Text

Responds to Violations Noted in Insp Rept 50-255/89-24. Corrective Actions:Comprehensive Program Initiated to Reverify Adequacy of IE Bulletin 79-14 Documentation Re Snubber Reduction Program
	ML18054B548
Person / Time
Site:	Palisades
Issue date:	04/12/1990
From:	Berry K; CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
	IEB-79-14, NUDOCS 9004200535
	Download: ML18054B548 (37)
	v • d • e

r consumers Power POW ERi Nii MICHlliAN'S PROGRESS General Offices: 1945 West_ Parnell Road, Jackson,.Ml 49201 o (517) 788-1636 April 12, 1990 Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 DOCKET 50-255 - LICENSE DPR PALISADES PLANT -

RESPONSE TO INSPECTION REPORT 89024; NOTICE OF VIOLATION Kenneth W Berry Director Nuclear licensing Nuclear Regulatory Commission (NRC) Inspection Report 255/89024, dated January 4, 1990 documented the results of an examination of design activities associated with the Palisades Snubber Reduction Program.

The deficiencies noted during this inspection resulted in issuance of three violations and two unresolved items in the areas of design control and design engineering program implementation.

Consumers Power Company's evaluation of the individual deficiencie.s has shown that they are not safety significant, however, the deficiencies collectively illustrate the need for a continuation of our current management efforts to achieve programmatic improvements, and to communicate quality standards and expectations in the area of design engineering.

The NRC required that this response be submitted within 30 days; however, discussion between members of our respective staffs extended the submittal date to April 12, 1990.

The actions we are taking in response to the items of noncompliance and unresolved_ items identified in the Inspection Report are summarized in Attachment 1 to this letter.

As committed in our January 10, 1990 submittal, we are also responding to the programmatic observations made in Appendix D of _the Office of Nuclear Reactor Regulation (NRR) Audit Team Report (TAC 75155), dated December 15, 1989.

These actions are summarized in Attachment 2 and we are in the process of implementing program improvements that reflect the results of our evaluation of the NRR observations.

Consumers Power Company understands the seriousness of the concerns raised in.the Inspection Report.

As we have previously acknowledged, there was* a time when we experienced a breakdown in controls for our design engineering program: the design program was inadequately implemented and program controls lacked effectiveness.

The effect of this previous lack of design engineering program controls was that errors were not always detected during design package reviews conducted prior to

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Nuclear Regulatory Commission Palisades Nuclear Plant Response to IR 89024 Violation April 12, 1990

. 2 field implementation.

These errors have been corrected, when identified, and _it has been subsequently determined for each case that the error was not safety significant.

We believe that the actions we had already undertaken prior to this inspection had improved design engineering program implementation, and resolved the previously existing program implementation inadequacies.

Beginning in 1986, Consumers Power Company undertook a substantial and comprehensive effort to restore the effectiveness of our design engineering program; The purpose of this effort was to achieve and maintain effective control of Plant design activities.

Among the actions that we have taken to ensure that we obtain this goal, we have established a central design authority, implemented a unified method for effecting facility changes, improved procedures that govern the design engineering process, and communicated the performance expectations held by management to personnel within the Design Engineering organization.

Consumers Power Company believes that these measures have resulted in several programmatic strengths.

These strengths include superior design engineering procedures, improved equipment performance, and competent and knowledgeable personnel.

We are satisfied with the improvements; however, we still recognize the need to remain abreast of changing industry standards, and the need to recognize weaknesses in established programs and make positive efforts when areas for improvement are identified.

The deficiencies identified in Inspection Report 255/89024 are predominantly related to our original IE Bulletin 79-14 (IEB 79-14) verification program.

When our IEB 79-14 program was originally performed, it consisted of field walkdown and reconciliation with the design and analytical basis for large bore piping and supports.

Although the quality of our original IEB 79-14 documentation packages does not satisfy the standards we maintain today, we believe that our original IEB 79-14 verification program efforts *adequately met the definition and intent of the IEB 79-14 program, as it was conceived in 1979, which was to provide a reasonable assurance of safety for nuclear plants by -

confirming the seismic adequacy of safety related piping design and as-built piping configuration

..I.

Nuclear Regulatory Commission Palisades Nuclear Plant Response to IR 89024 Violation April 12, 1990 3

In an effort to establish the expected documentation quality~ Consumers Power Company is currently undertaking a comprehensive program to reverify the adequacy IEB 79-14 documentation, _and is taking aggressive action to enhance the control and quality of its design engineering program in this area.. This IEB 79-14 reverification program, as detailed in our January 10, 1990 submittal, will identify and resolve any documentation or as-built concerns from the original IEB 79-14 effort to the extent necessary.

Kenneth W Berry Director, Nuclear Licensing CC Administrator, Region III, USNRC NRC Resident Inspector - Palisades Attachment

CONSUMERS POWER COMPANY To the best of my knowledge, information and belief, the contents of.this submittal are truthful and complete.

David P Hoffman, Nuclear Operations Sworn and subscribed to before me this 12th day of April, 1990.

Beverly :Ann Avery, Not y Public Jackson County, Michigan My commission expires December 7, 1992

ATTACHMENT 1 Consumers Power Company Palisades.Plant Docket No. 50-255 DETAILED RESPONSE TO ITEMS CITED IN NOTICE OF VIOLATION 50-255/89024 April 12, 1990 19 Pages

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NOTICE OF VIOLATION AND PROPOSED IMPOSITION OF CIVIL PENALTY Consumers Power Company

Palisades Nuclear Generating.Plant Docket No.

50~255 License No. DPR-20 EA 89-251 During an NRC inspection conducted on August 14 through December 7, 1989, violations of NRC requirements were identified.

In accordance with the "General Statement of Policy and_Procedure for NRC Enforcement Actions,"

10 CFR 2, Appendix.C (1989) the Nuclear Regulatory Commission proposes to impose a civil penalty pursuant to Section 234 of the Atomic Energy Act of 1954, as amended (Act), 42 U.S.C. 2282, and 10 CFR 2.205.

The particular violations and associated civil penalty are set forth below:

A.

10 CFR 50, Appendix B, Criterion III, requi~es that measures be established to assure that the design bases are correctly translated into design documents and that the design adequacy is verified and checked.

[255/89024-01)

1.

The design basis for Palisades Design Class I Pipe Supports,

as specified in paragraph 5.10.1.2 of the Updated Safety Analysis Report, requires that the calculated stresses in structural steel components be less than 1.1 times the minimum

yield strength of the support material for the Safe Shutdown Earthquake (SSE) load case.

Contrary to the above, measures were not adequate for modifications performed on 1987 to assure that the design bases was correctly translated into design documents and that the design adequacy was verified and checked, in that the stresses in the structural components identified as Detail "A-A" on Class 1 Pipe Supports No. EB1-S2 and No. EB1-S3 were subsequently calculated in October 1989 and found to be in excess of 1.25 times the minimum yield stress for the SSE load case.

2.

ANSI N45.2.ll, as committed to in the Palisades Quality Assurance Program, CPC-2A, requires design analyses to be documented in sufficient detail to permit auditing and verification of the adequacy of results.

Contrary to the above, engineering analyses performed in 1987 for the Palisades Snubber Reduction Program and in 1988 for Specification Change SC-88138 did not provide sufficient documentation to demonstrate the design adequacy.

No quantitative justification was provided for the acceptability of the following pipe supports which had load increase:

I GC1-Hl37 JB-14-6"-Hl97.4 JB-14-6"-Hl97.5 B.

10 CFR 50, Appendix B, Criterion V, requires that activities affecting quality be prescribed by documented procedures.and be accomplished in accordance with these procedures.

[255/89024-02]

1.

Paragraph 18 of the Palisades procedure, "Criteria for Evaluation of Supporting Structures for Safety-Related Piping Systems," Revision 1, January 17, 1980, requires, in part, that the allowable weld stress be based on the properties of the base metal at temperature.

Contrary to the above, Calculation PSR No.* 20, June 23, 1987, did not consider the base material properties at the operating temperature of 300 degrees F during the review of the adequacy of supports GC1-Hl37 and GCl-Hl.40 for support load increases.

2.

Paragraph 19 of -the Palisades procedure, "Criteria for Evaluation of Supporting Structures* for Safety-Related Piping Systems," Revision l, January 17, 1980, requires in part, that a friction force be included in the support evaluation when the relative displacement between contacting surfaces of pipe and attachment steel is greater than 1/16".

