ML18054A910

Responds to NRC 890628 Ltr Re Violations Noted in Insp Rept 50-255/89-07.Corrective Actions:Design Engineers & QA Personnel Provided W/Training on Structural & Welding Codes & Code Application to Weld Installation & Exam

ML18054A910
Person / Time
Site:	Palisades
Issue date:	08/10/1989
From:	Berry K CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 8908180078
Download: ML18054A910 (50)

Text

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• G11neral 1946 West Parn11ll Road, Jackson, Ml 49201 * (6171 788-1638 ** -----August 10, 1989 Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 DOCKET 50-255 -LICENSE DPR-20 -PALISADES PLANT -RESPONSE TO INSPECTION REPORT 89007 NOTICE OF VIOLATION Kenneth W Berry Director Nuclear Licensing Nuclear Regulatory Commission Inspection Report 255/89007, dated June 28, 1989, identified strengths in inservice testing programs and weaknesses relative to design control. These weaknesses resulted in three violations supported by numerous examples.

None of these examples were safety cant, but collectively they indicated a need for programmatic refinements and additional communication of management's expectations.

The NRC required a written response to be provided within 30 days, however, discussion between respective members of our staffs extended the due date to August 10, 1989. This letter. summarizes the actions to be taken. Details pertaining to the specific items are provided in the Attachments.

Since 1986 significant efforts have been undertaken by Consumers Power Company to provide for effective control of Plant design change activities.

These efforts have resulted from evaluation of performance by Plant Engineering and Corporate Engineering personnel, Quality Assurance personnel, NRC and the Institute of Nuclear Power Operations.

In achieving an effective design control process; procedures governing modification control activities have been revised, a single design authority has been established, changes to the facility are. being effected through a single unified approach and expectations and standards have been communicated to Design Engineering personnel.

Procedural upgrades have focused on translation of design input to the desired output, controlling and implementing the design change in the field and providing close coordination of the design with the needs of the Plant. In the past, the design authority for "minor" modifications has resided at the Plant while offsite engineering organizations retained the design for "major" modifications.

Establishing the Plant as the design authority for all changes to the facility has been effected by Plant sponsorship of all design control procedures, Plant approval for assignment of design individuals and Plant review of all work completed by non-Plant organizations.

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• Nuclear Regulatory Commission Palisades Plant Response to IR 89007 August 10, 1989 semi-annual design seminars and monthly design supervisor meetings which include Engineering, Construction and Testing and Quality Assurance personnel are being conducted to facilitate communication of procedural changes, standards and expectations.

2 Consumers Power Company believes, and as recognized within the Inspection Report, these efforts have resulted in programmatic strengths.such as; good design procedures, improved equipment performance and competent, knowledgeable personnel.

However, Consumers Power Company also recognizes that as industry performance standards are increased, weaknesses in established programs may develop which require additional effort. NRC violation 255/89007-01 presented 19 examples of inadequate design control related to design changes implemented at the Plant. The first seven of these examples were related to the failure to correctly translate design bases into drawings, procedures and instructions.

Five of the examples are acknowledged as presented and are attributed to the failure to; 1) follow established procedures, 2) provide adequate justification and documentation within cation packages or 3) provide for adequate technical reviews of installation efforts. Also, certain areas were identified where procedural enhancements and improved design guidance would preclude recurrences. er, the remaining two examples, 255/89007-0ld and Olg, are not acknowledged as* presented within the Inspection Report. For these two* examples we believe the design intent of the modification was preserved and verified by testing and that record drawings utilized reflect the as-built condition of the Plant. The nine examples were related to the failure to adequately verify and check design. Eight of the examples are attributed to the failure to; 1) follow established procedures, 2) document engineering decisions or 3) provide for adequate technical reviews. Also, certain areas were fied where procedural enhancements would preclude recurrence.

However, Consumers Power Company does not acknowledge the remaining example 255/89007-011.

For this example, the Inspection Report noted that a setpoint change was implemented without assuring the design intent of the system had not been compromised.

In review of the documentation supporting the design change, it was verified that design intent of the system was considered and documented within the modification package and had not been compromised.

The remaining three examples were identified as non-compliances for the to adequately delineate acceptance criteria.

Two of these examples are attributed to a lack of procedural guidance within modification procedures.

Consumers Power Company does not believe example 255/89007-0lq is valid as presented in that appropriate equipment selection criterion were applied during design and documented within the modification package

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• Nuclear Regulatory Commission

• Palisades Plant Response to IR 89007 August 10, 1989 3 In an effort to ensure the accuracy of the existing plant design basis is maintained, discrepancies identified within analyses supporting the cited design changes have been or will be dispositioned and documented.

As an effort to collectively utilize auditing agencies appraisals of our past performances, the identified deficiencies were presented to Design Change Engineers with emphasis placed on strict adherence to established procedures and the concept of Plant based modification engineering.

Enhancements being made to design change procedures regarding documentation of engineering judgement, substantiating input assumptions and* thorough technical reviews will be presented to design change engineers via personal letters, performance seminars and continuing training programs.

Enhanced design guidance is being developed for weld engineering.

Specifically, code training for weld neers is being conducted as well as design change procedure revision to "prompt" the use of existing weld engineering guidelines for proper code selection and specification.

In addition, as part of the Configuration Control Project, additional engineering guidance regarding cable sizing and raceway fill, designing fire barriers and fire stops, evaluating station and emergency power* system.component loads and cable routing including the effects of cable submergence, is being developed.

Additionally, more engineering guidance in the form of an engineering specification will be developed for the civil/structural discipline.

This specification will be developed by July 1990. . NRC violation 255/89007-02 presented two examples where socket fillet welds were not-verified to be in conformance with weld size requirements provided in welding specifications.

These examples are attributed to a failure to meet current expectations for the control of design change implementation.

To . avoid further non-compliance, design change procedures are being revised to present welding specifications input checklists and implementation drawings, and to provide for technical reviews of weld requirement inputs by Maintenance Planners.

Additionally, Design, Engineers and Quality Assurance personnel are.being provided with training on structural and welding codes and their application to weld installation and examination.

NRC violation 255/89007-03 was issued for a failure to implement and maintain Technical Specification low temperature overpressure (LTOP) setpoints which were changed through the specification change process. The violation is attributed to poor within the Technical Specification Change Request development process. When the LTOP setpoints were derived, Plant personnel failed to identify that the value included in the Technical cation did not account for calibration tolerance.

A letter of interpretation has been submitted to the NRR which documents ou"' '1-'osition and commits to revising the setpoints in a forthcoming Technical Specification Change quest. In the interim, surveillance procedures which provide for setting and verifying the LTOP setpoints.have been revised to remove the positive tion tolerance.

An evaluation will be conducted to determine where ments in the Technical Specification Change Request process can be made to preclude recurrence.

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• Nuclear Regulatory Commission Pal:isades Plant Response to IR 89007 August 10, 1989 4 The Inspection Report additionally requested a written response be provided for certain, specific examples of programmatic weaknesses.

The first weakness cited involved the addition of zener diodes in the safety injection tank pressure transmitter power supply without analyzing potential failure modes and without checking diode input voltage after installation.

The failure to fully analyze potential failure modes is attributed to personnel error. Administrative Procedures currently require that .a failure modes and effects analysis (FMEAs) be performed as part of the safety evaluation process. The periodic*

refresher training program for design engineers will include emphasis on FMEAs. The next weakness cited pertained to the backup nitrogen supply modification.

Specifically, an unauthorized design change was implemented when field nel implemented their own weld requirements after identifying that an priate weld was specified by the design engineer.

The condition is attributable to the fact that welding maintenance procedures are not. ly integrated with design control procedures, thus assuring that changes. in the field will be approved by engineering before they are undertaken.

The welding maintenance procedures will be better integrated with the design control procedures.

The third weakness pertained to utilization of different editions of the ASME Code relative to stress intensification factors utilized in analyses.

In summary, usage of the later addition of the ASME Code, as currently described in the Palisades.

Final Safety Analysis Report (FSAR), was discussed in an April 1980 meeting between Consumers Power Company and the NRC and found to be acceptable.

Our interpretation of the results of this meeting was submitted to the NRC in the draft form, revised FSAR pages in our Final Response to IE Bulletin 79-14 dated September 26, 1980. As indicated in our submittal to the NRC dated October 24, 1980, the use of different code editions was found to be acceptable, reviewed in accordance with 10CFR50.59 and placed in the Palisades FSAR. Therefore, usage of different code editions as presented in the FSAR currently represents our position and is believed to be acceptable.

The last weakness cited pertains specifically to the Engineering Design Change (EDC) form utilized to revise facility changes not listing calculations which may be affected by the particular EDC. Therefore, it was unclear whether technical reviewers had considered the effects of the EDC on the original analyses.

Consumers Power Company believes that existing procedural ments direct the EDC initiator to "reflect" the change in all affected tailed design documents; the engineering analysis was clearly identified in the procedure as being a detailed design document.

However, "engineering analyses" will be specifically added to the EDC form to ensure that technical reviewers consider effects on engineering analyses and provide documentation of this consideration

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• Nuclear Regulatory Commission

• • **Palisades Plant Response to IR 89007 August 10,

• 1989 5 The Inspection Report also requested that specific discussion be provided regarding unresolved items pertaining to welding. This discussion is ed on page 41 of Attachment

1. In summary, we acknowledge that no corrective actions have yet been directed towards reviewing previously made socket fillet welds for compliance with code requirements.

Consumers Power Company plans, however, to select an appropriate sample.of as-built welds and inspect the

• welds during the 1989 maintenance outage. The sample will be chosen to include a range of weld types. The purpose of the inspection will.be to verify that the weld characteristics (type and size) conform to requirements set forth in the repair inspection checklist and/or applicable welding code. Kenneth W Berry Director, . Nuclear Licensing-CC Administrator, Region III, USNRC NRC Resident Inspector

-Palisades Attachments OC0889-0167-NL04

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• * * .*_.* *; ATT0889-0167-NL04 ATTACHMENT 1 Consumers Power Company Palisades Plant Docket 50-255 DETAILED RESPONSES TO INSPECTION REPORT 89007 August 10, 1989 45 Pages . ' . * ,. . ..* *. , .. *** .. * *. ** ** .. *---"""--'---'----'-'-'""'"--=...;.......*

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• Violation (255/89007-0!A-S)

1. lOCFRSO, Appendix B, Criterion III, as implemented by the Palisades Operations Quality Assurance Program requires, in part, that the design bases be correctly translated into drawings, procedures, and instructions; that the design control measures provide for verifying or checking.the adequacy of the design; and that design control measures be applied to the delineation of acceptance criteria for inspections and tests. Contrary to the above, the following instances of inadequate design control were identified:

This is a Severity Level IV Violation.

This violation is sustained by 19 examples.

Though Consumers Power Company believes four of these are not supportive examples.

We do acknowledge the violation.

Our detailed response to each example follows: MI0789-1683A-TC01-NL02 1 ... . -::* -* .. ........ * -....... ,. .. *. .... *... ,.,_

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NRG Violation 255/89007-0la:

EA-FC-789-07, "Seismic Analysis of Auxiliary Feedw'ater Control ESSR 88714, 11 Revision l, August 24, 1988. [Refer to page 9 of NRG Report 50-255/89007 (DRS).] Example FC-789 contained multiple dimensional differences between the analysis model and the installation drawings.

The following examples are provided:

-The location of new support 8224 was analyzed at 6 11 from the 45° elbow. The piping drawing (M-101 Sheet 5113) *used to install the support specified a dimension of l'-7 1/2" from the elbow. This difference was not noted in the calculation.

-The length of pipe between Model Nodes 6276 and 6282 was analyzed as 5'-10" long. The installation drawing specifies S'-6" long. This difference was not noted in-the calculation.

Several additional -dimensional .discrepancies on the. new. bypass piping were . also noted between the analysis and installation drawing. These discrepancies ranged from 1 11 to 2-1/4" and were considered minor by the inspector.

none of these discrepancies were noted in the calculation.

