ML18054B125

Provides Information & Commitments Presented During 891117 Meeting W/Nrc Re Deficiencies Noted in Plant Documentation & Hardware Per IE Bulletin 79-02 & 79-14 Rework Items.Addl Info Will Be Submitted by 900110

ML18054B125
Person / Time
Site:	Palisades
Issue date:	11/21/1989
From:	Berry K CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
IEB-79-02, IEB-79-14, IEB-79-2, NUDOCS 8911280120
Download: ML18054B125 (15)

Text

consumers Power Kenneth W Berry Director PDWERINli Nuclear Licensing MICHlliA.N'S PROGRESS General Offices: 1945 West Parnall Road, Jackson, Ml 49201 o (517) 788-1636 November 21, 1989 Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 DOCKET 50-255 - LICENSE DPR PALISADES PLANT -

SAFETY-RELATED PIPING REVERIFICATION PROGRAM On November 17, 1989, a meeting at NRC White Flint offices with Consumers Power Company reviewed NRC concerns and Palisades Plant actions resulting from deficiencies identified in plant documentation and hardware associated with IE Bulletin 79-02 and IE Bulletin 79-14 rework items. Consumers Power Company committed to implement a Safety-Related Piping Reverification Program. At the close of this meeting the NRC requested Consumers Power Company to docket prior to plant startup the information and commitments presented during the meeting. It was recognized all details of the Piping Reverification Program would not be finalized and a subsequent submittal by January 10, 1990 would provide the additional information.

During recent NRC audit and inspection activities, piping stress calculation packages containing assumption errors, calculation errors, code application errors, and modeling errors were identified. The significance of any given error normally could not be determined without some level of reanalysis.

These types of errors had been previously noted by Consumers Power Company piping analysis personnel and upon reanalysis were found to be insignificant and only rarely required equipment rework.

Consumers Power Company has evaluated the NRC findings, reviewed previously identified deficiencies and concluded the majority of the errors and those errors with the most significance occurred while designing pipe supports. A sampling program was initiated by Consumers Power Company to substantiate this conclusion. As a result of our sampling effort and the recent inspection

. activity, a total of 43 pipe supports have been reevaluated. Only four pipe supports were found to be loaded in excess of the FSAR allowables and were reworked to meet the FSAR allowables. An additional nine pipe supports were reworked to gain additional margin above the FSAR allowables

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~1 Q ~* 05000255 0Cll89-0016-NL02 PDC A CMS ENERGY COMPANY

Nuclear Regulatory Commission 2 Palisades Nuclear Plant Safety-Related Reverification Program November 21, 1989 Following our review of the NRC findings, our sampling plan results and our prior years of experience using the existing stress package information, we have initiated a safety-related Piping Reverification Program. The program is designed to reverify adequacy of safety-related piping two and one-half inches in diameter and larger from a seismic design perspective. This program will involve (1) confirmation that piping isometric and pipe support drawings reflect the installed configuration and (2) assurance that the piping stress analyses and pipe support calculations also reflect the installed configuration and are in accordance with the licensing basis. Those conditions found not to be in conformance with the licensing basis will be brought into compliance.

The program will be divided into two phases. Phase 1 of the program will address the following piping systems or portions of piping systems prior to startup from the 1990 refueling outage (currently planned for mid-winter in early 1991):

Class 1 Piping Between the Primary Loop and the Second Isolation Valve High Pressure Safety Injection System Auxiliary Feedwater System Power-Operated Relief Valve Piping Between the Pressurizer and the Quench Tank The remaining safety-related piping two arid one-half inches in diameter and larger will be addressed in Phase 2 of the program. A report describing results of Phase.I and plans and schedule for Phase 2 will be provided by January 10, 1991.

Implementation of the program will involve physical walkdowns, piping stress analysis verification and pipe support calculation verification. Physical walkdowns will be used to verify piping geometry and.attributes as well as pipe support location, orientation and configuration. Stress analysis verification will consist of model verification for piping as-built configur-ation as well as other critical design inputs. Support calculation verification will involve evaluation of design parameters, methodology, conclusions, and reconciliation with the as-built support configuration and piping stress analysis.

Procedures will be developed to control the physical walkdown and the reconciliation of the as-built configuration to the piping stress analyses and pipe support calculations. Specifications will be developed addressing analytical methodology as well as design and acceptance criteria for the piping stress analyses and pipe support calculations. Documentation requirements for the piping stress analyses and pipe support calculations will also be included in the above documents. Programmatic concerns raised by the NRC in their recent inspections in this area will be addressed in the above procedures and specifications, or otherwise dispositioned. The procedures for the physical walkdown and the reconciliation of the as-built configuration with the analyses, as well as the disposition of the NRC's programmatic OC1189-0016-NL02

Palisades Nuclear Plant Nuclear Regulatory Commission Safety-Related Reverification Program November 21, 1989 3

concerns, will be available for NRC review on or about January 10, 1990. The analytical specifications will be available for NRC review by January 26, 1990. The January 10, 1990 s~bmittal will summarize the controls in the specifications and procedures.

Technical oversite of the safety-related piping reverification effort will be provided by a qualified individual(s) independent from the engineering organization performing the effort. This oversite will include technical review of a sample of the piping stress analysis and pipe support calculation verification packages.