Contrary to the above, Calculation PSR No. 20, June 23, 1987, did not consider the friction force during the review of the adequacy of pipe support GC1-Hl40 even though the calculated relative displacement exceeded 1/4".

-3.

Paragraph 20 of Palisades procedure, "Criteria for Reanalysis of Safety Pipe," Revision 1, November 12, 1979, requires that -

supports utilizing "U" bolts be modeled as two-way restraints.

Contrary to the above, the U-bolt for s~pport No. H-9 at Node 525 of Piping Stress Analysis EA-SC-88138-4, Revision 0, August 1, 1988, was incorrectly modeled as a one-way restraint.

C.

10 CFR 50, Appendix B, Criterion XVI, requires that measures be established to assure that conditions adverse to quality, such as deficiencies and nonconformances are promptly reported and corrected.

Also, in the*case of significant conditions adverse to quality, the measures shall assure that the cause.of the conditions is determined and actions are taken to preclude repetition.

[255/89024-03]

1.

Contrary to.the above,the calculations performed during the IEB 79-14 project for the main steam piping supports EB1-S2Q, Revision 0, August 20, 1981, EB1-S3Q, Revision 0, August 20,

I 1981, and safety Injection Piping Support HC3-Rl33.l, Revision 2, December 3, 1981, concluded that portions of the pipe support assemblies were not in conformance with the applicable design criteria.

The licensee failed to take appropriate corrective action in that the identified co~ditions adverse to quality were not promptly corrected or addressed until 1987.

2.

Contrary to the above, on April 23, 1987, the licensee

.rediscovered the design deficiencies described in C.1 above, for supports EB1-S2Q and EB1-S3Q, as part of the snubber reduction program.

The licensee also identified similar deficiencies for supports EB1-S5Q and EB1-S6Q.

The licensee failed to take appropriate corrective action, in that with the unit in an operating status, the licensee: (a) did not determine the safety significance of the deficiencies with respect to the operability of the associated piping system; (b) did not correct the deficiencies until the unit outage in December 1987; (c) did not initiate a corrective action document until questioned by the NRC in October 1989; and (d) did not take any action to preclude repetition until the need was.identified by NRC inspectors.

Collectively, these violations have been classified as a Severity Level III problem (Supplement 1).

Cumulative Civil Penalty :- $75,000 (assessed equally among the seven violations).

j Response to NRG Violation 255/89024-01:

Example A.l Calculation PSR No. 41, "Main Steam System", Revision 0, April 7, 1987.

[Refer to page 4 of NRG InspectiOn Report 50-255/89-024 (DRS)]

A Consumers Power Company internal memorandum dated April 23, 1987 identified that the original support assessments performed during the IEB 79-14 project for pipe supports EB1-S2, EBl~S3, EB1-S5 and EB1-S6 were incomplete.

The memorandum also noted that the l" diameter U-bolts.associated with these pipe supports needed to be replaced with U-bolts made from a high strength material, and in some cases replaced with larger diameter,U-bolts.

Subsequent review of Bechtel pipe support calculation No. 03341-EB1-S3, Rev 0, Dated November 11, 1980, identified additional discrepancies.

Although the cover sheet states that the pipe support is "structurally adequate", and that no field modification is required, the catalog item evaluation concluded that the structural attachment,.the 20 kip snubbers, and the 1 11 diameter U-bolt, failed to meet the design criteria.

There were no other calculations or evaluations that reconciled these components.

Additionally, there was no structural evaluation of the

.components identified in detail "A-A" which connects the U-bolt to the rest of the pipe support.

The U-bolts for the four pipe supports in question were replaced under Specification Change SC-87-205, which was initiated on December 15, 1987., The Specification Change changed the U-bolt material for the four pipe supports to a higher strength material, and increased the U-bolt diameter from l" to 1-l/4" for pipe supports EB1-S3 and EB1-S6.

These changes were necessary in order to correct the previously identified design deficiencies.

The calculations used to verify the adequacy of the U-bolt changes were documented in EA-SC-87-205-01, Rev 0, dated December 14, 1987, "Replacement of U-Bolts on Main Steam Line Hangers".

However, no documentation could be found that supported an evaluation of the structural components identified in detail "A-A" which had been modified to accept the larger diameter U-bolt.

Subsequently, calculation EA-03341-S3-C/S-l, "Pipe Support Calculation of Restraint EB1-S3",.Rev 0, Dated November 11, 1989, was performed.

This calculation concluded that the structural component in question exceeded the allowable stress criteria by at least 14% for the current design loads.

The FSAR states that the allowable stress for this situation is 1.1 times the yield strength of the material.. The

. calculated strength was approximately 1.25 times the yield strength of the material.

Reason for Violation The deficient conditions described above were first identified to the Plant in the previously cited internal memorandum.

Although these discrepancies had existed since initial installation, they were not

. 1"'*

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discovered until the main steam line U-bolts were reviewed as a part of the Snubber Reduction Program.

In response to the memorandum, an effort was initiated to determine the operability status of pipe support EB1-S3.

The discrepancy described in the internal memorandum was promptly discussed with our pipe support analyst, and it was determined that the as-built hanger configuration was s:tructurally adequate to perform it's designed functiort.

The results of our communications with the analyst also indicated that the as-built configuration of hanger EB1-S3 did not meet the design criteria for material strength of U-bolts, but that the ultimate strength of the hanger would not exceeded, and that some design margin of safety would still be present. *These results were documented in a July 27, 1987 letter.

No.further action was taken to resolve the identified pipe support deficiencies because the personnel involved believed that they }lad been satisfactorily resolved.

During assessment of the U-bolt discrepancies we did not adequately recognize or communicate that the standard for evaluation of pipe support acceptability was the FSAR design criteria, and that any evaluation performed using different criteria needed to have a* documented justification. Additionally, we did not adequately document the evaluation that was performed to determine that the hanger was adequate to perform its design function and maintain piping system integrity.

The engineering analysis that was performed when the U-bolts were

. replaced has also been identified as incomplete since it did not adequately address the effects of enlarging the U-bolt penetrations on detail "A-A" components.

When the U-bolt penetrations on detail "A-A" were enlarged to allow for larger diameter U-bolts, an analysis should have been performed to verify that the components on detail "A-A" were structurally adequate andthat the modified configuration had retained sufficient strength to suppor~ the applied loads. We concur with the NRG evaluation of this deficiency as an example of noncompliance.

Corrective Action Taken and Results Achieved An analysis of the original, as-built hanger configuration was performed utilizing Code Case N-411.

The purpose of t_he N-411 reanalysis was to assess the operability of the pipe support and piping system during the period when the Plant was operated with the

deficient U-bolt conditions uncorrected.

Although the analysis method of Code Case N-411 was not available when the hanger was originally designed, the results of this analysis indicate that the original hanger installation was adequate to satisfy the approved interim operability criteria established in our November 22, 1989 submittal.

Additionally, the pipe supports have now been modified to meet the FSAR design criteria.

Corrective Action Taken to Prevent Recurrence Design engineers have been made aware of the circumstances surrounding this condition, and administrative control of design engineering activities has been strengthened through procedural

improvements.

Administrative Procedure 9.11, Rev. 4, "Engineering Analysis" contains several recently addedrequirements and clarifications: 1) attention to detail is emphasized, 2) the bases for design input assumptions-will be clearly stated-and documented,

3) dimensional information used in design evaluations will be verified by field walkdown, 4) and the bases used to justify engineering assumptions will be substantiable.

A technical review checklist has also been added to this procedure to provide further assurance that these activities are performed during the technical review.

Date When Full Compl:iance Will be Achieved Full compliance with design criteria has been achieved for pipe supports EB1-S2 and EB1-S3.

Example A.2. Support GC1-Hl37:

Calculation PSR No. 20, "Safety Injection, Containment Spray, and Shutdown Cooling System", Revision 0, June 23, 1987; Support No. GC1-Hl37 (Q).

[Refer to page 6 of NRC Inspection Report 50~255/89-024 (DRS)]

The design loads for pipe support GC1-Hl37 (Q) increased by approximately 30% due to Snubber ~eduction Program modifications.

However, no quantitative basis documenting the basis for determining continued acceptability of the pipe support was provided, as specified in ANSI N45.2.ll, and the Palisades Quality Assurance Program.