Reason for Violation During the evaluation*

of the design of the bypass piping system numerous changes in design dimensions were encountered due--to pipe, support and valve operator-interferences.

At a certain point in the analysis process, it was decided to build* the design *to. the drawing and* effect the final analysis reconciliation*when the as-built data were recorded on a marked-up drawing. The analysis reconciliation with the as-built was never made. This violation was due' to inadequate documentation of the* justification for analytical input and failure to follow established procedures.

Corrective Action Taken, and** Results Achieved-All engineering groups have been briefed as to the results of this inspection.

These briefings were completed on August 2, 1989. The above noted cies have been satisfactorily dispositioned and the finite element piping analysis model has been updated. Corrective Actions to be Taken to Avoid Further Non Compliance Interim All design change engineers will be briefed as to the reported violations by personal letter. These letters will require that all engineers involved in design changes scheduled for installation in 1989 review existing design ages for similar problems and correct any identified problems.

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.-:*** , .. ,,. .. Long-Term Enhancements will be made to plant administrative design control procedures to further clarify the requirements that strict alignment between engineering analyses, associated/accompanying drawings, and as-built condition must be verified and documented prior to declaring modified systems/equipment operable.

In additionf a program will be developed to provide periodic refresher training to all design change engineers on design change-related administrative dures. Date When Full Compliance Will be Achieved The personal briefings by letter will be issued by September 1, 1989. dural enhancements will be completed by January 1 1 1990. The program for periodic training will be in place by March 1, 1990. NRC Violation 255/89007-0lb:

EA-FC-789-07, "Seismic Analysis of Auxiliary Feedwater Control ESSR 88714" Example b.l -For the south bypass loop, the Young's Modulus was specified as 27.4 E6 psi instead of 27.9 E6 psi. This is equivalent to analyzing this portion of pipe with properties at 300° instead of 70°. This discrepancy was not noted in the analysis.

Reason for Violation The use of the.27.4 E6 psi value for the Young's Modulus represents a 1.8 cent error with *regard* to the correct value of 27 .9 E6* psi value. The impact of such an error is expected to be an underprediction of thermal expansion stress of no more than 1.8 percent. This resulted from inadequate tion of technical review and failure to follow existing procedures.

Corrective Action Taken. and. Results= Achieved*

All engineering groups have been* briefed as to the results of the inspection.

These briefings were completed on August 2, 1989. Corrective Actions to be Taken to Avoid Further Non Compliance Interim All design change engineers will be briefed as to the reported violations by personal letter. These letters will require that all engineers involved in design changes scheduled for installation in 1989 review existing design ages for similar problems and correct the problems.

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• ** Long-Term to plant administrative design control procedures will be made tog -Provide the technical reviewer a review checklist with a "prompt" to justify the numerical values of all constants and variables utilized as inputs to the analysis (the checklist will provide a comprehensive set of "prompts" to ensure an overall accurate, thorough and auditable analysis). -A mechanism for the reviewer to note minor errors which would not necessitate a reanalysis.

In addition, a program will be developed to provide periodic refresher training to all design change engineers on design change-related administrative dures. Date When Full Compliance Will be Achieved The personal briefings by letter will be issued by September 1, 1989e The procedural enhancements and training on the enhancements will be completed by January 1, 1990. The program for periodic training will be in place by March 1990. Example b.2 The location of the center of gravity (CG) for the new bypass valves was analyzed at 19 11 from the pipe centerline.

The location specified on the vendor-drawing was 22 11* This represents a 15% increase in the moment arm which was not noted in the calculationo Reason for Violation The piping analysis was set up from preliminary data. The valve assembly weight was included in the model. However, the weight placement was not sistent with-*the final drawing received.from the vendor. The existing mentation does not indicate whether or not the analyst reviewed the center of gravity data from the vendor drawing. The analysis certainly was not run to accommodate This violation occurred due to failure to account for vendor information as analytical input and failure to follow established procedures.

Corrective Action Taken and Results Achieved All engineering groups have been briefed as to the results of the inspection.

These briefings were completed on.August 2, 1989. The calculation was revised to incorporate the correct vendor data and was found to be acceptable.

Corrective Actions to be Taken to Avoid Further Non Compliance Interim Same as that required for Violation l.a. MI0789-1683A-TC01-NL02 4 ...... ... \*::..:***** . . . . *,_ . . .*. . . . .:::*-

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• Long-Term Enhancements to plant procedures will be made Ensure that vendor information/recolillllendations are accounted for ical input and that justification be provided for departure from information/recommendations, as such Provide the technical reviewer a review checklist with a 11 prompt" to assure that vendor information/recommendations are appropriately accounted for. A program will be developed to provide periodic refresher training to all design engineers on design change-related plant administrative procedures.

A 11 punch 11 list or equivalent will be developed to track items requiring verification when data becomes available.

Date When Full Compliance Will be Achieved The personal briefings by letter will be issued by September 1, 1989. These procedural enhancements will be in-place by January 1, 1990 as will required training on these enhancements.

The program to provide refresher training will be in place by March 1, 1990. Example b.3 In addition to the above noted discrepancies for modeling the bypass piping, other dis.crepancies were noted in the model of the original auxiliary feedwater piping. The inspector could not determine whether these discrepancies were inherent in the original data or whether they occurred during the transcription of the original model into the current piping analysis.

However, notes in the piping model stated the following: "Bechtel analysis is a bit off from ISO here." -"Bechtel has modeled elbows only with SIFs. Elbows are used here." -"Review ISO for pipe schedule change." These notes led the inspector to question the validity of the assumption made in the calculation concerning the correctness of the original input data. CPCo Response The three notes recorded by the inspector do not necessarily imply errors in the original input analysis.

The notes reflect free text written into the ADLPIPE computer model by the translator of the ME101 Bechtel model for the review by the piping analyst. The specific analysis model/ISO discrepancy was small. However, the note advised the analyst that a choice needed to be made for analysis record runs. MI0789-1683A-TC01-NL02 5 . :.'.-.** -

• There is nothing wrong with modeling elbows with SIFs and flexibility characteristics.

However, the note merely advises the analyst that comparing ADLPIPE elbows and ME101 elbows for counting of elbows for model benchmarking will not yield consistent results and that the MElOl model will require more review to ensure model consistency.

The note with respect to pipe schedule change is again for the benefit of the analyst. No error is implied. No corrective action is required.

Example b.4 The additional discrepancies in the mod*el of the auxiliary feedwater piping were as follows: -For flow element FE-0736, the weight of 192 lbs was modeled at node 211 instead of node 205 *. Although this was only a 4-1/2" error on a 6 11 pipe, the flange pair was analytically modeled with the weight concentrated at one edge instead of at the middle of the flanges. For Valve M0-0754, the 460 lb weight was modeled at the centerline of the pipe at node 267. The weight should have been specified at the valve CG at node 268, 18" out from the pipe centerline.

The horizontal response spectra used in the analysis was inconsistent with the spectra given in Specification C-175. The spectra used was lower and not as broad as those given in the Specification.

-Piping .between the nodes 252 and 253 was modeled as 4", schedule 40, instead of 6 11 , schedule 80. The above discrepancies are further examples of violation of 10 CFR 50, Appendix B, Criterion III in that the licensee failed to correctly translate the design into the drawing (255/89007-0lb).

Reason for Violation The placement of the flow element weight, the placement of the valve operator weight and the pipe schedule discrepancy constitute discrepancies which should be picked up in the review process. The reason for the violation has been attributed to an inadequate technical review and failure to follow established procedures.

The horizontal response spectra employed in the original IE Bulletin 79-14 analysis of the Palisades piping systems were based upon the Taft 1952 record. The digit,ized data and a straight-edged set of plots from those data were

.. to Consumers Power Company by Bechtel in 1976. The horizontal response spectra used in the piping analysis were derived from these digitized data. The straight-edged plots were used for building and equipment tion seismic work

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Because the straight-edged plots were very difficult to read and because it was desired to incorporate building and equipment spectra in a single seismic ification specification, the straight-edged plots were redrawn and incorporated into Specification C-175. It is expected that the horizontal spectra of C-175 could be slightly higher and broader than the straight-edged spectra. However, that was not the purpose for drawing them. Although the C-175 horizontal tra should be very similar to the straight-edged horizontal they should be used for building analysis and equipment qualification only. They should not be used for piping analysis.

The correct horizontal response spectra for safety related piping systems at Palisades which use the initial plant seismic design basis are those included in the stress packages as developed from the digitized spectra._

New piping systems or modifications involving substantial changes to existing systems will employ the spectra and procedures in Specification M-195. Corrective Action Taken and Results Achieved All engineering groups have been briefed as to the results of the inspection.

These briefings were completed on August 2, 1989. Corrective Actions to be Taken to Avoid Further Non Compliance Interim Same as for Violation Item l.a. Long-Term Enhancements to plant procedures will be made to: -Provid*e the technical reviewer a checklist with a comprehensive set of "prompts" to ensure an overall accurate, thorough and auditable analysis.

These "prompts" will specifically require that the reviewer check the validity of all analytical input and assumptions.

-Provide. the basis for the selection of design as governing, and -Provide a technical review checklist with.a prompt to concur that governing design criteria (input) have been justifiably selected.

-Identify applications in which C-175 or M-195 would be used. Furthermore, a program witl be developed to provide periodic refresher training to engineering personnel on design change related plant administrative dures. Date When Full Compliance Will be Achieved The personal briefings by letter will be issued by September 1, 1989. dural enhancements will be made by January 1, 1990 as will all required training on the enhancements.

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• NRC Violation 255/89007.0lc:

Consumers Power Company Drawing M-101 Sheet 5113, Revision O, "Piping Isometric, Auxiliary Feedwater Control Valve CV-0736A and CV-0737A Bypass Piping." [Refer to page 12 of NRC Report 50-255/89007 (DRS)o] Example -The size of the fillet weld was determined by the requirements of Welding Specification WPS-11.21, Revision 2; however, for the socket welded fittings, the size of the fillet weld was not specified on this drawingo In reviewing the Repair Inspection Checklist (RIC) for the welds in question, the weld size specified is 1 1/2 11* This is misleading in that this is the size of the pipe and not the size of the fillet weld. In order for the welder to determine the size of the fillet weld, the pipe wall thickness must be obtained and a calculation of 1.09 times the wall thickness must be per-. formed. Although this is a relatively simple calculation, it is a design function and* as such must be controlled.

There is no documentation to demonstrate that this design activity was performed.

In addition, there are *no controls in place to check and verify this design activity.

Reason for Violation Specifying welding requirements (such as applicable code, weld material, weld type and weld size) is an engineering function.

If properly administered by procedure, the maintenance planner can (and has) effectively prescribe welding details* for the field provided that adequate input from engineering exists as a basis. In the past, engineering input has been limited to welding tion and/-0r structural analysis engineering sketches.

which have lacked size dimensions for the welds *. As a result, the planner has failed to provide the proper size on the Repair Inspection Checklist (RIC) thereby requiring the field welder to determine and install the proper weld size. This practice fails to meet current expectations for control of design change implementation.

The plant administrative design control procedures required and currently require that the design change project engineer determine code requirements for assigned projects (Reference 4), and plant maintenance procedures required and currently require that the maintenance planner specify applicable code and weld parameters after consultation with the Engineering Department (Reference 3). These procedures have not been effectively integrated to support one another to ensure that weld specifications from engineering were accurately translated into installation planning, installation, and post-installation The following actions have been/will be taken to ensure the administrative dures relating to weld specifications are properly integrated with the Maintenance Department.

