning snubber iteterference, the rear brackets and clamps of rigid struts and snubbers will also be inspected as part of the installation inspection for the remaining three units.
The inspection procedures will contain a specific provision to identify potential rear end bracket and pipe clamp interferences.
Any identified interferences will be appropriately dispositioned.
4.
ASSESSMENT GF THE NEED TO ENHANCE THE RIGOR OF OUR EXAMINATION OF DATA RECEIVED FROM OUR APPROVED VENDORS RELATED TO QUALIFICATION AND PRODUCTION TESTING OF THEIR PRODUCTS.
The Commonwealth Edison Quality Assurance Program includes the requirement for review of testing and test data related to qualification and production testing of vendor products.
As a result of the identified concerns regarding the review of test data, the following actions are being taken by Commonwealth Edison:
1.
A review of existing procurement specifications will be performed to identify equipment remaining to be delivered to Byron /Braidwood for which qualification or production testing remains to be performed.
2.
A review of the pending qualification and production testing will be performed to determine which of these tests will require mandatory source witness inspection.
The source inspection of the testing will then be performed and special attention will be given to the type of problems recently experienced with snubbers and EAM.
3.
For the tests which will be subject to a mandatory source witness inspection resulting from (1) and (2) above, and other qualification and production testing deemed appropriate, Sargent & Lundy will perform a specific documented evaluation of the test data.
4.
Normally scheduled Commonwealth Edison audits will periodically cover the above requirements to assure implementation.
, As our Architect-Engineer, Sargent & Lundy has the responsibility for review ar.d acceptance of vendor data.
In order to procedurally address Sargent & Lundy's examination of data received from vendors, Sargent & Lundy General Quality Assurance Procedure GQ-3.08 titled'" Design Calculations" was revised on January 21, 1985 to require the following:
" External sources of design input may include recognized industry or academic publications as well as commercial sources provided the Preparer is satisfied with the reliability of the input, its applicability to the design, and that the input is not'in disagreement with commitments made iri the SAR or Project Design Criteria.
If the design input is to be used generically on a project, then it shall be included in a Project Design Criteria and/or applicable Departmental Standard.
External sources of design input shall be referenced in the calculations."
Due to other concerns, we are aware of the need to document engineering judgement.
A procedure is now in place at Sargent &
Lundy that addresses the formal documentation requirements for certain engineering judgement.
We believe the added emphasis to Commonwealth Edison's vendor source inspection point program, the revision to Sargent &
Lundy's Procedure GQ-3.08, and our heightened awareness to better document engineering judgement and vendor test data evaluation will provide for an acceptable level of review.
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