Reason for Violation Instead of performing a detailed reanalysis of each affected pipe support during the Snubber Reduction Program, a scaling method was typically used.

The results of these scaling method based pipe support evaluations were documented in the text of the design change package, and were based on the assumption that original pipe support calculations were correct.

We concur with the NRG evaluation of this*

deficiency as an example of noncompliance.

Corrective Action Taken and Results Achieved A sample verification was performed of pipe supports that had significant load increases due to the Snubber Reduction *Program.

The results of calcul.ations performed during this verification effort indicate that none of the assumptions made in the scaling based'pipe

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support evaluations were compromised, and that scaling of loads is a reasonable approach for evaluating increases in pipe support loading.

An evaluation was also performed that demonstrates that pipe support GC1-Hl37 was marginal, but fully adequate to meet design criteria in both the original as-built configuration, and the configuration that resulted from piping system modifications performed under the Snubber Reduction Program.

We have recently reduced pipe support loads by over BOX for pipe support GC1-H137 by removing adjacent pipe support GC1-Hl36.2. *As a result, the factor of safety for GC1-Hl37 has been greatly increased, and the loads on GC1-Hl37 are no longer of a magnitude that makes its acceptability marginal.

Removal of pipe support GC1-Hl36.2 had only a negligible effect on the loads experienced by other pipe supports in the piping system; however, it greatly reduced GC1-Hl37 loads.

An update of the stress calculation for support GC1-Hl37 is currently in progress which will formalize the record file for pipe support GC1-Hl37, and to reflect removal of pipe support GC1-Hl36.2.

Completion of the record file update is expected by July 1, 1990.

Corrective Action Taken to Prevent Recurrence Administrative control of design engineering activities has been strengthened through procedural improvements.

We have provided

.additional guidance on the importance of documenting the bases employed to justify cases where engineering judgment is used in Administrative Procedure 9.11, Rev. 4, "Engineering Analysis".

Administrative Procedure 9.11, also reflects several other new requirements and clarifications: 1) attention to detail will be emphasized, 2) the basis for design input assumptions will be clearly stated and docuIIiented, 3) dimensional information used in design evaluations will be verified by field walkdowns, 4) and the bases used to justify engineering assumptions will be substantiable.

A technical review checklist has also bee.it added to this procedure to provide further assurance that these activities are performed during the technical review.

Although the previously mentioned sample verification has demonstrated the validity of scaling as a method for evaluating pipe support load increases, in the future we will routinely perform pipe

~upport calculations to evaluate the effects of load increases rather than-using the scaling method.

Procedural guidance will be provided to design engineers to clarify when it is acceptable to scale existing calculations instead of performing a calculation.

This procedure is scheduled to be in place by September l, 1990.

Date When Full Compliance Will be Achieved Full compliance with design criteria has been continuously maintained for pipe support GC1-Hl37.

. ~

Example A.2. Pipe Supports JB-14-6"-Hl97.4 and JB-14-6"-Hl97.5:

Steam Supply for Auxiliary Feedwater Pump Turbine Driver K-8, Piping Stress Calculation No. EA-SC-88138, Revision 0, August 1, 1988.

[Refer to page 15 and 16 of NRC Inspection Report 50-255/89-024 (DRS)]

During review of Piping Stress Calculation No. EA-SC-88138 it was determined that design loads for pipe support Hl97~5 had increased by a factor of 2.53 in the vertical direction following implementation of a modification.

The calculation also stated that the two-sided 3/16" weld on the 2x2 tube steel was inadequate and needed to be increased to 3/16" all around.

No analysis was provided to justify this change.

Based upon preliminary calculations, it was believed that a 3/16" weld all around would still result in stresses that exceed the FSAR stress limits by approximately 10%.

Additionally, the post-modification design loads for pipe support Hl97.4 were increased by approximately 70%.

However, there was no quantitative justification to demonstrate the adequacy of either of the pipe support designs.

Reason for Violation The engineering judgment based evaluation that was noted in the package for calculation EA-SC-88138 was performed in order to determine the acceptability of a potential overstress condition for pipe supports Hl97.4 and Hl97.5.

The conclusion reached by this evaluation has been shown to be adequate and correct, despite the lack of a documented, quantitative evaluation.

We acknowledge that a basis for justifying the evaluation performed for calculation EA-SC-88138 was not documented sufficiently to allow substantiation, and that this does not meet NRC expectations.

We. concur with the NR,C.

evaluation of this deficiency as an example of noncompliance.

Corrective Action Taken and Results Achieved An operability determination has been performed for pipe supports Hl97.5 and Hl97.4.

This evaluation considered only the comments and deficiencies which affect key stress parameters and demonstrate a significant impact on pipe support evaluation results.

The results of this preliminary evaluation indicate that pipe supports Hl97.5 and Hl97.4 are acceptable for use in the present configuration, without modification An update of the entire stress package (EA-ESSR-90708-02) for pipe supports Hl97.5 and Hl97.4 is currently in final review.

This update will include complete evaluation of comments and evaluation of updates required by modifications, and is expected to be complete by April 30, 1990.

Corrective Action Taken to Prevent Recurrence Administrative control of design engineering activities has been strengthened through procedural improvements.

We have provided

).

additional guidance on the importance of documenting the bases employed for justifying cases where engineering judgment is used in Administrative Procedure 9.11, Rev. 4, "Engineering Analysis".

Administrative Procedure 9.11 also reflects several other new requirements and clarifications: 1) attention to detail will be emphasized, 2) the basis for design input assumptions* will be clearly stated and documented, 3) dimensional information used in design evaluations will be verified by field walkdowns, 4) and the bases used to justify engineering assumptions will be substantiable.

A technical revie~ checklist has also been added to this procedure to provide further assurance that these activities are performed during the technical review.

As an additional corrective action, Specification M-195, "Requirements for the Design and Analysis of Palisades Plant Safety Reiated Piping and Instrument Tubing" is being revised to incorporate recent experience.

Revision of Specification M-195 is expected to be complete by April 30, 1990.

Date When Full Compliance Will be Achieved Full compliance with FSAR design criteria has been continuously maintained for pipe supports JB-14-6"-Hl97.5 and JB-14-6"-Hl97.5.

NRG Violation 255/89024-02:

Example B.l. Support GC1-H137(Q):

Pipe Support Design Review Pertaining to Procedures and Drawings, Safety Injection System, Support No. GC1-Hl37 (Q).

[Refer to pages 7 and 8 of NRG Inspection Report 50-255/89-024 (DRS))

During review of the design for pipe support GC1-Hl37 (Q), it was determined that the allowable weld stresses in the pipe support evaluation were incorrectly assumed for both the Operating Basis Earthquake (OBE) and Safe Shutdown Earthquake (SSE) because carbon steel at ambient conditions was assumed for the pipe support material.

The actual pipe support installation was stainless steel at an operating temperature of 300 degrees F.

Paragraph 18 of the "Criteria for Evaluation of Supporting Structures for Safety-Related Piping Systems" Rev. 1, January 17, 1980, states that the properties of the base material will be considered at operating temperature.. Neither the correct operating temperature, nor the correct base material.were used for evaluation of pipe support GC1-Hl37 (Q).

Reason for Violation The condition described in the example above resulted when incorrect operating temperature and base material data from the original pipe

support documentation pack&ge was used to evaluate the effect of pipe support load increases at a later date.

The pipe support documentation package was generated during the original IEB 79-14 program. and the presence of incorrect operating temperature and base material data is due to inadequate implementation of procedures used during the original IEB 79-14 project.

The error was compolinded when the errors in the support load increase evaluation were not detected by the technical review process.

We concur with the NRG evaluation of this deficiency as an example of noncompliance.

Corrective Action Taken and Results Achieved An evaluation was performed that demonstrated that pipe support GC1-Hl37 was marginal, but adequate to meet design criteria in.both the original, as-built configuration; and the configuration that resulted from Snubber Reduction Program piping system modifications.

The deficiencies identified in. this. example would not have substantially altered this conclusion.

We have recently reduced pipe support loads by over SOX for pipe support GC1-Hl37 by removing adjacent pipe support GC1-Hl36.2.

As a result, the factor of safety for GCl-Hl37 has been greatly increased, and the loads on GC1-Hl37 are no longer of a magnitude that makes its acceptability marginal.