Prior to actions taken as a of recent self-identified failures to verify weld size (Reference 7), no specific requirements existed to verify characteristics (weld, type, size contour) of installed welds. Although Nuclear Operations Department Standards suggest inspection hold points for weld installation verification, working level administrative procedures did not specify a hold point requirement except for fit up. MI0789-1683A-TC01-NL02 8 .. '. **. _.\**.**.*.:

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• * ---------Corrective Action Taken and Results Achieved -All engineering groups have been briefed as to the results of this inspectiono The briefings were completed on August 2, 1989. The Inservice Inspection (ISI) Section of the Plant Projects Engineering Department has effected the role of Design Authority for weld engineering by revising the RIC to identify critical weld parameters and require ISI cal review of the maintenance planner's specifications.

The purpose of the review is to ensure that appropriate welding codes are complied with in the areas of weld installation and post-installation examination.

Revision to the RIC was completed as part of the revision to the plant administrative procedure for control of special processes (Reference 3). -The ISI Section (as well as planners, welders and welding supervisors) has received specific training with respect to welding codes and technology to augment their existing collective knowledge.

-In addition, the RIC .. was revised to issue. the. weld minimum leg length to the field. This will eliminate the need for the field welder to calculate the length. The aforemenqoned ISI review will assure that this specification is provided.

-Finally, the RIC has been revised to require verification of weld size * (RIC now requires that weld is inspected for size, porosity, undercut,-etc.)

Training materials for the welder tra1n1ng progression course have been revised-to emphasize fillet weld terminology and conformance of the completed weld to the.design specification.

Corrective Actions to be Taken to Avoid Further Non Compliance Interim -* Same as that required for Violation Item l.a. Long-Term

-Enhancements to plant design control and maintenance procedures will be made to more effectively integrate engineering into weld specification and mately into weld planning and verification:

Appropriate welding codes will be included in the Design Input Checklist (Reference

2) to "prompt" the design engineer to specify appropriate weld requirements (for installation and examination) in the facility change package as part of both conceptual and detailed engineering.

In addition, a generic guideline will be developed to support the design engineer throughout the weld design process. MI0789-1683A-TC01-NL02 9 . :*.-...

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• Design control procedures related to engineering analyses (Reference

1) will explicitly require that all drawings accompanying structural/seismic analyses provide detailed weld information (type, size, material) for input to the planner. The procedures will also require that sizing calculations be* performed as part of the analysis.

Finally, a technical review checklist will be provided to require that the reviewer ensures that weld information be accurately represented on the analysis drawings.

-Plant maintenance procedures (Reference

3) will require that the maintenance planner utilize the contents of the facility change package to complete the RIC in specifying for the field weld installation and examination ments. The procedure will require that the planner consult the Design Input Checklist and structural/seismic engineering analyses.

Relative to weld verification, the design control program and related welding program will be evaluated and enhancements developed as necessary to ensure that administrative and quality verification controls exist to consistently verify that field installation satisfies design requirements (ie, input vs output). Interim actions related to changes to the RIC and !SI group review of the RIC (as described above) will remain in effect

• Design and quality assurance engineers will be trained on the appropriate structural and piping weld codes and their application to weld installation and examination.

The engineers will also be trained on the above procedural enhancements.

Finally, a program will be developed to periodically train design and quality assurance engineers on the aforementioned codes and their application, and on the weld-related design control and maintenance procedures.

In summary, it is expected that these actions will ensure that proper welding requirements (type, material, size) are specified by engineering, planned by maintenance (with a check on planning by engineering), and in turn, verified by quality control. Date When Full Compliance Will be Achieved The engineering group briefing has been The personal briefings by letter will be issued by September 1, 1989. Procedure enhancements and required training on the enhancements will be completed by January 1, 1990. The program for periodic refresher training will be developed by March 1, 1990. NRC Violation 255/89007-0ld:

EA-T-FC-722-501-01 "Calculation of Acceptance Criteria for Modification Test Procedure T-FC-722-501," January 13, 1987. [Refer to page 16 of NRC Report 50-255/89007(DRS).]

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• Example The calc.ulation on page 2 of the engineering analysis states that the total volume of gas contained in the nitrogen bottles at 2000 psig is 209 scf. This value is incorrect in that it is the usable cylinder volume as given in lation EA-FC-722-02.

The actual volume is approximately 228 scf. By using the incorrect value, the calculated acceptance criteria for pressure drops were higher and, therefore, were nonconservative.

CPCo Response CPCo does not acknowledge this example as a of violation of 10CFR50, Appendix B, Criterion III for the following reasons. 1. As indicated by EA-FC-722-02, the design intent of this modification is to supply a nitr.ogen header pressure from an initial minimum bottle pressure of 2,000 psig down to 150 psig to ensure that the associated control valves would be brought to their safety-related position and maintained in that position for the -required time period.

• 2. In accordance with the design intent of this modification, the usable volume of nitrogen is that volume contained in the bottle from 2,000 psig to 150 psig or 209 scf as calculated by EA-FC-722-02, Sheet 10 of 13. The usable volume of 209 scf is utilized as a conservative value to establish the number of nitrogen bottles required for each station to meet system design requirement.

3. Although not specifically stated in *the body of EA-T-FC-722-501-01, the value of the "usable" volume of nitrogen (209 scf) was utilized in lishing test acceptance criteria rather than the "total" volume of nitrogen (228 scf) to confirm the design intent, verify estimated leakage rates, and confirm system margins. The test procedure clearly tests the design intent of this modification.

Based up_on the above, we feel that this example does not support a violation of lOCFRSO, Appendix B, Criterion III has occurred.

However, certain actions will be undertaken to remedy this minor deficiency and prevent its recurrence:

Interim -All design change engineers will be briefed as to the reported violation by personal letter and by engineering group presentation.

The letter briefings will be completed by September 1, 1989. The group presentations were pleted on August 2, 1989. -EA-T-FC-722-Ji will be revised to clearly indicate that "useable" volume has been utilized to calculate the acceptance criteria rather than "total" volume. MI0789-1683A-TC01-NL02 11 . **-.*****:*:

--:*;*-: ... :**:*,'. ... . . .. :-. ... ;-: . .;. :: ;*.* .. ...... _ ...... . ':. :*;*.-**

Long-Term The actions identified as being taken in the interim are considered complete and effective in responding to this identified condition; no further action is required.

Date When Full Compliance Will be Achieved The engineering analysis will be revised by September 1, 1989. NRC Violation 255/89007-0le:

FC-756 11 HPSI Pump Miniflow Bypass Modification.

19 [Refer to page 18 of NRC Report 50-255/89007 (DRS).] Example Input into the AOLPIPE, Inc (AOL) piping stress analysis, contained in FC-756, contained multiple dimensional differences from the as-built dimensions.

Bechtel's stress.isolmetric drawing 03378, sheet 4 of 5, Revision 1, and drawing Revision 4, showed a dimension of 29 7/8 inches between pump 66A and the elbow. The as-built dimension is 13 1/2 inches. Both (ADLPIPE, Inc.) AOL's and B.echtel's stress analyses used 27 7/B inches. This dimensional discrepancy was not documented during the NRC IEB 79-14 program, nor was it corrected in Bechtel's and AOL's stress analyses.

Further, this discrepancy is in conflict with the assumptions contained in analysis No CS-ESSR 87-144 that purportedly demonstrated that the Bechtel drawings are correct. The inspector also noted that the input data used in the modification portion of the piping system was inconsistent with as-built drawing No 03378, Sheet 4 of 5, Revision 2. The licensee reviewer was not aware of the above dimensional discrepancies.

Failure to correctly translate the design into the drawings is considered an example of violation of 10CFR50, Criterion III. Reason for Violation The dimensional discrepancy associated with the 27 7/8 versus 13-1/2 inch lengths was a result of the analyst relying on data being transmitted from the field and not checking the installation personally.

The smaller discrepancies between the ADL and as-built drawing records were recognized by the analyst when he was provided a marked-up drawing of the as-built configuration.

The analyst acknowledged receipt of the as-builts via memo and stated that the as-built configuration was acceptable and no reanalysis was required.

The reason for the violation was inadequate analytical assumption resulting from a failure to perform a system walkdown and failure to follow established dures. Corrective Action Taken and Results Achieved All engineering groups were briefed on the results of this inspection.

The briefings were completed on August 2, 1989. The dimensional discrepancies noted have been satisfactoril*y dispositioned and documented.

MI0789-1683A-TC01-NL02 12 *::**. -.* ...... .

... :* ..... **', ..

• Corrective Actions to be Taken to Avoid Further Non Compliance The following corrective actions will be taken to prevent Interim Same as that required for Violation Item 1.a. Long Term Procedural enhancements will be made to ensure

-The analyst "walks down" the area of interest *to confirm all as-built (or intended as-built) data is utilized in the analysis.

This confirmation must be made prior to declaring modified structures or equipment operable.

-By approval of the facility change "Responsible Engineer, 11 the above bility for as-built data confirmation may be delegated to field construction by controlled procedure or work order instruction.

-In the event the analyst concludes that no further "analysis" is necessary, the reconciliation of such shall be documented as part of a controlled analysis revision which ensures technical review. A program will be developed

• to provide refresher training on design change related prQcedures.

This training will be directed towards all design change engineers.

_ Finally, a portion of the Configuration Control Projec.t involves the walkdown and field verification of piping as-built dimensions to confirm the accuracy of our stress isometric drawings.

Verification of the stress isometric ings for a sample system is planned for 1990 to assess theneed and extent of further verification activities.

CPCo will perform any required walkdowns by no later than the 1990 refueling outage. Date When Full Compliance Will be Achieved Personal briefings by letter will be issued* by September 1, 1989. Procedural enhancements and required training on the enhancements will be completed by January 1, 1990. The periodic training program will be in place by March 1, 1990. Walkdown and field verification of stress isometric drawings requiring verification will be completed by the 1990 refueling outage. ' NRC Vio*lation 255/89007-0lf:

FC-756 "HPSI Pump Miniflow Bypass Modification." [Refer to page 19 of NRC Report 50-255/89007 (DRS).] Example The as-built sketch used in the analysis for FC-756 contained a nine inch dimensional error. MI0789-1683A-TC01-NL02.

13 *:.:*:: ** . *' ... . i.: .*

_ .. * . ,, .. *-* :-*._:: ... ; . . ' . : :: ... :. ' ........

The as-built sketch for the modification near pump 66A was sent from the site to the engineering office for review. The inspector noted that this sketch contained.a dimensional error. the 2 1-6 1/2" dimension was incorrectly marked on the sketch. This dimension was off by nine inches. Failure to correctly translate the design into the drawing is considered an example of violation of lOCFRSOP Appendix B, Criterion III. Reason for Violation As a result of required piping changes for this modification, a seismic analysis and Stress Package 03378 update were requested by the site. Included with the request were M-107 Sh 2247/2248 which indicated the existing configuration, and proposed modification.

Using the drawings as input 1 the system was modeled on ADLPIPE to generate the system stresses after the modification.

The existing drawings (sent as part of the request) were marked "Issued As-Built per NRC IE Bulletin 79-14.11 After the analysis was performed, a pre-installation walkdown was performed.

During the walkdown the referenced dimensional discrepancy was noted. The seismic analyst was contacted to evaluate the change. As a resultp the analyst issued a letter stating *that since stresses in the area were low, based on his judgement, the change was acceptable.

When the construction was complete, the seismic analyst compared the as-built to the dimensions used in the preliminary analysis.

It was determined the analysis was acceptable with the dimensional variance .... Stress Package 03378 was annotated to reflect this information.

The above-information describes*

the circumstances surrounding.the modification however does not indicate a root cause. The discrepancy is not directly related to the modification except that the modification brought a previous error to light. That is, the drawings used were certified as being dimensionally correct per Bulletin 79-14, when in reality there was an error. Corrective Action Taken and Results Achieved The engineering groups were briefed as to the inspection results. These ings were completed on August 2, 1989. The above noted discrepancy has been satisfactorily

• dispositioned by analysis.

Corrective Actions to be Taken to Avoid Further Non Compliance The. following corrective actions will be taken to prevent recurrence:

Interim Same as that required for Violation Icem l.a. Long-Term

• The "long-term" actions prescribed for Violation Item l .e will prevent rence. MI0789-1683A-TC01-NL02 14 .

..... ,, --.

• . -

• Date When Full Compliance Will be Achieved The dates established for.actions related to Violation Item l.e apply here as well. NRC Violation 255/89007-0lg:

FC-756 "HPSI Pump Miniflow Bypass Modification.eu

[Refer to page 19 of NRC Report 50-255/89007 (DRS).] Example Pipe support drawings in p1p1ng support Calculation No 03378 of FC-756 did not adequately describe the required weld sizes. Pipe support drawings DCl-8198.1 and DC1-Hl96.2 contained in support tion No 03378 were reviewed.

The inspector found that one drawing showed fillet welds at the structural joints but no weld sizes were specified.

The other drawing showed a 3/16 inch fillet weld with a note "assumed." As a result, the design bases of the welds were not adequately translated into the drawings.

CPCo Response As part of the evaluation of this example, M-107 Sh 2254/2255 were reviewed which are detail drawings for the subject hangers. The two ports *cited were not modified or installed as part of FC-756. The supports were only evalua.ted regarding stresses in relation to the modification.

In both cases, the_drawings are Rev 0 and are issued as-built per IE Bulletin 79-14. It appear-s that this is a situation where documentation from the 79-14 effort may not be completely However, when past discrepancies were identified, there was no signficant impact on analytical conclusion.

Neither drawing DC1-H198.l nor DC2-Hl96.2 were utilized as design input to FC-756. After further discussion on this issue with NRC Region III via telecon on July 26, 1989 and review of the drawings referenced by the inspector, it was determined that these drawings were initial IEB 79-14 calculation file ings of preliminary status. These drawings do not represent the final hanger detail drawings referenced above. Since these calculation file drawings are not "record" drawings reflecting as-built condition, and are not referenced (by intent) in our Facility Change Design Document Checklist, they are not input to our facility change process. No further action is required since neither a design change control deficiency nor inaccurate record (as-built) document exists. Therefore, CPCo does not acknowledge this example. However, reference example e. for actions to be taken to ensure accurate dimensions are utilized as* analysis inputs. NRC Violation 255/87007-0lh:

FC-731 "Regulatory Guide 1.97 Transmitter Replacement." [Refer to pages 19 and 20 of NRC Report 50-255/89007 (DRS).] Example The seismic stress calculation assumed an incorrect center of gravity which was not identified during the checking process.

15 . *.*:.,*-*

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• The analysis criteria shown on page 3 required the center of gravity (CG) of the instruments/equipment to be considered in the seismic stress calculationso A review of the rack support bent plate on page 27 found that the CG of the instruments was not considered in the seismic stress calculations.

As a the forces and moments at the rack support attachment were inadequately lated. Reason for Violation The analysis addresses the adequacy of instrument racks inside the containment building.

For the GWO 7906, FC-731 job, the work involved modifying all four instrument racks. Three of the racks are tied together while the fourth one is by itself. The racks are made out of Unistrut attaching to the containment liner plate using bent plates. The instruments are mounted on the mounting plate which in turn is* bolted to the Unistrut.

Analytical error based on the failure to consider the center of gravity is acknowledged.

The reason for this is an error made by the analyst, inadequate technical review and.failure to follow established procedures.

Corrective Action Taken and Results Achieved and the The analysis has been revised to include the center of gravity analytical results represent an acceptable as-built condition.

groups have been briefed as to the results of this* inspection.

were completed on August 2, 1989. All engineering These briefings Corrective,Actions to be Taken to Avoid.Further Non Compliance To prevent recurrence of this or similar discrepancies, the following corrective actions will be taken: Interim Same* as* that required for Violation Item* La. Long-Term The Plant Administrative Procedure will be enhanced by the incorporation of a technical review checklist consisting of a comprehensive set of review "prompts." One of the "prompts" will require that the reviewer ensure that all analysis objectives be carried through to completion.

' In addition, a program will be developed to provide periodic refresher training to all design engineers on design change-related administrative procedures.

Date When Full Compliance Will be Achieved The personal briefings letter will be issued by September 1, 1989. Procedural enhancements, as well as required training on the enhancements, will be pleted by January 1, 1990. The program for periodic refresher training will be in place by March 1, 1990. MI0789-1683A-TC01-NL02 16 *.:.

• NRC Violation 255 /89007-0li:

FC-731 "Regulatory Guide 1. 97 Transmitter Replacement." [Refer to page 20 of NRC Report 50-255/89007 (DRS).] Example The calculated bending stress "fbx" shown on page 27 of the analysis was in error. The 5,645 psi should be 5,976 psi. The checker did not identify this calculational error. Reason for Violation Analytical error based on the inaccurate bending stress is acknowledged.

The analysis has been revised to incorporate the accurate "fbx" value and the analytical results represent an acceptable as-built condition.

Corrective Action Taken and Results Achieved All engineering groups have been briefed as to the results of this inspection.

These briefings were completed on August 2, 1989. Corrective Actions to be Taken to Avoid Further Non Compliance To prevent recurrence of this or similar discrepancies, the following corrective actions will be taken: Interim Same as that required for Violation Item La. Long-Term*

Same as that required for Violation Item l.h with the exception that a "prompt" will be included on the technical review checklist to require that the reviewer verify the accuracy of all analysis calculations.

Date When Full Compliance Will be Achieved The dates specified for Violation Item l.h apply to this item also. NRC Violation 255/89007-0lj:

FC-567 "Core Cooling Instrumentation Modification." [Refer to page 22 of NRC Report 50-255/89007 (DRS).] Example FC-567 did not address the impact of the increased load on the inverters, bypass regulators on the battery chargers.

The inspector observed that the licensee performed calculations to analyze the impact of the increased loading on the preferred AC bus supply breakers, cabling to the preferred busses from their respective inverters and on the DC batteries.

However, no calculations or analyses were evident which addressed MI0789-1683A-TC01-NL02 17 . . :_ . =** *; ..... ... : . . .. **. ::: :.

. :.' ..... -. the impact on the inverters, bypass regulator or the DC system battery chargers.

This resulted in a concern for the capability and capacity of these Class lE systems to perform their safety-related functions.

The inspector concluded that the licensee had failed to employ adequate design controls during the design stage of the facility change in that the full impact of the increased loading was not analyzed.

In response to the inspector's cern, the licensee verified the present loading on the respective inverters and battery chargers which includes the increase resulting from the instrumentation additions.

The inspector concurs that based on the licensee's reported inverter and battery charger outputs, plus the anticipated emergency loading, per the Design Basis document, the inverters, bypass regulator and battery chargers will not be overloaded.

However, the licensee failed to employ adequate design controls which would have included analyses of all impacted components.

Reason for Violation Facility Change FC-567 (Core Cooling Instrumentation) added a Reactor Vessel Level Monitoring System (RVLMS) to the plant design. Addition of this system resulted in an increased load of 600VA on each of preferred busses, YlO and Y20, the associated DC to AC inverters, bypass regulator and DC system. In reviewing this design change, the inspector identified that, although the effect of the increased load on the batteries was determined, the facility change did .not. address the impact of the increased load on the inverters, bypass regulator or the battery chargers * . * . The apparent failure to adequately verify and check design resulted from inadequate documentation of assumptions and engineering judgement utilized to determine the impact of the load additions to the preferred busses. The effect of the load increase on the batteries was determined based on the undocumented assumption that the batteries were the limiting component.

In order to mine the effect of the increased load on the batteries, the new loading on each of the preferred buses and thus the loading on each of the inverters was determined.

No documentation was provided, however, comparing the revised load on the invertors against their design. rating. A similar situation existed for the battery chargers.

The new battery load profile was determined based on the increased loads, however, no documentation of the effect of the new load profile on the battery charges was provided.

Subsequent evaluations have been performed to document that the load additions to the preferred buses performed by FC-567 did not result in overloading inverter, battery charger or bypass regulator.

The results of these evaluations are summarized below: 1. The maximum loadings on the YlO and Y20 buses during emergency conditions are 4378VA and 5456VA respectively.

This includes the loads added by FC-567. The design rating of the invertors is 6000VA and thus the tors are not overloaded.

MI0789-1683A-TC01-NL02 18 ., .... .._: .*

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. ' * -* _._ .. _. . ... ..... * ... 2. The steady state constant DC current requirements during emergency tions is 253 amps for the most heavily loaded battery (Battery No 2) after approximately ten minutes. This is less than the 400 amp combined rating of the two battery chargers connected to each DC bus. The battery chargers thus have sufficient capacity to provide the DC steady state load with capacity remaining for restoration of the batteries following the discharge during the first ten minutes. 3. The bypass regulator is utilized to provide temporary power to a preferred bus from a non-class lE source to allow maintenance to be performed on an inverter.

The initial response made to the inspector regarding operation of the bypass regulator was incorrect.

The bypass regulator is not shed during accident conditions and could be subject to the emergency load. Operation with the bypass regulator energizing the preferred buses is, however, restricted by Administrative Procedures to less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (eight hours for some buses). This restriction minimizes the amount of time that the bypass regulator would be subject to providing power to a preferred bus during accident conditions.

The limiting component of the bypass regulator is the isolation transformer*

This transformer is rated at 5000VA. As discussed earlier, the maximum loading on bus Y20 is 5456VA. Thus the load on the bypass regulator could be exceeded if it were connected to bus Y20 during an emergency condition.

This discrepancy had been previously identified by the Configuration Control Project and Discrepancy Report F-CG-88-002 was initiated.

This discrepancy was quently closed out by assuring that the output voltage of the bypass regulator will be maintained at acceptable levels at up to 150% of the nameplate rating of the tr...an*sformer.

Corrective Action Taken and Results Achieved All engineering groups havebeen briefed on the results of this inspection.

These briefings were completed on August 2, 1989. -An engineering analysis was per.formed documenting that the inverter and

• battery charger were not overloaded as a result of this modification.

-The Configuration Control Project had.previously identified the concern with the bypass regulator and has subseq'uently resolved and closed out the crepancy.

Corrective Actions to be Taken to Avoid Further Non Compliance To prevent recurrence of this or similar discrepancies, *the following corrective actions have or will be taken: Interim Same as that required for Violation Item l.a

• MI0789-1683A-TC01-NL02 19 *.*:-: * .. " .-* *,::: >*'. -:-,,. .. ........ **. *'* ----

• Long Term Upgrades have been initiated to our station load analysis program to account for full aystem impact of load additions.

In the the load carry1ng ability of load carrying components will be assessed in addition to assessing power supplies.

Specifically, the load carrying capability of the battery chargers and preferred power inverters will be assessed, along with battery capacity whenever load is added to the 120V preferred AC system. Periodic training as proposed for Violation Item l.a will feature the ities of modifications support groups such as: Power Resources and Systems Planning (for load addition and -Systems Protection and Planning (for breaker and -Energy Supply Services Civil Section (for structural analyses).

It is expected that this training wil-1 maintain the design engineer's awareness as to what must be taken into account when adding electrical or mechanical load to plant systems. Date When Full Compliance Will be Achieved Personal briefings letter will be issued by September 1, 1989. The station load analysis program upgrades will be completed by September 1, 1989. A gram for the periodic training on the capabilities of support groups will be in place by -March 1, 1990. NRC Violation 25S/89007-0lk:

FC-760-02 "Control Room Emergency Lighting." [Refer to pages 23.and 24 of NRC Report 50-255/89007 (DRS).] Example This FCcontained an unverified assumption in that the assumption that emergency lighting fixtures were rigit was never proven. Engineering Analysis EA-FC-760-2-001 was performed to analyze the mounting of the lighting fixtures to be installed.

Section V of this document, referring to the DC lighting fixtures, states in part "Assume the lighting fixture is rigid **** " This assumption is not justified in the analysis document and, in fact, the fixture (McMasters-Carr Lampholder, Catalog No 1700Kl2) employs a swivel joint. The lighting fixtures are not safety-related, but mounting is considered critical since they are in the control room and failure could endanger personnel or safety-related devices

• MI0789-1683A-TC01-NL02 20 : * .. *:-. -**-. . .. .