Removal of pipe support GC1-Hl36.2 had only a negligible effect on the loads experienced by other pipe.supports in the piping system; however, it greatly reduced GC1-Hl37 loads.

An update of the stress calculation for support GC1-Hl37 'is currently in progress which will formalize the record file for pipe support GC1-Hl37.

Completion of the record file update is expected by July 1,.1990.

Corrective Action Taken to Prevent Recurrence We are currently in the process of implementing a program to reverify the original IEB 79-14 results.

As a part of this program, we will review our present as-built pipe support configuration against design documents and existing analyses in order to identify and reconcile anomalous conditions such as the operating temperature and base material discrepancies identified in this example.

Specification C-173, "Technical Requirements for the Analysis of Safety Related Pipe Supports". is being revised, and will conservatively evaluate material and temperature effects during pipe support analysis.

This Specification will be used during the IEB 79-14 reverification program, as well as in the development of future pipe support modifications.

Incorporation of these additions to Specification C-173 is expected to be complete by April 30, 1990.

As an additional corrective action, design engineers have been made aware of the circumstances surrounding this condition, and have been notified that data contained in the pipe support documentation packages is not to be used to for design evaluation purposes unless

1../

it has been field verified. Administrative control of design engineering activities has also be.en strengthened. Administrative Procedure 9.11, Rev. 4, *"Engineering Analysis" contains several recently added requirements and clarifications: 1) dimensional information used in design evaluations will be verified by field

-walkdown, 2) attention to detail is emphasized, 3) the bases for design input assumptions will be clearly stated and documented, 4) and the bases used to justify engineering assumptions will be substantiable.

A technical review checklist has also been added to this procedure to provide further assurance that these activities are performed during the technical review.

Date When Full Compliance Will be Achieved Full compliance with FSAR design criteria has been continuously maintained for pipe support GCl-Hl37.

Example B.1. Support GC1-Hl40(Q):

Pipe Support Design Review Pertaining to Procedures and Drawings, Safety Injection System, Support No. GC1-Hl40 (Q).

[Refer to pages 7 and 8 of NRG Inspection Report 50-255/89-024 (DRS)]

Paragraph 18 of the "Criteria for Evaluation of supporting Structures for Safety-Related Pip~ng Systems" Rev.. 1, January 17, 1980, states that the properties of the base material will be considered at operating temperature.

Neither the correct operating temperature, nor_the correct base material were used for evaluation of pipe support GCl-Hl40 (Q).

Although carbon steel at ambient conditions was assumed for the pipe support material, the actual pipe support installation-was stainless steel at an operating temperature of 300 degrees F.

Reason for Violation The condition described in the example above resulted when incorrect operating temperature and base ~aterial data from the original pipe support documentation package was used ~o evaluate the effect of pipe support load increases at a later date.

The pipe support documentation package was generated during the original IEB 79-14 program, and the presence of incorrect operating temperature and base material data is due to inadequate implementation of procedures used during the original IEB 79-14 project.

The error was compounded when.

the errors in the support load increase evaluation were not detected by the technical review process.

We concur with the_NRC evaluation of this deficiency as an exampie of noncompliance.

Corrective Action Taken and Results Achieved A calculation has been performed by Impell (GC1-Hl40, Rev. 1, November 14, 1989) to evaluate the as-built condition of pipe support

'(/

GC1-Hl40.

The results of this evaluation indicate that the current pipe support installation is adequate to meet FSAR design criteria, and th~t no modifications are needed.

Corrective Action Taken to Prevent Recurrence We are currently in the process of implementing a program to reverify the original IEB 79-14 results.

As* a part of this program, we will review our present as-built pipe support configuration against design documents and existing analyses in order to identify and reconcile anomalous conditions such as th~ operating temperature and base material discrepancies identified in this example.

Specification C-173, "Technical Requirements for the Analysis of Safety Related Pipe Supports", is being revised, and will conservatively evaluate material and temperature effects during pipe support analysis.

This Specification will be used during the IEB 79-14 reverification program, as well as in the development of future pipe support modifications.

Incorporation of these additions to Specification C-173 is expected to be complete by April 30, 1990.

As an additional corrective action, design engineers have been made aware of the circumstances surrounding this condition, and have been notified that data contained in the pipe support documentation packages is not to be used to for design evaluation purposes unless it has been field verified.

Administrative control of design engineering activities has also been strengthened.

Administrative

,Procedure 9.11, Rev. 4, "Engineering Analysis" contains several recently added requirements and clarifications: l) dimensional

_information used in design evaluations will be verified by field walkdown, 2) attention to detail is emphasized, 3) the bases for design input assumptions will be clearly stated and documented, 4) and the bases used to justify engine,ering assumptions will be substantiable.

A technical review checklist has also been added to this procedure to provide further assurance that these activities are performed during the technical review.

Date When Full Compliance Will be Achieved Full compliance with FSAR design criteria has been continuously maintained for pipe support GC1-Hl40.

Example B.2. Support GC1-Hl40(0):

Pipe Support Design Review Pertaining to Procedures and Drawings, Safety Injection System, Support No. GC1-Hl40 (Q).

[Refer to pages 7 and 8 of NRC Inspection Report 50-255/89-024 (DRS)]

Paragraph 19 of "Criteria for Evaluation of Supporting Structures for Safety-Related Piping Systems" Rev. l, January 17, 1980, states that a friction force shall be included in the support evaluation when the relative displacement between contacting surfaces of pipe attachment

.. *~

steel and structural attachment steel is great;er than 1/16".

Although the thermal movement was greater than 1/4" for pipe support GC1-Hl40 (Q), friction force was not considered in the pipe support evaluation.

Reason for Violation The condition described in the example above resulted when incorrect pipe support evaluation results contained in the original pipe support documentation package were used as a basis to re-evaluate pipe support GC1-Hl40 in 1987.. The pipe support documentation package was generated during the original IEB 79-14 program, and did not include an evaluation of the effects of friction loads.

Absence of a friction force evaluation in the pipe support documentation package was due to inadequate implementation of procedures during the IEB 79-14 program.

The documentation package error was compounded when the omitted friction load evaluation.was not detected during technical review of the pipe support re-evaluation.

We concur with the NRC evaluation of this deficiency as an example of noncompliance.

Corrective Action Taken and Results Achieved A calculation has been performed by Impell (GC1-Hl40, Rev. l, November 14, 1989) to evaluate the as-built condition of pipe support GC1-Hl40.

The results of this evaluation indicate that the current pipe support installation is adequate to meet FSAR design criteria, and that no modificat~ons are needed.

Corrective Action Taken to Prevent Recurrence We are currently in the process of implementing a program to reverify the original IEB 79-14 results.

As a part of this program, we will review our present as-built pipe support configuration against design documents and existing analyses in. order to identify and reconcile anomalous conditions such as the friction load discrepancy identified in this example.

Specification C-173, "Technical.Requirements for the Analysis of Safety Related Pipe Supports", is being revised, and will conservatively evaluate friction effects during pipe support analysis.

This Specification will be used during the IEB.79-14 reverification program, as well as in the development of future pipe support modifications.

Incorporation of these additions-to Specification C-173 is expected to be complete by April 30, 1990.

As.an additional corrective action, design engineers have been made aware of the circumstances. surrounding this condition, and have been notified that data contained in the pipe support documentation packages is not to be used to for design evaluation purposes unless it has been field verified. Administrative control of design engineering activities has also been strengthened.

Administrative Procedure 9.11, Rev. 4, "Engineering Analysis" contains several recently added requirements and clarifications: l) dimensional

information used in design evaluations will be verified by field walkdown, 2) attention to detail is emphasized, 3) the bases for design input assumptions will be clearly stated and documented, 4) and the bases used to justify engineering assumptions will be substantiable.

A technical review checklist has also been added to this procedure to provide further assurance that these activities are performed during the technical review.

Date When Full Compliance Will be Achieved Full compliance with FSAR design criteria has been continuously maintained for pipe support GCl-Hl40.

Example B.3:

Steam Supply for Auxiliary Feedwater Pump Turbine Driver K-8, Piping Stress Analysis, Calculation EA-SC-88138-4, Revision 0, August l, 1988.

[Refer to page 14 of NRC Inspection Report 50-255/89-024 (DRS)]

Example During review of piping stress calculation EA-SC-88138-4, Revision 0, it was discovered that Node 525 was modeled a a Y restraint, despite the fact that it is a U-bolt connection.