-*. ' **:*:.-:***

Reason for Violation The McMasters-Carr Lampholder, Catalog No 1700Kl2 fixture has been used for the control room emergency lighting design associated with The fixture employs a swivel joint for adjusting only. The adjustment is made in one plane only. The mechanism used is a bolted connection and the lamp tion is fixed in place by the friction from tightening the bolt. Tightening the bolt keeps the joint tight in service and keeps it from swiveling.

The assumption of rigidity of the fixture service was based upon the analyst's interpretation of catalog data. That assumption is considered appropriate.

Plant administrative design control procedures required, and currently that all analytical assumptions be documented, acknowledged in terms of icance and technically reviewed (Reference 1). The identified discrepancy results from failure to implement this procedural requirement.

Corrective Action Taken and Results Achieved All e.ngineering groups have .. been briefed as to the results of this inspection.

The briefings were completed on August 2, 1989. Corrective-Actions to be Taken to Avoid Further Non Compliance Interim

• Same* as that required for Violation Item 1.a.

• Long-Terni-

-Develop a program to provide periodic refresher training on "the requirements of plant administrative design change procedures related to engineering analyses.

Date When Full Compliance Will be Achieved The personal briefings letter will be issued by September 1, 1989. The program for periodic refresher training will be in place by March 1, 1990. NRC Violation 255/89007-011:

SC-87-090 Water Leak Detection Set Point. [Refer to page 27 of NRC Report 50-255/89007 (DRS).] Example Specification Change No 87-090 changed the Service Water (SW) leak detection set point from 75 gpm to 300 gpm verifying what size of SW piping break in the containment air coolers would result in a 300 gpm delta-flow alarm

• MI0789-1683A-TC01-NL02 21 -.... * *** *i,.. :. * .. *:. -... . -.

• CPCo Response The containment SW leak detection system monitors SW flow into and out of the reactor building and provides an alarm in the control room when a preset differential flow is exceeded.

SC-87-090 changed the differential flow alarm set point from 75 gpm to 300 gpm. The instrumentation loops for the leak detection system consist of flow elements 1 differential pressure transmitters with square root output and a differential flow switch with a time delay output. A time delay of approximately 15 seconds is incorporated to eliminate nuisance alarms due to flow noise spikes and still allow timely indication of leakage. The SW leak detection system is utilized as a post accident monitor. During accident conditions, without all control rods water leaking inside the containment building can dilute the containment building sump water to a boron concentration low enough to allow the reactor to return to a power state. As noted in Engineering Analysis EA-SC-87-090-1, the basis for the original alarm set point of 75 gpm was engineering judgement.

Further, the new 300 gpm set.point.was selected based on the total inaccuracies of the instrumentation loop, times the full scale flow of the transmitters.

Use of instrument acies within the engineering analysis provides a conservative determination based on instrument capabilities.

As noted in the inspection report, the engineering analysis did not provide justification that the set point meets the design intent of the SW leak tion systeqi..

However, the adequacy of the set point with respect to the tion system.design intent was presented and evaluated as part of the written l0CFR50.5-9

.. (Safety Evaluation) analysis for the SC. The safety evaluation is part of the SC package and was reviewed with other supporting documentation comprising the SC package by the Plant Review Committee (PRC) on March 2, 1987. Therefore, Consumers Power Company does not acknowledge this example as a lation of 10CFR50, Appendix B, Criterion III. NRC Violation 255/89007-0lm:

SC-87-163 "Upgrade Feedwater Flow Transmitters." [Refer to pages 27 and 28 of NRC Report 50-255/89007 (DRS).] Example Specification Change No 87-163 added a series voltage zener diode to the feedwater flow transmitter instrument loop for Transmitter Nos FT-0701 and FT-0703 without specifying the required zener diode design parameters.

Reason for Violation upgraded FW flow transmitters FT-0701 and FT-0703 to Rosemount units. The supply voltage requirements for an 1151 DP transmitter is 12 Vdc to 45 Vdc (4 mA to 20 mA current loop). The transmitter will operate within this voltage range as a function of load resistance.

The load resistance for the FW flow transmitters is approximately 300 ohms. The nominal supply voltage requirements for the transmitter as determined from the Rosemount functional specifications was approximately 19 Vdc. MI0789-1683A-TC01-NL02 22 *__:_-.*-:-

• • .. ,._ *. ,. : o:. *..*.. * .. **:

• .* , ..... *. :, .. ,._ *,*

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• As part of the SC, a zener diode was installed -in the series current loop to lower the power supply output voltage to the operating voltage of the Rosemount flow transmitter.

During development of the SC, the design criteria for the zener diode, that is the required voltage was determined to be 11 Vdc. This design criteria is shown on Drawing F-69 Sh 1, Rev 22 of SC-87-163.

As a result of this criterion being stated within the SC package, the proper zener diode was installed and as stat-ed in the inspection "the zeners were performing their function." Therefore, Consumers Power Company does not specifically acknowledge-this example as stated. While the design criterion was detailed sufficiently within the SC to provide for installation of the proper zener diode, Consumers Power Company acknowledges the need for design packages to contain documentation which provides the bases for engineered changes. The failure to include the required enigneering analysis which served as the basis for the design criterion presented within SC-87-163 has been attributed to a weakness within the SC process regarding documentation of engineered decisions.

Corrective Actions Taken and Results Achieved In that the proper zener diode was prescribed and installed, and resulted in the equipment affected by the modification being capable of performing their design function, no immediate corrective actions have been undertaken.

All engineering groups were briefed on the results of this inspection.

The briefings were completed on August 2, 1989. Correctiv.e .Actions to be Taken to Avoid Further Non Compliance Interim Same as that required for Violation Item 1.a. Long-Term To ensure that adequate bases are developed to justify the change and that these bases are technically reviewed and documented within the specification change package, plant *administrative procedures (Reference

5) will be revised either to require that a formal engineering analysis (per Reference

1) or a new SC change justification form be utilized for the following:

To provide a reason for the change (in part by describing why the existing condition is less than desired and why the change will improve as-built dition), ., *:ra describe the design basis function of the system within which this change is being made and justification that this function will be maintained, -To identify the full impact change will have on the system within which this change is being made and on potential interfacing systems, MI0789-1683A-TC01-NL02 23 ;---* . *:..:.::--.**

.*--... :. .. :.

,, .. ,, .. -....... ;:., :*. * .. **.:::;1-'* . :"'.*

• .. * *.*. .... :. -To identify critical functional or physical features that must be met by the change to achieve the desired as-built condition (this may require formal engineering analysis per Administrative Procedure 9.11), and -To describe how these critical features will be verified (eg, inspection or test). Date When Full Compliance Will Be Achieved The personal briefings letter will be issued by September 1, 1989. The revision to administrative procedures will be completed by January 1, 1990. In addition, a program will be developed by March 1, 1990 to provide engineers with periodic refresher training on SC-related administrative procedures.

NRC Violation 255/89007-0ln:

SC-88-069 "Upgrade Safety Injection Tank Pressure Transmitters." [Refer to pages 29 and 30 of NRC Report 50-255/89007 (DRC).] Example Specification Change No 88-069 added a series voltage regulating zener diode to the safety injection tank. pressure transmitter instrument loops for Transmitter Nos PT-0361, 0367 , 0369, and 0371 without specifying the required zener diode design parameters.

Reason for_Violation SC-88-069 safety injection (SI) tank pressure transmitters, PT-0363, PT-0367, ..PT-0369 and PT-0371 to Rosemount units. This modification, like SC-87-163, introduces a zener diode in series current loop to lower the power supply output voltage to the operating voltage of the Rosemount pressure mitter. During development of the SC package for this modification, engineering analyses.

were performed to* determine the design criterion for the zener diode. However, as evidenced by the transmitter voltage measurements taken during the inspection, an error was made .in the analysis.

This error was not identified during design reviews of the modification package due to the lack of a mented engineering analysis within the SC package. Further, after modification installation, no preoperational testing specific to transmitter operating age was conducted.

Therefore, the failure to attain a completed modification with all equipment operating within manufacturer prescribed operating ranges has been attributed to weaknesses within the Specification Change process regarding documentation of engineered options and adequate preoperational testing. Corrective Action Taken and Results Achieved The power supply output voltage, zener diode vuic:age and transmitter voltage for all the upgraded Rosemount transmitters associated with SC-88-069 were measured.

As indicated within the inspection report, the transmitters were found to be operating outside their nominal operating of 14 Vdc to 45 Vdc by.up MI0789-1683A-TC01-NL02 24 .. * '* .,. **:*

• • to 12.62 Vdc. As a result of this finding, all other installed transmitters having zener diodes in their circuit had power supply, zener diode and mitter voltages measured.

From these measurements, two additional non-safety related transmitters (PT-5117 and PT-0927) were identified to be operating outside their prescribed nominal* operating range. Due to these findings, SC-89-162 was generated to replace the improper zener diodes. As part of this modification package, an engineering analysis was completed and technically reviewed to assure proper zener diode selection and to provide documentation of design criterion.

The analysis was completed on August 1, 1989. Additionally, work orders were generated on June 5, 1989 to inspect the transmitters that were operating outside their nominal operating range. Presentations to all engineering groups have been conducted to. brief engineers as to the NRC engineering team inspection results. These presentations were completed on August 2, 1989. Corrective Actions to be*Taken to Avoid Further Non*Compliance Interim Personal letters will be sent to all engineers by September 1, 1989 describing the NRC observed weaknesses and requiring that the engineer look at SC's rently being engineered for similar problems.

Long Term -The plant administraive procedure (Reference

5) revisions described for tion l.m apply as do the following:

-Revise plant administrative procedures (Reference

1) to provide the technical reviewer of an engineering analysis a checklist to assure a thorough, accurate and auditable analysis.

The checklist would feature a set of "prompts" in part to verifyall analytical input, assumptions and calculation.

-Revise administrative procedures (Reference

5) to require that pre-operational testing be specified as part of SC engineering either in a work request or test procedure prior to technical review of the SC engineering package. In addition, require that the test specification align with the critical features identified as part of the documented change basis (see procedure changes identified for Violation Item l.m). Date When Full Compliance Will be Achieved Administrative procedures will be revised by January 1, 1990. Training on the procedure revisions will also be complete on January 1, 1990. In addition, a program will be in place by March 1, 1990 to provide periodic refresher training on SC-rela.ted procedures.

SC-89-162 will be performed by November 15, 1989. The work orders to inspect the affected transmitters will be completed by December 1, 1989. MI0789-1683A-TC01-NL02 25 ; .....

NRC Violation 255/89007-0lo:

SC-88-069 "Upgrade Safety Injection Tank Pressure Transmitters." [Refer to pages 29 and 30 of NRC Report 50-255/89007 (DRS).] Example Specification Change No 88-069 did not consider the effect of instrument loop loading on the power supply; as a result, the load adjustment resistor setting which matches impedance for maximum power transfer was not specified or adjusted.

Reason for Violation SC-88-069 upgraded safety inJection (SI) tank pressure transmitters, PT-0363, PT-0367, PT-0369 and PT-0371 to Rosemount units. This modification, like SC-87-163, introduces a zener diode in series current loop to lower the power supply output voltage to the operating voltage of the Rosemount pressure mitter. While reviewing this SC the inspector reviewed the SI tank pressure loop power supply manual. As-stated intheinspectionreport; "the Foxboro Model 610A power supply is designed to furnish power to a single electronic transmitter.

The nominal DC output voltage is 80 volts. The manual also states that the output load resistance must be 600 ohms +10; -20 percent. The SC package did not determine the load resistance.

The manual provided detailed instructions to sum the input resistances of all the receivers in the loop (excluding the and to adjust the load adjustment dial on the power supply to the difference,,between the loop resistance and 600 ohms. Subsequentcto the inspection on July 25, 1989, plant engineering personnel contacted the power supply vendor to discuss the inspector's concern regarding the affects of increased load resistance on the power supply. During this conversation the vendor noted that the specific requirement for a load tance of 600 ohms applies only to Foxboro transmitters connected to Foxboro power supplies and that applied power supply load resistance is based on the voltage requirements of the associated transmitter.

The voltage requirements of the Rosemount transmitters installed under SC-88-069 are addressed in the modification package, however, documentation was not provided regarding resultant.

power supply l6ad resistance.

Failure to include applicable documentation within the modification package has been attributed to a lack of guidance being provided within Administrative Procedure 9.04, fication Changes." Corrective Action Taken and Results Achieved Presentations of the inspection results were made to all affected engineering groups. These presentatioris were completed on August 2, 1989. Corrective Actions to be Taken to Avoid Further Non Compliance Personal letters will be sent to all engineers describing the NRC engineering inspection results by September 1, 1989. The letters will require that neers review SC packages currently being engineered for similar problems.

MI0789-1683A-TC01-NL02 26 -:* : .. \_._ .:: ;*'.. ,. -** :=. * .. : ** _.-:-: .*

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• -The plant administrative procedure revisions (and training) described for lation Items l.m and l.n effectively respond to this item also. Date When Full Compliance Will be Achieved Administrative procedures will be revised by January 1990. Training in the procedure revisions will also be complete on January 1, 1990. In addition, a program will be in place by March 1, 1990 to provide periodic refresher ing on SC-related procedures.

NRC Violation 255/89007.0lp:

SC-88-102 "Upgrade Containment Pressure Transmitter PT-1812." [Refer-to pages 31 and 32 of NRC Report 50-255/89007 (DRS).] Example Specification Change No 88-102 installed a different model containment pressure transmitter for Transmitter No PT-1812 without performing a seismic analysis to determine the acceptability of installing the new transmitter on the old mounting.

Reason for-Violation SC-88-102 upgraded containment building pressure transmitter, PT-1812 to a Rosemount pressure transmitter.

The pressure loop affected by the modification provides indication only and is not required to be operable for any analyzed event. The pressure transmitter is mounted off piping associated with ment Penetrcation MZ-17 and is physically located between the manual instrument isolation valve and the manual containment isolation valves. The manual instrument isolation valve is maintained open to allow pressure transmitter operation.

Therefore, the primary containment boundary includes PT-1812. While processing SC-88-102, engineering personnel

• failed to identify that the pressure transmitter constituted part of the containment boundary.

This ure is attributed to the following factor: The administrative procedure for Specification Changes (Reference

5) requires that the engineer consult the Equipment Data Base (EDB). The EDB-Q-Listing identifies the pressure retaining and structural (seismic) requirements to be met by the equipment.

The existing Q-Listing in the EDB for PT-1812 indicates that the transmitter function is not safety-related, there are no pressure retaining requirements, and that the structural mounting is not safety-related.

This specific Q-Listing needs to be reviewed and revised as necessary.

Given accurate EDB information, the existing_

SC checklist "prompts" which also existed at the time this deficiency occurred, are sufficient to identify the governing design codes, standards and regulatory guides to be complied with. Corrective Actions Taken and Results Achieved A formal seismic engineering analysis has been initiated to document the adequacy of the existing transmitter mounting and the associated tubing. MI0789-1683A-TC01-NL02 27 ;..&.. ',' :*,* : .:*:

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--;_,.

The results of the inspection have been presented to all engineering groups. These presentations were completed on August 2, 1989. Corrective Actions to be Taken to Avoid Further Non Compliance The existing Q-List interpretation for PT-1812 will be reviewed for accuracy and revised as necessary.

In addition, if it is determined that the tation is in error, other interpretations will also be reviewed to identify the breadth of the discrepancy.

These additonal reviews will cover, as a minimum, interpretation for other instrumentation serving pressure retaining functions.

If additional reviews indicate the need, additional clarification in tive P.rocedures related to Q-List interpretation (Reference

6) will be provided and engineers will be trained. Further, a review will be conducted to ensure the seismic qualification of other similar configurations.

In addition, a program to provide periodic refresher training on procedures related to Q-Listing will be developed.

Finally, a portion of the Configuration Control Project involves the tion of the Q classification for approximately 16,000 components in the Plant's equipment data base. This activity is currently scheduled to be completed by the end of-1990 and will provide a sound technical basis for future tions. Date When F.ull Compliance Will Be Achieved The existing Q-List interpretation for PT-1812 will be reviewed for accuracy and revised necessary) by September 15, 1989. If it is concluded that the PT-1812 interpretation is in error, interpretation for other similar tions will be completed by November 1; 1989. If these additional reviews tate the need for procedural clarification, the procedures will be enhanced by January 1, 1990 and all engineers*

will be trained on the enhancements by this date. The program for periodic refresher training on Q-Listing will be in place by March 1, 1990. The additional seismic review will be completed by October 1, 1989. NRC Violation 255/89007-0lg:

EA-FC-722-10 "N2 Backup Test Evaluation for Station 5," February*21, 1987. [Refer to page 15 of NRC Report 50-255/89007 (DRS).] Example The stated that the nitrogen usage rate was 32.5 psig AP/hour based on the test results from Functional Test T-FC-722-501-01.

However, the test results failed to account for the post test calibration shift of 5 psig for of the pressure gauges. By incorporating this additional factor, the usage rate is increased to 33.75 psig AP/hour. MI0789-1683A-TC01-NL02 28 . *. . -:* ** '7. *. ,-'* .. .*,* . ,* ' *.*.:* ... * *,'* *. **-:*** ."'-*,' .. * .** .... *.,_ ..

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. ' *

• Using the above rate in the calculation reduces the "actual operating period" from 10.3 days to 9.93 days. This is below the assumed acceptance limit given in the original calculationo No safety significance was attributed to this occurrence; however, the instrument accuracy requirements specified in the test procedure were inadequate as noted belowo -Procedure No T-FC-722-0501, "CV Air Supply -N2 Backup Performance Test," Revision O, February 6, 1987. Under Special Tools/Equipment, a 0-3000 psig pressure gauge is called for. The accuracy specified is +/- 2% minimum. This equates to a +/- 60 psig accuracyo The acceptance criteria for three of the four nitrogen stations ranged from 24 psig to 68 psig over the four hour span of the performance test. CPCo Response CPCo does not acknowledge this example as a violation of Appendix Criterion III Design Control," based upon the following.

1. Page 6 of 32 of "Palisades Nuclear Plant Modification Procedure No T-FC-722-501," and "Temporary Change to a Change No FFC-87-006, specified calibrated analog pressure gauges, 0-3000 psig, +/- 2% minimum accuracy and that these gauges shall be calibrated in accordance with 2.4, reference paragraph 6.1.5. 2. The intent of specifying a minimum accuracy of the test gauges was to allow qualified test personnel the. flexibility to utilize test gauges of a higher degree"of accuracy if available.

3. The intent of Reference 2.4 (Palisades Nuclear Plant Administrative dure S.07, "Control of Measuring of and Test Equipment"), paragraph 6.1.5, is to require performance of pre-and post-calibrations of the test gauges. These calibrations were performed as Pre-and Post-Calibrations of the gauges are utilized to determine/verify the actual gauge accuracy as utilized during the test. 4. As stated in paragraph 1 of page 16 of NRC Report No 50-255/89007 (DRS), "Additional reviews by the inspector disclosed that the pressure gauges actually used has a specified accuracy of +/- 1%. In addition, pre-test and post-test calibration data indicated that the actual accuracy was closer to +/- 0.1%." This statement reinforces the intent of specifying and the requirement to perform pre-and post-calibrations (reference Item 83) of the gauges. 5. Acceptance criteria for Palisades Nuclear Plant Modification Procedure No T-FC-722-501 are established via calculation and are not affected by gauge inaccuracies which are linear and constant throughout the test range

• 29 ... .: . ' . . ...... *--*

• Based upon the above the specification of test gauges, 0-3000 psig, +/- 2% accuracy was appropriate and in accordance with Palisades Nuclear Plant Administrative Procedures--.

Plant administrative design control procedures (Reference

2) required, and currently require, that modification test procedures feature requirement

-The use of calibrated test equipment of the proper range and accuracy to determine conformance to specified acceptance criteria, -Test equipment be identified along with its calibration status, and -Acceptance criteria (with appropriate tolerances) be specified to effectively determine whether critical design requirements have been satisfied.

Thus, no corrective action is deemed necessary.

NRC Violation 255/89007-0lr:

SC-87-163 "Upgrade Feedwater Flow Transmitters." [Refer to pages 27 and 28 of NRC Report 50-255/89007 (DRS).] Example Specification Change No 87-163 added a series voltage regulating zener diode to the FW flow transmitter loop for Transmitter Nos FT-0701 and FT-0703 without specifying

__ the measurement .of. the power supply, zener, and transmitter voltage as acceptance*

criteria to determine if the transmitter loop was operating within its-design limits. Reason for Violation SC-87-163 upgraded FW flow transmitters FT-0701 and FT-0703 to Rosemount units. The supply voltage requirements-for a 1151 DP transmitter is 12 VDC to 45 VDC (4 mA to 20 mA current loop). The transmitter will operate within this voltage range as a function of load resistance.

The load resistance for the FW flow transmitters is approximately 300 ohms. The nominal supply voltage requirement for the transmitter as determined from the Rosemount functional specifications was approximately 19 Vdc. As part of the SC a zener diode was installed in the series current loop to lower the power supply output voltage to the operating voltage of the Rosemount flow transmitter.

During the inspection, the NRC inspector identified that the SC package did not contain post installation power supply output voltage urements.

Further, it did not contain zener diode and transmitter operating voltages following modification.

The failure to adequately specify necessary preoperational testing requirements on the work orders which implemented the SC has been attributed to weaknesses within Administrative Procedure 9.04. Currently, no guidance exists as to the type of which may be appropriate, nor does the procedure specify the need to document testing performed on implementing work orders or within the SC package.

30 . *** ..............

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.. *. . .. ., .. ... ......

Corrective Actions Taken and Results Achieved As noted within the inspection reportp the power supply output voltage, and the zener diode and transmitter operating voltages were measured.

From these urements it was determined that all components were performing their design function within manufacturer specifications.

Presentations have been made to engineers discussing the results of the recent NRC engineering inspection.

These presentations were completed on August 2, 1989. Corrective Action to be Taken to Avoid Further Non Compliance Personal letters will be sent to all engineers on or before September lp 1989 describing the results of the NRC inspection and requiring that SC's currently being managed be reviewed for similar problems.

Date When Full Compliance Will be Achieved The procedure revisions for Violation Items l.m and l.n will effectively respond to this item. NRC Violation 255/89007-0ls:

SC-88-069 "Upgrade Safety Injection Tank Pressure Transmitters." [Refer to pages 29 and 30 of NRC Report 50-255/89007 (DRS).] NRC Identi£ied Discrepancy Specificai:ion Change No 88-069 added a series voltage regulating zener diode to the safety injection tank pressure transmitter loops for Transmitter Nos PT-0363, 0367, 0379, and 0371 without specifying the measurement of the power supply, zener, and the transmitter voltage as acceptance criteria to determine if the transmitter loop was operating within its design limits; and also did not specify acceptance criteria for determining the acceptability of changing the load adjustment resistor in the power supply. Reason for Violation Consumers Power Company's response regarding the failure to specify acceptance criteria to determine if the transmitter loop was operating within its design limits in the preoperational stage is provided in our response to Violation Item l.m. In regard to the post modification stage of this SC, the failure to establish a program to periodically measure the pressure transmitter loop voltages has been attributed to plant personnel not considering all potential failure modes and effects in the circuit design. Acceptance criterion for determining the acceptability of changing the load adjustment resistor in the power supply were not specified in the SC package. The manual for the Foxboro 610A power supply stated that the output load resistance for the power supply must be 600 ohms + 10; -20 percent. In matory conversations with the vendor on July 25, 1989, the requirement for load resistance was said to be based on transmitter limitations, not power supply limitations.