According to the Bechtel "Criteria for Reanalysis of Safety Related Piping", this restraint should be modeled as a Y-Z restraint in order to avoid thermal expansion difficulties.

Reason for Violation The discrepant condition described in the example above resulted from errors in the original piping analysis model.

-The incorrect piping analysis model.was relied upon during preparation of an Auxiliary Feedwater modification package.

The errors in the original piping system model resulted from inadequate control of technical and programmatic work activities during the original IEB_ 79-14 project.

We concur with the NRC evaluation of this defici~ncy as an example of noncompliance.

- Corrective Action Taken and Results Achieved The effects of the U-bolt modeling discrepancy described above have undergone preliminary evaluation.

Based on the results of this evaluation, it has been determined that the as-built piping and pipe support configuration are acceptable and do not require modification.

Engineering evaluation EA-ESSR-90708-02 is currently being prepared as a revision to pipe stress analysis package 03358C.

This evaluation will finalize the* incorporation of audit and inspection related comments, and is expected to be complete by April 30, 1990.

Corrective Action Taken to Prevent Recurrence

. We have notified design engineering personnel that the information contained in our original IEB 79-14 design files must be field verified if it is to be used as design input information.

Additionally, we have provided additional guidance on the importance of field verifying dimensional information that is used to perform design evaluations in Administrative Procedure 9.11, Rev. 4, "Engineering Analysis".

We are also in the process of revising the specification used for pipe stress analysis.

Specification M-195, "Requirements for the Design and Analysis of Palisades Plant Safety Related Piping and Instrument Tubing", is currently in the process of being revised in order to increase usability and include recently learned lessons.

Changes to this Specification are expected to be complete by April 30, 1990.

Date When Full Compliance Will be Achieved Full compliance with FSAR design criteria has been continuously maintained for pipe support GC1-Hl40.

NRC Violation 255/89024-03 Example C.l:

Pipe Support Design Review Pertaining To Corrective Action, Main Steam System, Support No.s EB1-S2.and EB1-S3.

[Refer to page 9 of NRC Inspection Report 50-255/89-024 (DRS)]

In August 1981, a Bechtel design review concluded that the structural attachments, snubbers, and U-bolts for main steam pipe supports EB1-S2 and EB1-S3 would fail under the specified seismic loads.

Documentation was not issued at this time to ensure that the identified items would be promptly corrected, and corrective actions were not taken to replace the inadequate U-bolts until December 1987.

A similar condition was also identified in August 1981 involving Safety Injection System pipe support HC3-Rl33.l, during review of calculation 03319-24, November 1979.

This calculation was performed to evaluate the as-built configuration of an IEB 79-14 modification.

The results of this calculation indicated that hardware modifications were required to increase the size of the anchor plate and expansion anchor bolts; however, the modification had not been implemented.

Example C.2:

Pipe Support Design Review Pertaining To Corrective Action, Main Steam System, Support No.s EB1-S2 and EB1-S3.

[Refer to page 9 of NRC Inspection Report 50-255/89-024 (DRS)]

A corrective action document was not issued until 1989 to determine the root cause of deficient conditions identified on August 20, 1981 and April 23, 1987 for pipe supports EB1-S2Q and EB1-S3Q, or for deficiencies identified on April 23, 1987 for pipe supports EB1-S5Q and EB1-S6Q.

Additionally, following identification of these

deficiencies on April 23, 1987, the safety significance of the deficiencies with respect to piping system operability was not evaluated, corrective actions were not taken until the unit outage in December 1987, and actions were not taken to prevent repetition.

Reason for Violation The original calculations for pipe supports EBl~s2 and EB1-S3 were performed by Bechtel in November 1980, and indicated that the pipe support components would fail under seismic loading.

Approximately one year later, in November 1981, Bechtel approved this same calculation, without making any changes, by noting that the calculation results were acceptable on the calculation cover sheet.

No doc'1mentation was provided to justify the engineering basis employed to reach this conclusion.

The Plant was first notified that an apparently failed pipe support calculation package had been approved without documented justification in an internal memorandum dated April 23, 1987.

In order to resolve questions we had related to operability of the affected pipe supports we promptly contacted the analyst and indicated that we needed to know if the pipe supports were still capable of performing their design function.

The results of this discussion indicated that although the U-bolts did not meet the design criteria for material strength, the hanger was acceptable for use since it would not be stressed beyond its ultimate strength, and it still had some factor of safety remaining to assure that it could perform its safety function.

This discussion was documented in a letter to the analyst dated July 27, 1987.

The Plant personnel involved in the discussions did not recognize at the time the operability determination was performed that the evaluation did not resolve the deficient condition described in the April 23, 1987 letter.

This oversight occurred because the personnel involved did not realize that an evaluation that utilized criteria other than.the pipe support design criteria described in the FSAR still constituted a nonconforming condition. It was.assumed by the personnel involved that the nonconforming condition had* been resolved when it was determined that the pipe supports were adequate to perform their safety function.

A corrective action document was not initiated in 1987 when the pipe support deficiencies were identified during the Snubber Reduction Program due to an incorrect perception held by the personnel who evaluated the pipe support deficiencies that the nonconforming condition involving pipe supports EB1-S2, EB1-S3, EB1-S5, an-d EB1-S6 had been resolved when the pipe supports were found to be structurally adequate.

Failure to initiate a corrective action document for the pipe support deficiencies prevented management awar~ness of the deficient conditions, and delayed corrective action program evaluation of the pipe support deficiencies.

The pipe

support deficiencies identified in the April 23, 1987 letter should have been addressed at the time of discovery by a corrective action document.

We concur with the NRC evaluation of these deficiencies as examples of noncompliance.

A similar example involving Safety Injection System pipe support HC3-Rl33.l was identified during review of calculation 03319-24, November 1979.

The results of this calculation indicated that hardware modifications were required to increase the size of the anchor plate and expansion anchor bolts.

However, this modification had not yet been implemented at the time of the inspection.

A corrective action document was not initiated until 1989, and corrective actions were not taken promptly for this item because management did not identify it as a deficient condition.

We concur with the NRC evaluation of these deficiencies as examples of noncompliance.

Corrective Action Taken and Results Achieved for C.l and C.2 A modification has been implemented to replace the U-bolts for pipe supports EB1-S2, EB1-S3, EB1-S5, and EB1-S6, and a calculation has been performed to assess the pre-modification structural adequacy of the piping system associated with these pipe supports.

The pipe support evaluation for pipe supports EB1-S2, EB1-S3, EB1-S5, and EB1-S6 were performed using approved interim operability criteria in order to determine whether the pipe support installation was adequate prior to modification.

The evaluation results indicate that the pipe supports were acceptable in the as-built condition, prior to modification.

A calculation has been performed by Impell (HC3-R133.l, Rev. 1, November 13, 1989) in order to determine the adequacy of pipe support HC3-R.133.1.

The results of this evaluation indicate that the pipe support is adequate in the as-built condition.

Based on the results of this evaluation, the modification package that was initiated to increase the size of the anchor plate and expansion anchor bolts is no longer considered necessary.

Corrective action documents have been initiated for the deficient conditions associated with the Main Steam System and Safety Injection System pipe supports deficiencies in 1989.

Corrective Action Taken to Prevent Recurrence for C.l and C.2 Management has placed increased emphasis on defining what constitutes a nonconforming item and documenting nonconforming conditions.

As a part of this effort, Design Engineers and System Engineers have recently attended seminars related to the corrective actions which resulted from design engineering related deficiencies identified within the original IEB 79-14 program.

Among the topics covered at

- these seminars were, the.need to correctly perform operability determinations and the need to initiate corrective action documentation when nonconforming conditions are identified.

Administrative control of design engineering activities has been strengthened through procedural* improvements.

We have provided additional guidance on the importance of documenting the bases employed to justify cases where engineering judgment is used in Administrative Procedure 9.11, Rev. 4, "Engineering Analysis".

Date When Full Compliance Will be Achieved Full compliance with FSAR design criteria has been achieved for pipe supports HC3-Rl33.l, EB1-S2, EB1-S3, EB1-S5, and EB1-S6.

Unresolved Item 255/89024-04:

Example: Pipe Welded Attachments Local Stress Evaluation.

[Refer to pages 9 and 10 of NRG Inspection Report 50-255/89-024 (DRS)]

During review of documents associated with the Snubber Reduction Program, it was noted that the localized stresses induced into the

piping component by welded attachments such as lugs and. cylindrical attachments were not evaluated to ensure that the combined stresses on the attachments,*weld configurations, and pipe were within code allowable limits. It was also noted that FSAR Table.5.2-3 cites ASA B31.l (1955) as the piping code used; however, FSAR Section 5.10, "Systems and Components", states that the piping system was designed to USAS B31.l (1967).

'J1le Bechtel pipe support design criteria for Palisades, dated January 1980, also refers to USAS B31.l (1967).

The 1967 edition of USAS B31.l has a paragraph which states that integral (welded) attachments*

are an integral part of the piping component, and that consideration should be given to localized stresses induced into the piping component by the integral (welded) attachments.

Additionally, the Bechtel Pipe Support Manual (PSM), Section 3.11.1, dated April 21, 1980, establishes the following guidelines: 1) welded attachments on process pipe should be avoided where possible, and 2) where an attachment is welded to a piping system as a part of a pipe support, local stresses will be induced in the pipe wall around the area of the attachment.

Pipe stresses obtained in an analysis without consideration given to the local stresses are referred to as general pipe stresses and [will] be checked against the allowable stresses.

Reason for Unresolved Item Code requirements in place at the time the original IEB 79-14 work was being performed we~e vague and inconclusive regarding the analysis of localized stresses, and calculations were not generally

performed when considering localized stresses since they were not explicitly required.

Non-calculational methods of evaluating the effects of localized stresses, such as the Kellogg method, on pipe support and piping system integrity were typically used at the time when our original IEB 79-14 calculations were being performed.

Response

It is uncertain how the effects of localized stresses were addressed in the original IEB 79-14 program.

However, we will evaluate the effects of local stresses induced by integral welde~ attachments in our IEB 79-14 reverification program.

This evaluation will examine a sample of ten integral welded attachments in order to determine if a generic concern exists.

The evaluations will be performed in accordance with current industry practice under Specification M-195, "Requirements for the Design and Analysis of Palisades Plant Safety Related Piping and Instrument Tubing".

This Specification is currently being revised to include integral welded attachments in the evaluation and design of nuclear piping system supports, and will be used *during future pipe support analyses.

Incorporation of the locaiized stres_s revisions into Specification M.,-195 is expected to be complete by April 30, 1990.

Unresolved Item 255/89024-05 Example: Steam Supply for Auxiliary Feedwater Pump Turbine Driver K-8, Field Verification.

[Refer to page 16 of NRC Inspection Report 50-255/89-024 (DRS)]

During a system ~alkdown of general system configuration, it was noted that line HB-4-1" appeared to be inadequately supported.

This small bore steam trap line is attached* to steam supply pipe EB-13-4" just upstream of the Auxiliary Feedwater Turbine.

Response

Line HB-4-1" had been decoupled from the main pipe run during development of stress package 03358 and was riever reincorporated into a seismic analysis package.

A pipe stress calculation has been performed for this pipe run, and additional supports have been installed under SC-89-338.

The system installation is now in full accordance with FSAR design criteria. Criteria for decoupling of piping systems during stress analysis are provided in Specification M-195, "Requirements for the Design and Analysis of Palisades Plant Safety Related Piping and Instrument Tubing".

Implementation of the guidance provided in this Specification will prevent deficiencies such as the one described in this example from occurring in the future.

--*-----~

ATTACHMENT 2 Consumers Power Company Palisades Plant Docket No. 50-255 RESPONSE TO APPENDIX D OBSERVATIONS FROM ~HE NRR TEAM AUDIT REPORT (TAC 75155)

ON PIPE HANGERS AND SUPPORTS

. April 12, 1990.

12 Pages

Restatement of Appendix D Observation 1:

[Ref. NRR TAC Report 75155; Appendix D, page 2, and Appendix C, pages 1 and 2)

The original !EB 79-14 stress packages contain numerous deficiencies and cannot be used to demonstrate the adequacy of the safety-related

,piping and pipe supports at Palisades.

Response

The presence of deficiencies in our original IEB 79-14 stress packages can be attributed to inadequate management overview of vendor supplied design engineering services during the original IEB 79-14 project.

We did not recognize that the condition of our piping and pipe support stress packages warranted increased attention until recently because numerous previous reviews of our original !EB 79-14 program had not identified the number of package errors as a concern, and the package errors that were identified had not been safety

significant.

In recognition of the potential significance that stress package -deficiencies can have, we have committed to perform a reverification of our IEB 79-14 program stress packages.

We believe that the IEB 79-14 reverification program will establish that the errors contained in our original IEB 79-14 stress packages did not result in a safety significant condition.

The stress package deficiencies identified to_ date have affirmed this position.

We have notified design engineering personnel that the information contained in our original !EB 79-14 design files must be field verified if it is to be used as design input information.

Further, the engineers have been informed that design input information that is obtained from the original !EB 79-14 files must either be verified through field observation, in accordance with recent changes to Administrative Procedure 9.11, "Engineering Analysis"; or reanalyzed to confirm accurate representation of the as-built pipe support configuration.

The quality of information contained in our pipe*

support design files will be upgraded during the IEB 79-14 reverification program, as committed to the NRC in our letter dated January 10, 1990.

Restatement of Appendix D Observation 2 Operability criteria should be _prepared to permit evaluation for continued interim operation even though systems may be outside FSAR limits.

Response

It is acknowledged that some of our pipe supports may not meet FSAR design criteria; however, it is also recognized that the codes used to satisfy the FSAR design criteria for piping and pipe supports,

employ substantial factors of safety.

As a result, it is possible to

. -~'..

~** ****---

develop and implement a set of operability criteria that will allow continued safe operation of the plant on an interim basis in cases where pipe supports that do not meet FSAR design criteria, but that are otherwise structurally adequate, are identified.

By letter dated November 13, 1989, Consumers Power Company submitted a set of proposed interim operating criteria to the NRC for review and approval.

As a result of discussi6ns with the NRC staff, clarification and revision of the interim operability criteria was provided via a Consumers Power Company letter to the NRC dated November 22, 1989.

We have received verbal concurrence that the interim operability criteria, as clarified in our letter dated November 22, 1989, are.acceptable.

These interim operating criteria will be applied when conditions exceeding the FSAR design criteria are identified during the IEB 79-14 reverification effort in order to determine if the condition is allowable.

Restatement of Appendix D Observation 3 The use.of vertical ground motion spectra for all elevations is specified in Specification M-195.

This is not in conformance with current techniques and therefore does not satisfy the requirements of R. G. 1.84 when implementing Code Case N-411 damping.

Justification

for this apparent lack of conformance must be provided.

Response

Consumers Power Company does not concur with the NRC assessment of this issue.

In our letter to the NRC dated July 28, 1986, we outlined how we intended to meet the requirements of R. G. 1.84.

This correspondence identified that we would use the response spectra*

outlined in NUREG/CR 1833, which is abounding vertical spectra.

NUREG/CR 1833 states that amplification of the vertical spectra for both the Auxiliary Building and Containment structure are negligible.

Therefore, the use of a single bounding spectra is both simplistic and accurate.

Changes to Specification M-195, "Requirements for the Design and Analysis of Palisades Plant Safety Related Piping and Instrument Tubing" will highlight our use of a single response spectra.

Changes to this specification are expected to be complete by April 30, 1990.

Restatement of Appendix D Observation 4 Specification C-173(Q) used for the design of pipe supports has the following anomalies which require justification:

a)

FSAR allowables are increased by a factor of 1.1.

b)

Load rated capacity of components is increased by 80~ for SSE

loadings.

c)

The interaction equation used for anchor bolts is not linear.

d)

Modifications for existing pipe supports permit use of new allowable stresses.

Response

We are revising specification C-173, "Technical Requirements for the Analysis of Safety Related Pipe Supports", to be more specific with respect to implementing FSAR design requirements.

The revised Specification will be used during the IEB 79-14 reverification project, and is expected to be complete by April 30, 1990.

Our re~ponse to the apparent anomalies is as follows:

a)

Specification* C-173, "Technical Requirements for the Analysis of Safety Related Pipe Supports", reflects the FSAR allowables.