The new Rosemount transmitters installed per SC-88-069 do MI0789-1683A-TC01-NL02 31 . . : ' :* -. . : -. *-: ... ... , .... ....

not have this load restriction and hence do not have acceptance criteria as delineated in the manual. Therefore this item by itself is not a violation of 10CFR50-, Appendix B, Criterion III. It is noted however that the new Rosemount transmitters have voltage limitations and this is discussed in our response to Violation Item l.n. Corrective Actions Taken and Results Achieved Same as that taken for Violation Item l.n. Corrective Actions to be Taken to Avoid Further Non Compliance Procedural revisions and tra1n1ng described for Violation Item l.n will ively respond to this item. Additionally, preplanned and periodic control sheets (preventive maintenance activities) will be established to provide for periodic measurements of loop voltages.

Date When Full Compliance Will be Achieved The control sheet program will be established by October 1, 1989. Violation

'255/87007-02a-b) lOCFRSO, Appendix B, Criterion X as implemented by the Palisades Operations Quality Assurance Program requires, in part, that a program for inspection of activities-,affecting quality be established and executed by or for the zation performing the activity to verify conformance with the documented instructions, procedures, and drawings for accomplishing the activity and that examinations, measurements, or tests of materials or products processed be performed for each work operation where necessary to assure quality. Contrary to the above: This is a Severity Level IV Violation.

NRC Violation 255/89007-02a:

CPCo Drawing M-101 Sheet 5113, Revision O, "Piping Isometric, Auxiliary Feedwater Control Valve CV-0736A and CV-0737A Bypass Piping." [Refer to pages 12 and. 13 of NRC Report 50-255/89007(DRS).]

Example A secondary aspect, associated with the socket welds, pertains to the quality control (QC) inspection of the completed fillet welds. The RIC forms have a column for "QC verification" but for the socket welds in question, the size of the fillet welds was not inspected by QC. Line No 16 of the RIC form, which specifies the weld, size, gap, and type of joint was marked "NA" (not applicable) for all the welds in question under the QC Verification column. Although all of the welds received a Nondestructive Testing (NDT) Visual Examination (VT), it is not clear if the size of the welds was verified during these examinations.

Since the size of the socket fillet welds was not specified on the drawing, nor noted on the RIC form, the NDT examiner would MI0789-1683A-TC01-NL02 32 ** : * '."'!'* *.* :: * .. * ..... ' ..... : ..... '.'-:_: ....... *.** -:-*:**.**

• • . ,

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• have had to determine the required size in the same manner as previously described for the welder. No notation of size nor record of the size calculation was in the documentation provided with the NDT-VT data. In addition, the VT report did not list fillet weld gauges under "Visual Aids Used" giving further indication that the size of the welds was not checked. As a point of clarification, it should be noted that the VT performed on the socket fillet welds was in accordance with American Welding Society (AWS) Dl.l requirements.

This is a structural welding code and allows portions of fillet welds to be undersized by 1/16". This is inconsistent with the requirement of ANSI 831.1, Power Piping Code which specifies minimum fillet weld sizes. If the size of the-socket fillet welds was verified by the stated VT examinationp it cannot be assured that the weld meets the ANSI 831.1 Code requirements.

Reason for Violation The failure to merit conformance of the size of the socket fillet welds has been attributed to a lack of engineering input to and technical review of the maintenance planning for the welding process. Prior to actions taken as a result of recent self-identified failures to verify weld size (Reference 7), no specific requirements existed to verify characteristics (weld, type, size contour) of installed welds. Although Nuclear Operations Department Standards suggest inspection hold points for weld installation verification, working level administrative procedures did not specify:a hold point requirement except for fit up. Corrective'"Action-Taken and Results Achieved Presentations to all engineering groups have been conductep to review the results of this inspection.

These presentations were completed on August 2, 1989. -The Inservice Inspection (ISI) Section oP the Projects Engineering Department has assumed the role of Design Authority for weld engineering by revising the RIC to technically review the maintenance planner's specifications.

The purpose of the review is to ensure that appropriate welding codes are complied with in the areas of weld installation and post-installation examination.

-The RIC has been revised to issue the-weld minimum leg length to the field. This will eliminate the need for the field welder to calculate the length. The aforementioned ISI review will assure that this specification is provided.

-Reference Violation 255/89007-0lc for other applicable actions being taken. Corrective Actions to be Taken to Avoid Further Non Compliance Specifying welding requirements (such as applicable code, weld material, weld type and weld size) is an engineering function.

If properly administered by procedure, the maintenance planner can (and has) effectively prescribe welding MI0789-1683A-TC01-NL02 33 :. . . ' . : . -; : . ':* *:-. . . ': . *. *: **.::-. ,._ *. ,*,_ ... , ..

details for the field provided that adequate input from engineering exists as a basis. In the past, engineering input has been limited to welding tion and/or structural analysis engineering sketches which have lacked size dimensions for the welds. As a result 11 the planner has failed to provide the proper size on the Repair Inspection Checklist (RIC) thereby requiring the field welder to determine and install the proper weld size. This practice fails to meet current expectations for control of design change implementation.

Although plant administrative design control procedures required and currently require that the design change project engineer determine code requirements for assigned projects (Reference 4), and plant maintenance procedures required and currently require that the maintenance planner specify applicable code and weld parameters after consultation with the Engineering Department (Reference 3), these procedures had not been effectively integrated to support one another to ensure that weld specifications from engineering were accurately translated into installation planning, installation, and post-installation verification.

As a result, the following actions have been/will be taken to prevent rence: Interim Same as that required for.Violation Item.l.a.

Long-Term

-Enhancements to .plant design.control and maintenance procedures will be made to more effectively integrate engineering into weld specification and mately -into weld planning and verification:

Appropriate welding codes will be included in the Design Input Checklist (Reference

2) to prompt the design engineer to specify appropriate weld requirements (for installation and examination) in the facility change package as part of both conceptual and detailed engineering.

-Design control procedures related to engineering analyses (Reference

1) will explicitly require that all drawings accompanying structural/seismic analyses provide detailed weld information (type, size, material) for input to the planner. In addition, the procedures will require that sizing culations be performed as part of the analysis.

Finally, a technical review checklist will be provided to require that the reviewer ensure that weld information be accurately represented on the analysis drawings.

Plant maintenance procedures (Reference

3) will require that the maintenance planner utilize the contents of the facility change package to complete the RIC in specifying for the field weld installation and examination ments. The procedure will require that the planner consult the Design Input Checklist and structural/seismic engineering analyses.

Interim actions related to changes to the RIC and ISI group review of the RIC (as described above) will remain in effect. MI0789-1683A-TC01-NL02 34 *. *, -;.*; .. *:. ' ....... .. *' .... -. . *.::*****

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• -Design and quality assurance engineers will be trained on the appropriate structural and piping weld codes and their application to weld installation and examination.

The engineers will also be trained on the above procedural enhancements.

A program will be developed to periodically train design and quality assurance engineers on the aforementioned codes and their application, and on the related design control and maintenance procedures.

In summary, it is expected that these actions will ensure that proper welding requirements (type, material, size) are specified by engineeringp planned by maintenance (with a check on planning by engineering)p and in turn verified by quality control. Date When Full Compliance Will be Achieved The personal briefings by letter will be issued prior to September lp 1989. Procedure enhancements and required training on the enhancements will be pleted by January 1, 1990. The program for periodic refresher training will be developed by March lp 1990. NRC Violation 255/89007-02b:

SC-89-072 (Deviation Report D-PAL-89-043).

[Refer to page 32 of NRC Report 50-255/89007 (DRS).] Example This report documented the undersized fillet welds on socket welded fittings -for SC-89-072.

This specification change was necessary to provide an interim solution to primary coolant system leakage from cold leg drain valves. The change required the of a new length of two inch schedule 160 pipe with a socket welded cap on each of the four loop drains. Inspection of all eight socket fillet welds indicated that none of them met the Code required size of 3/8 inch. During the inspector's review* of the deviation report, there were several concerns that apparently were not addressed.

First, although the corrective actions appear to recognize that the current RIC form does not give the welder sufficient information (specifically the size of the fillet weld), there was no recognition that QC did not and was not required to verify the size of the fillet weld. The.undersized condition was not discovered until the authorized inspector (AI) pointed it out to the licensee.

All of the welds had been reviewed and by the licensee's program and yet the size had never been verified.

This is considered another example of violation of 10CFR50, Appendix 8p Criterion X, in that the size of the socket fillet welds was not verified (255/89007-02b).

Reason for Violation Specifying welding requirements (such as applicable code, weld material, type and weld size) is an engineering function.

If properly administered by procedure, the maintenance planner can (and has) effectively prescribe welding MI0789-1683A-TC01-NL02 35 . *.".,'T:'

• .*. ': .. *: ***-._ .. ___

• • .,. . ... . * .. *.:* *. '. . .. --. _.,*.

• details for the field provided that adequate input from engineering exists as a basis. In the past, engineering input has been limited to welding tion and/or structural analysis engineering sketches which have lacked size dimensions for the welds. As a result, the planner has failed to provide the proper size on the Repair Inspection Checklist (RIC) thereby requiring the field welder to determine and install the proper weld size. This practice fails to meet current expectations for control of design change implementationo Corrective Action Taken and Results Achieved -Presentations to all engineering groups were conducted to brief engineers as to the results of this inspection.

The presentations were completed on August 2, 1989. -The Inservice Inspection (ISI) Section of the Projects Engineering Department has assumed the role of Design Authority for weld engineering by revising the RIC to technically review the maintenance planner's specifications.

The purpose of the review is to ensure that appropriate welding codes are complied .with in the areas of weld installation and post-installation examinationm

-The RIC has been revised to issue the weld minimum leg length to the fieldo This will eliminate the need for the field welder to calculate the length. The aforementioned ISI review will assure that this specification is provided.

Corrective Actions to be Taken to Avoid Further Non Compliance Although .plant administrative design control procedures required and currently require that the design change project engineer determine code requirements for assigned projects (Reference 4), and plant maintenance procedures required and currently require that the maintenance planner specify applicable code and weld parameters after consultation with the Engineering Department (Reference these procedures had not been effectively integrated to support one another to ensure that weld specifications from engineering were accurately translated into installation planning,.

installation, and post-installation verification.

As a result, the following actions have been/will be taken to prevent rence: Interim Same as that required for Violation Item l.a. Long-Term

-Enhancements to plant design control and maintenance procedures, and to ESS Departmental guidelines will be ***ade by January 1, 1990 to more effectively integrate engineering into weld specification and ultimately into weld ning and verification:

-Appropriate welding codes will be included in the Design Input Checklist (Reference

2) to prompt the design engineer to specify appropriate weld requirements (for installation and examination) in the facility change package as part of both conceptual and detailed engineering.

MI0789-1683A-TC01-NL02 36 . , ... 'i--: ' .. . .. ' ....... ;** _. **:*: I

. '* ,. ' .. Design control procedures related to engineering analyses (Reference

1) will explicitly require that all drawings accompanying structural/seismic analyses provide detailed weld information (type, size, material) for input to the planner. In addition, the procedures will require that sizing culations be performed as part of the analysis.

Finally, a technical review checklist will be provided to require that the reviewer ensure that weld information be accurately represented on the analysis drawings.

-Plant maintenance procedures (Reference

3) will require that the maintenance planner utilize the contents of the facility change package to complete the RIC in specifying for the field weld installation and examination ments. The procedure will require that the planner consult the Design Input Checklist and structural/seismic engineering analyses.

-Interim actions related to changes to the RIC and ISI group review of the RIC (as described above) will remain in effect. -Design and quality assurance engineers will be trained on the appropriate structural and piping weld codes-and their application to weld installation and examination.

The engineers will also be trained on the above procedural enhancements.