These allowables are 1.1 times the AISC allowables, and not 1.1 times the FSAR allowables.

b)

We believe that the.SOX load increase is consistent with industry practice, as reflected in MSS SP-58, ANSI B31.l, and ASME Section III, Subsection NF.

c)

We acknowledge that the interaction equation used for evaluating anchor bolts is not linear; however, it is in accordance with industry accepted criteria for the evaluation of existing anchor bolts.

d)

All of our calculated pipe stresses meet FSAR criteria, regardless of the method used to perform the calculations.

Restatement* of Appendix D Observation 5 Specifications providing design criteria and procedures for pipe stress analysis should be prepared.

Response

We are in the process of expanding the specification used for pipe stress analysis.

Specification M-i95, "Requirements for the Design and Analysis of Palisades Plant Safety Related Piping and Instrument Tubing", is currently in the process of being revised in order to increase usability and include recently learned lessons.

Changes to this specification are expected to be complete by April 30, 1990.

Restatement of Appendix D Observation* 6 The mixed use of various codes and standards, and different editions of each, seems to occur without any justification: As an example, a stress intensification factor is used from one edition of a standard, while the allowable stress for the same problem is taken from another

edition. Justification for the mixed use of codes and standards must be provided.

Response

A reconciliation of the Codes that are applicable for use when performing piping system and pipe support stress analyses under Specifications C-173, "Technical Requirements for the Analysis of Safety Related Pipe Supports", and M-195, "Requirements for the Design and Analysis of Palisades Plant Safety Related Piping and Instrument Tubing", has been completed.

The documents which will incorporate the results of this Code reconciliation are expected to be complete by December 31, 1990.

Restatement of Appendix D Observation 7 Several apparent QA deficiencies were noted which indicate that the Palisades QA Program and.its implementation should be reviewed:

a)* The main steam U-bolt overstress problem was identified in 1987; however no nonconformance report, or equivalent, was written.

b)

Calculations were prepared by a supervisor, checked by an engineer who worked for the supervisor, and approved by the supervisor who prepared the calculation.

c)

The cover sheet of a calculation is signed off as structurally adequate when in fact the calculation showed numerous failures.*

Response

a)

The Plant personnel who evaluated the main steam system pipe support deficiencies did not recognize that the evaluation they had performed to address structural adequacy of the pipe supports did not resolve the nonconforming condition described in the April 23, 1987 letter.

This oversight occurred because the personµel involved did not realize that, regardless of an evaluation, a pipe support configuration that c;loes not meet the FSAR design criteria is a nonconforming condition. It had been assumed by the personnel involved in evaluation of the pipe support deficiencies that the nonconforming condition had been resolved when it was determined that the pipe supports were adequate to perform their safety function.

Failure to initiate a corrective action document for the pipe support deficiencies prevented management awareness of the deficient conditions, and delayed corrective action program evaluation of.the pipe support deficiencies.

The pipe support deficiencies identified in the April 23, 1987 letter should have been addressed at the time of discovery by a corrective action document.

A corrective action document has been initiated to evaluate the deficiencies identified in this concern.

b)

Although this practice is not specifically disallowed by controlling Administrative Procedure 9.11, "Engineering Analysis",

it is discouraged.

The supervisor involved in this observation has been informed that this reviewer/initiator interface is not the preferred method for obtaining independent technical review.

c)

We have identified -that several of the original IEB 79-14 cover.

sheets were signed-off despite the fact that the calculations contained within had not been updated.

Packages that had identified discrepancies were evaluated, and the discrepancies were resolved prior to restart from the last outage.

A review of the QA audit program is in progress, and will attempt to identify why deficiencies of this type have not been identified under the existing program.

Restatement of Appendix D Observation 8 50.59 Evaluations a)

HELB CRITERIA - How are the changes consistent with Generic Letter 87-11?

b)

Code Case N-411 damping values

1)

In the evaluation, the requirements of R. G._ 1.84, Rev. 24, were not mentioned nor was R. G. 1.84. referenced in the reference list.

2)

The code ease requires the use Qf current envelope spectra.

In the FSAR update it is stated that the spectra consistent with the center of mass of the piping system will be used.

This FSAR commitment is not consistent with the code case.

3)

It is indicated that the 3X damping value was used for frequencies above 10 Hz when developing the response spectra.

This does not comply with Code Case N-411 damping.

The spectral curves included in the evaluation (per Code Case N~411 damping) are therefore misleading.

c)

Evaluations for the code changes from B31.l-55 ed. and*B31.l-67

ed. to B31.l-73 ed. and ASME Section III could not be found.

Response

a)

Generic Letter 87-11 provided relaxation of the original HELB requirements specified in our FSAR.

A review of the FSAR. Generic Letter 87-11, and our design practices for new and existing systems indicates that we conform to Generic Letter 87-11.

Current requirements are addressed in Specification M-195.

"Requirements for the Design and Analysis of Palisades Plant Safety Related Piping and Instrument Tubing".

This Specification is being revised to reflect our interpretation of HELB criteria with regard to the existing systems addressed in IEB 79-14.

Revision of Specification M-195 is expected to be complete by April 30. 1990.

b)

1)

This reference is not clear in our FSAR or implementing

. procedures.

Specification M-195.,;Requirements for the Design and Analysis of Palisades Plant Safety Related Piping and Instrument Tubing" is currently being revised to satisfy R. G. 1.84. Rev. 24.. Revision of Specification M-195 is expected to be complete by April 30. 1990.

2)

The center of mass approach will not be used in conjunction with Code case N-411.

This approach was only used with the original plant spectra.

3)

This issue was discussed in our letter dated July 28, 1986, and was accepted in a letter from the NRC dated October 20,.

1986 from A. C. Thadani (NRC) to K. W. Berry (CPCo). The NRC response stated that, "therefore for the case of Palisades.

the intention by CPCo to use the 3% spectra instead of the 25 spectra from NUREG/CR 1833 is determined to be reasonable."

We will review the FSAR and make clarifications as is necessary to demonstrate how we apply these spectra.

Specification M-195. "Requirements for the Design and Analysis of Palisades Plant Safety Related Piping and Instrument Tubing". will also be revised iri. order to provide clarification of how the spectra is applied.

Revisions to specification M-195 are expected to be complete by April 30, 1990.

c)

A reconciliation of the Codes that are applicable for use when performing piping system and pipe support stress analyses under Specifications C-173, "Technical Requirements for the-Analysis of Safety Related Pipe Supports". and M-195, "Requirements for the Design and Analysis of Palisades Plant Safety Related Piping and Instrument Tubing", has been completed.

The documents which will incorporate the results of this Code reconciliation are expected to be complete by December 31, 1990.

i c*.:,

Restatement of Appendix D Deficiencies 1 and 2:

[Ref. NRR TAC Report 75155;Appendix D, page 2, and Appendix C, pages 1 and 2)

The questions raised in Appendix C remain open.

. a)

Provide justification for the use of the Reactor Building response spectra rather. than the Auxiliary Building response spectra for

the analysis of the main steam supports.

b)

Provide justification for not using Code Case N-411, together with ll. G. 1. 84 requirements, in analysis of the main steam supports since this is the FSAR commitment for any modifications of piping systems.

c) d)

e)

Provide justification for using Code Case N-411 with the original FSAR response spectra to demonstrate plant operability since this approach is not in accordance with the requirements of R. G. 1,84.

Provide resolution of the discrepancies reported in the 9/5/89 inspection report for Support No.s EB1-H6, EB1-H9.

Provide NCR's which have been issued against the deficient main steam supports.

Response

a)

The pipe stresses and support loads associated with the Main Steam System solid vertical supports are considered to be driven by the same horizontal seismic response as the Containment Building (Reactor Building) structure shell to which it is connected.

In accordance with the initial Palisades design criteria, the vertical seismic response of the Containment structure (Reactor Building) and Auxiliary Building are the same.

The _main steam lines are not designed to realize any lateral support in the Auxiliary Building.

b)

Code Case N-411 has typically been used at Palisades to evaluate new piping, and modifications to existing systems that result in a change to the dynamic properties of the associated piping system.

The U-bolt and U-bolt' attachment modification was a repair intended to restore the pipe support to its originally intended design, and did not alter the dynamic characteristics of the associated piping system.

c)

Code case N-411 was employed in a preliminary operability determination using the original Plant response spectra following a precedent established by an operability criteria that was in

place. at a different facility.

At the time that this operability determination was performed, it was for our information only.

Subsequently, an operability evaluation was performed using a new

d)

Palisades Plant operability criteria which did not need to employ the N-411 Code Case or its associated Regulatory Guide requirements.

There was no NRC Inspection Report issued on September 5, 1989, however, the deficiencies for pipe supports EB1-H6 and EB1-H9 were identified as comments by the NRC Region III inspectors during that week.

A calculation (EBl-9) has been.completed for pipe supports EB1-H6 and EB1-H9 using the as-built data.

The results from this calculation indicate that these pipe supports are acceptable in its current configuration.

e)

The NCR' s (Deviation Reports) which were issued against the deficient main steam supports are available for review.

Restatement of Appendix D Deficiency 4:

[Ref. NRR TAC Report 75155; Appendix D, page 3]

CPCo has performed a new evaluation for calculation 05956.

Based on the calculation results, the anchor bolts for one support were found to be deficient and the support is being modified.

A review of the new calculation indicates that most of the deficiencies noted by the NRC inspector ~ere corre.cted.

One exception, however, was the fact that the mass of the strainer between nodes 102 and 105 of the anaiytical model was neglected.

The significance of the omission must be evaluated.

Response

The strainer was not included in the reanalysis because its representation on isometric drawings was unclear, and the field walkdown did not translate existence of the strainer into the piping system analysis.

The analysis has been revised to include the strainer and its resultant loads.

The results of an evaluation performed after incorporation of the strainer into the support analysis indicates that the pipe stress and support loads are acceptable in the as-built configuration, and that no modification are required.

Restatement of Appendix D Deficiency 6:

[Ref. NRR TAC Report 75155; Appendix D, page 3]

The errors in the calculation for support HB35-H933 were properly addressed in a revised calculation.

The revised calculation for support HB35-H933 has not yet been provided.

Response

Calculation HB35-H933 was performed, and included an evaluation of

- pipe support HB35-H934.

The results of this calculation indicates that both pipe supports were acceptable in the as-built configuration.

The calculations are available for NB.C review.

Although no modifications were required to establish the acceptability of these pipe supports, additional weld length has been

provided to improve the margin of safety and capacity.

Restatement of Aooendix D Deficiency 7: *[Ref. NRR TAC Report 75155; Appendix D, page 3, and Appendix C, page 2]

The questions raised in Appendix C remain open.

Referenced calculation. EA-SC-89-338-04; K-8 Steam Turbine Trip Piping Stress Analysis.

a). Provide justification for use of the Auxiliary Building spectra for piping in the Turbine Building.

b)

Provide justification for not considering seismic and thermal anchor motions at points where the l" pipe connects to the 4" pipe.

se a)

Currently, there is not an available response spectra for the Turbine Building.

However, use of the ground response spectra for the Turbine Building response has typically been considered to be an appropriate assumption.

Code Case N-411 and the Auxiliary Building base slab spectra were employed for evaluation of two recent stress packages involving equipment located in the Turbine Building.

These two evaluations involved the Auxiliary Feedwater System suction piping arid discharge piping, b.oth of which are located underground in the Auxiliary Feedwater Pump Room portions of the Turbine Building.

An analysis was also performed for the Auxiliary Feedwater discharge piping with the original Plant ground spectra.

As a result of the response spectra profile, Code case N-411 analysis results* provided. more conservative results* in the Auxiliary Feedwater discharge piping analysis. It is also expected that the Code case N-411 results would have been.

similarly conservative for the Auxiliary Feedwater suction piping; however, a ground spectra evaluation was not performed for this piping.

As was discussed in the analysis of the Auxiliary Feedwater System discharge piping, Code Case N-411 results were used for the analysis of record for both Auxiliary Feedwater System problems because it is believed that Code Case N-411 input and methodologies are more consistent with current industry design and analysis practice.

b)

When the analysis of refe~ence-was performed, construction of the

piping system had not yet been completed.

As a result, the analysis of reference was not intended to be a final as-built design analysis for the piping system.

The stress evaluation associated with the 4'! pipe itself was *still under review when the analysis was performed, as was evaluation of the l" pipe.

Additionally, the seismic and thermal anchor movements* associated with the 4" pipe had been determined to be negligible based upon field in~pection results, and were therefore not included in the preliminary design analysis.

When the final design analysis was prepared the l" pipe analysis was revised to reflect as-built information, and a documented evaluation of the thermal and seismic anchor movements associated with the 4" pipe was included in the analysis package.

Restatement of Appendix D Deficiency 9:

[Ref. NRR TAC Report 75155; Appendix D, page 3, and Appendix C, page 3]

The questions raised in Appendix C remain open.

Reference calculation no. 03367, Support No. GC1-H712.

Provide the basis to retain modeling assumptions identified as errors by NRC inspectors.

Response

The stress package associated with calculation 03367 is being completely updated in order to reflect recent piping system modifications and address NRG comments.

Additionally, Specification M-195, "Requirements for the Design and Analysis of Palisades Plant Safety Related Piping and Instrument Tubing" is being revised to include lessons learned from the original IEB 79-14 program.

Changes to this specification are expected to be complete by April 30, 1990.

Restatement of Aooendix D Deficiency 11: * [Ref. NRR TAC Report 75155; Appendix D, page 3, and Appendix A, page 2]

The Calculations for Support No.s GC1-Hl26 and GC1-Hl40 did not consider friction forces.

Response

Calculations have been performed by Impell to d_etermine the as-built acceptability of these pipe supports"(GC1-H140, Rev. 1, November 14, 1989, GC1-Hl26, Rev. 1, November 14, 1989).

The results of these calculations indicate that the installed pipe support configurations are acceptable, and the the effect of neglecting friction forces was negligible for these pipe supports.

The IEB 79-14 reverification

- program that is currently in progress will provide assurance that friction forces are adequately addressed for other pipe supports.

Restatement of Appendix D Deficiency 12:

[Ref. NRR TAC Report 75155; Appendix D, page 3]

The calculation addressing the deficiency is being performed.

The adequacy of Support EB1-H25 will be assessed when the calculation is provided.

Response

Calculations have been performed by Impell (EB1-H25, Rev. 0) to update the engineering analysis for this pipe support and determine its as-built acceptability.

The results of this calculation indicate that the installed pipe support configuration is acceptable.

Restatement of Appendix D Deficiency 13:

[Ref. NRR TAC Report 75155; Appendix D, page 3, and Appendix C, pages 3 and 4]

The questions raised in Appendix C remain open.

Reference Support No.s GCl-136.2 and GCl-137.

a)

State the boundary (support). conditions which were used or assumed for the referenced supports to restrain the pipe in the computer run for the snubber reduction effort.

b)

Provide the maximum movements in the three coordinate directions at the junctions of the two supports and the pipe for loading conditions of dead, thermal, and SSE, or the combination of them.

c)

Provide detail sketches of the two supports between the pipe and the concrete anchorage surface including the size and length of structural members, welds, base plates and*anchor bolts.

d)

Provide calculations showing the maximum.tensile and shear forces or stresses in the lugs and welds of the two supports for the loading combination that will generate the maximum forces or stresses.

e)

If local stresses generated in the pipe from the lug are not considered, provioe justification for neglecting them.

Response

Existing drawings and calculations for pipe supports GCl-136.2 and GCl-137 were discussed in detail with the NRC reviewers during their audit.

These discussions included a review of original load

determinations for pipe supports GCl-136.2 and GCl-137, and an expression of our intent to remove pipe support GCl-136.2.

Removal of pipe support GCl-136.2 reduced the load on pipe support GCl-137 by over BOX, and only negligibly increased the pipe stress and pipe support _loads elsewhere in the piping system.

Removal of pipe support GCl-136.2 has effectively eliminated all load consideration for both supports, and the adequacy of pipe support GCl-137 in the*

present configuration is evident.

It is uncertain how the effects of localized stresses were addressed in the original IEB 79-14 program.

However, we will evaluate the effects of local stresses that ate induced on piping systems by integral welded attachments in our IEB 79-14 reverification program.

This evaluation will examine a sample of ten integral welded attachments in order to determine if a generic concern exists.

The evaluations will be performed in accordance. with current industry practice under Specification M-195, "Requirements for the Design and Analysis of Palisades Plant Safety Related Piping and Instrument Tubing".. This Specification is currently being revised to include integral welded attachments in the evaluation and design of nuclear

piping system supports, and will be used during future pipe support analyses.

Incorporation of the localized stress revisions into Specification M-195 is expected to be complete by April 30, 1990.

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