A program will be developed to periodically train design and quality assurance engineers on the aforementioned codes and their application, and on the related design control and maintenance procedures.

In is expected that these actions will ensure that proper welding requirement-s (t-ype, material, size) are specified by engineering, planned by maintenance (with a check on planning by engineering), and in turn verified by quality control *. Date When Full Compliance Will be Achieved *The personal briefings by letter will be issued prior to September 1, 1989. Procedure enhancements and required training on the enhancements will be pleted by January 1, 1990. The program for periodic refresher training will be developed by March 1, 1990. NRC Violation 255/89007-03:

SC-87-344 Low Temperature Over Pressure Set Points. [Refer to page 28 of NRC Report 50-255/89007 (DRS).] Technical Specification (TS) No 3.1.8.1.a requires a low temperature sure (LTOP) power operated relief valve (PORV) lift setting of < 310 psia for Tc < 300°F and TS 3.1.8.1.b requires a LTOP PORV lift setting 575 psia for Tc < 430°F. Contrary to the above, between August 9, 1988 and February 27, 1989, the PORV as-left setting exceeded the TS requirement on 17 occasions.

This is a Severity Level IV violation.

MI0789-1683A-TC01-NL02 37 *.-* ***-**-;* . * '* .* .. *.** ...... :* :,-* ;;;. ..... : .. :* .. , .

., I .. * * :* .. * .. Reason for Violation changed the LTQP protection system set points for temperature switches TS-0115 and TS-0125. The LTOP system provides primary coolant system {PCS) overpressure relief capability to protect the reactor vessel from the potential for brittle fracture.

The Palisades LTOP system is a two channel system which relieves PCS pressure through either of two PORV's. Channel A relieves through PRV-1042B and channel B relieves through The system is enabled at two settings.

When the PCS cold leg temperature is less than or equal to 300°F, the lift set point for the PORV is less than or equal to 310 psia. When the PCS cold leg temperature is greater than 300°F but less than 430°F, the set point for PORV opening is less than or equal to 575 psia. Above 430°F the LTOP system is not required to be enabled. The LTOP system set points are derived from plant heatup and cooldown limits specified in Plant Technical Specifications.

The set points reflect the ature and pressure limits calculated according to the requirements of Appendix G to 10CFR50, using the methodology provided in Regulatory Guide 1.99, Revision 2. These set points were enacted with the issuance of Amendment 117 to the Palisades operating license on November 14, 1988. At the time the 310 and 575 psia LTOP PORV set points were proposed on the Technical Specification change request which resulted in the issuance of Amendment 117, existing Technical Specifications did not recognize the need for LTOP above_300°F.

Instrumentation existing at this time did not operate above 600 psia had a recognized accuracy of +/- 22 psia. Therefore, the 310 and 575 psia points were selected to provide the maximum practical operating window allawed by exi.sting plant components while remaining bound by 10CFR50 Appendix G limits. The proximate cause of this condition is that the set point value which results from the addition of instrument inaccuracies is not conservative with the lift point specified in Technical Specifications.

This condition has been attributed to poor documentation within the Technical Specifications regarding the fic lift point value. When the technical specification value was derived, Engineering personnel subtracted instrument inaccuracies from the 10CFR50 Appendix G limit and arrived at the 310 and 575 psia set points found in Technical Specifications.

The intent of the Technical Specification lift point value is to ensure compliance with Appendix G. The typical set point methodology, if applied to this situation, would be to provide the applicable Appendix G limit in TS and then control the actual set point, adjusted for instrument inaccuracies, through Technical Specification Surveillance dures. As noted in the inspection report, the issue was identified in parallel by both the and plant personnel.

At the plant, the issue was identified during a review of the set point methodology process utilized at Palisades.

Plant Engineering personnel identified that the PORV lift point had been set at the technical specification values of 310 and 575 psia. Setting the lift points at the technical specification value, neglecting instrument accuracies, could result in the actual lift points being 332 and 597 psia when maximum instrument inaccuracies are accounted for. A review of past performances of MI0789-1683A-TC01-NL02 38 * '

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. : , ........ ,-:*. Technical Specification Surveillance Procedures M0-27A through D which provide for functional testing of the LTOP system, revealed that 29 of the 31 times lift set points (310 or 575 psia) were checked, the set point was greater than the technical specification limitc While the lift point did exceed the technical specification limit, it was within the acceptance values provided by 10CFR50 Appendix Ge Corrective Actions Taken and Results Achieved Plant Engineering personnel reviewed the basis for Technical Specification 3.1.8.1 and Technical Specification Surveillance Procedures which set the PORV lift points and verified that even if the largest positive instrument inaccuracy was added to the technical specification lift point, the 10CFR50 Appendix G limit would not be exceeded.

Upon further review it was additionally identified that the curve utilized in defining the Appendix G limit has incorporated a 30 psia measurement inaccuracy.

In that a Technical Specification change request is being prepared for submittal in support of LTOP protection system modifications to be performed during an upcoming maintenance outage, a letter of interpretation was submitted to the NRC on July 12, 1989 which presented Consumers Power Company's position regarding continued compliance with 10CFR50 Appendix G. Technical Specification Surveillance Procedures M0-27C and M0-27D 9 which provide setting and the PORV lift set points were revised on May 11, 1989 to remove the + 22 psia tolerance.

Corrective_Actions to be Taken to Avoid Further Non Compliance A Technical Specification change request will be submitted which delineates the requi.red PORV lift set points to assure continued compliance with 10CFR50 Appendix G limits following LTOP protection system modifications.

• An tion of the Technical Specification change request development process is being undertaken to determine where enhancements in the review process are required to preclude future occurrences.

Date When Full Compliance Will be Achieved Continued compliance with the lift set point value specified in the Technical Specifications has been assured by submittal of Consumers Power Company's letter dated July 12, 1989 and the rev1s1ons to M0-27C and M0-27D. The cal Specification change request supporting the planned LTOP protection system modifications will be submitted by October 1, 1989. The evaluation of the Technical Specification change request development process will be completed by November 1, 1989. NRC Open Item 255/89007-04:

Consumers Power Company Drawing M-101 Sheet 5113, Revision O, "Piping Isometric, Auxiliary Feedwater Control Valve CV-0736A and CV-0737A Bypass Piping, 11 [Refer to page 13 of NRC Report %-255/89007

(.DRS).] Example 'An additional aspect was associated with the size of socket fillet welds: The inspector noted that the current design practice used by the licensee is sistent with the original Code of construction.

The current practice utilizes MI0789-1683A-TC01-NL02 39 '* ... ; .... "' ,****:::. *:;* .... -,, . ' . "*;. :.-: .. ..;.-.: ,;.* ... -.. . ..... .. .. .... :*"'* '/' --

• later editions of 831.1 Code which specify the lo09 times the nominal p1p1ng wall thickness.

The original Code of construction required 1.25 times the nominal wall thickness. -From a technical standpoint the current practice is acceptable; however, this inconsistency has not been delineated by the licensee in the FSAR. Pending revision of the this item is considered open (255/89007-04).

Reason for Violation Construction codes related to 831.1 have not been reconciled 1n a document useable to the modifications engineer.

Corrective Action Taken and Results Achieved Presentations have been made to all engineering groups on the results of this inspection.

These presentations were completed on August 2, 1989. Corrective Actions to be Taken to Avoid Further Non Compliance Interim Same as that required for Violation l.a. Long-Term Palisades

&taff will complete a reconciliation of all construction codes to the latest edit:,.ion of 831.1. This. action would provide for standardization of code usage-and simplify the determination of code requirements.

This effort will also address the structural welding code AWS Dl.l. Such reconciliation will be documented in plant administrative design control procedures ence 4). In addition, a periodic training program covering procedural welding requirements will be developed.

Upon completion of the reconciliation the FSAR will be updated to* identify applicable codes and standards and their application.

Date When Full Compliance Will be Achieved The personal briefings letter will be issued by September 1, 1989. The ciliation of construction codes will be completed and implemented into plant. design control procedures by January 1, 1990. Training on these procedural revisions will also be complete by January 1, 1990. The periodic training program will be in place by March 1, 1990. The FSAR will be updated in the next revision following January 1, 1990. NRC Unresolved Item 255/89007-06:

SC-89-072 (Deviation Report D-PAL-89-043).

[Refer to page 32 of NRC Report 50-255/89007 (DRS).] MI0789-1683A-TC01-NL02 40 .--' ...

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Example The second concern pertains to the generic aspect of the problem. The licensee appeared to recognize the programmatic weakness which contributed to the problem by revising the RIC form to include the specific weld size. However, there appeared to be no corrective actions directed toward reviewing previously made socket fillet welds for compliance with Code requirements.

Based on the added complication that the sizes of fillet welds in general apparently have not been verified under the licensee's program, reviews of past work may not be sarily limited to socket welded fittings.

Pending a review of the licensee's justification as to why additional inspection of previous fillet welds is not required, this is considered an Unresolved Item (255/89007-06).

CPCo Response CPCo acknowledges that no corrective actions have yet been directed towards reviewing previously made socket fillet welds for compliance with code ments. CPCo plans, however, to select an appropriate sample of as-built welds and inspect the-welds during the 1989 maintenance outage. The sample will be chosen to include a range of weld types. The purpose of the inspection will be to verify that the weld characteristics (type and size) conform to requirements set forth in the Repair Inspection Checklist and/or applicable welding code. These field verifications and resulting report will be completed by December 1, 1989. NRC Unresolved Item 5: Consumers Power Company Drawing M-101 Sheet 5113, Revision O,. "Piping Isometric, Auxiliary Feedwater Control Valve CV-0736A and CV-07-3JA Bypass Piping." [Refer to page 14 of NRC Report 50-255/89007 (DRS).] NRC Identified Discrepancy A further concern associated with the p1p1ng installation drawing pertains to the attachment weld for a bypass piping fitting onto the existing run pipe. For this situation, the drawing did not specify the type of joint nor the weld reinforcement required.

However, the specified fitting is a "Weldolet" and as such has an exisitng weld prep on it and requires no additional design work. Also, the size of the fillet weld cover is specified in the welding procedure for this type of full penetration branch line connection.

The problem arose during the review of the RIC forms for the four branch connection welds. Although these are full penetration single bevel groove welds, with fillet weld reinforcement, the RIC form labels these welds as "F.W." indicating a fillet weld. For Gap Thickness, the RIC form specifies "NA" which would be appropriate for a fillet weld but not for a full penetration weld. Since this attachment must be a full penetration weld, there was no documentation able to assure that the proper penetration has been achieved using the fied fillet weld. Additional review by the inspector of the NDT Examination Reports revealed another deficiency.

According to liquid penetrant (PT) examination report sheet No MKV-01, welds No 2 and No 13 on line 1/2 did not receive a PT examination as required by Specification M-152(Q) "Field Fabrication and Installation of ASME Section Xi Piping fication in a Nuclear Power Plant," Revision 14, September 30, 1986, paragraph MI0789-1683A-TC01-NL02 41 .... ,.. .,.. . .... '* .. -. . --.. *.**-:**.'... . **.* .. -.

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9.1.1. Pending verification that all four branch attachment welds are full penetration welds and resolution of the PT this is considered an Unresolved Item (255/89007-05).

CPCo Response Reference NRC Violation 255/89007-02a.

MI0789-1683A-TC01-NL02 42 . -=--***_µ_*

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. -* ATT0889-0167-NL04 ATTACHMENT 2 Consumers Power Company Palisades Plant Docket 50-255 LIST OF REFERENCES August 10, 1989 1 Page ' ... *.* .. *.* .. ' .: ... :***_-..

References

.. lo Plant Administrative Procedure (AP) 9.11 "Engineering Analyses" --i I 2. AP 9.03 "Facility Change" 3. AP 5.06 "Control of Special Processesn

4. AP 9.06 "Code Requirements for Maintenance and Modifications" 5. AP 9.04 "Specification Changes" 6. AP 9.30 "Q-List" 7. Deviation Report D-PAL-89-43

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