The results of this program will include a physically-verified set of piping isometric and pipe support drawings as well as confirmation that the piping stress analyses and pipe support calculations reflect the installed configur-ation and are in accordance with the plant design and acceptance criteria.

As-built piping and pipe support configurations that are found to be outside of FSAR design criteria will be_addressed in accordance with the Palisades Plant corrective action program. Such _conditions will be first reviewed against a set of interim operating criteria (reference CPCo letter dated November 13, 1989 or later applicable revision) that are being reviewed by the NRC. If these criteria are satisfied, then the condition will be corrected at the next available opportunity, but no later than prior to startup from the 1990 refueling outage for the Phase 1 systems. Correction of any deficiencies in the Phase 2 systems will be accomplished on a schedule that is mutually agreed upon by the NRC and Consumers Power Company. If the interim operating criteria are not satisfied, then system operability will be evaluated in accordance with the Palis?des Plant Technical Specifications and appropriate actions taken as required. In addition, reportability will be evaluated per 10CFR50.

Also, during the November 17 meeting, it was pointed out that Consumers Power Company had moved very quickly from a set of identified deficiencies to the conclusion that only the pipe support calculations and installations were of poor quality, and that this conclusion was not the result of a comprehensive root cause analysis. Consumers Power Company agrees with this observation and will perform an independent review by our Quality Assurance and Nuclear Safety Services Departments to determine necessary programmatic changes and the appropriate scope of any reanalysis and field verification. This root cause review will be completed by March 31, 1990. The Piping Reverification Program will be adjusted and/or complementary activities will be initiated based on the root cause analysis results.

The final request was to explain how the NRC concerns, raised during the recent inspections and audits, would be addressed. As indicated above the programmatic concerns will be incorporated in the Piping Reverification Program. The location of these resolutions will be identified by a cross 0Cll89-0016-NL02

Nuclear Regulatory Commission Palisades Nuclear Plant Safety-Related Reverification Program November 21, 1989

• 4 reference. Resolution qf specific deficiencies will be provided with the early January submittal. A summary of the NRC Region III identified deficiencies is included in Attachment 1. Attachment 2 is a response to a specific question about completion of the IE Bulletin 79-02 project.

Kenneth W Berry Director, Nuclear Licensing CC Administrator, Region III, USNRC NRC Resident Inspector - Palisades Attachments OC1189-0016-N102

ATTACHMENT 1 Consumers Power Company Palisades Plant Docket 50-255

SUMMARY

OF NRC REGION III IDENTIFIED DEFICIENCIES November 21, 1989 6 Pages OC1189-0016-NL02

• 1 Deficiency No 1 During Inspection 50-255/89024 the inspector noted that the original hanger calculation, completed in support of IEB 79-14, for EB1-H29 (S3) and EB1-H32 (S6) evaluated catalog items for their specified loads. The results presented in the calculation indicate that all catalog items fail, however, the calcula-tion cover sheet stated no modification is required. An attempt was made to reconstruct the events that led to this conclusion, however, resolution was

• not achieved. As a part of the snubber reduction program the system was remodeled in many different ways to correlate the Bechtel pipe stress data.

The most reasonable correlation was achieved using different response spectra than was originally used, which resulted in lower loads. At this time, it was noted that the hanger U-bolt was in an overstress condition and as such, a modification (SC-87-205) was completed to replace U-bolts as necessary. A rigorous engineering evaluation was not completed at this time to evaluate other support components. Calculations performed as a result of the inspec-tion revealed the U-bolt connection assembly and structural attachment to be in excess of FSAR allowables. A modification has been completed to upgrade the U-bolt connection assembly and structural attachment to be capable of withstanding FSAR loads.

Deficiency No 2 As a result of field verification completed during the inspection, it was noted that original IEB 79-14 hanger calculations for main steam line supports EB1-H29, H32, H6A, and H9 did not reflect as-built conditions for_member size, weld size, and weld location. Engineering analyses were performed to address the as-found condition. Results indicated EB1-H29 and H32 to exceed FSAR allowables (reference Deficiency No 1), while as-built conditions for EB1-H6A and H9 are acceptable.

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• 2 Deficiency No 3 The original pipe stress analysis completed in support of IEB 79-14 considered pipe restraint direction at supports EB10-R227.1 and R227.C in a different direction than they physically restrain. Subsequently completed actions have included remodeling the pipe stress analyses to properly reflect as-built support restraining direction, and analysis of supports as necessary due to load increases using the loads derived in the pipe stress analysis. The revised calculations indicate all supports to be within FSAR allowables.

Deficiency No 4 For some pipe stress packages, Bechtel Power Corporation only provided summary calculations, with the understanding the original detailed *calculation would be available upon request. The detailed pipe stress analysis for Stress Pack-age 05956 could not be located in a timely fashion by Bechtel. Subsequent actions completed include modeling of the system using the ADLPIPE computer code, and evaluation of supports as necessary. Results indicate all supports to be within FSAR allowables. The original detailed calculations have subse-quently been located and provided to the NRC inspector.

Deficiency No 5 Record drawings and field verification completed during the inspection for hanger HC3-R133.1 indicate the as-built configuration to have a 8" x 8 11 x 1/2" baseplate with 5/8" diameter anchors. The original support calculation evalu-ated this size baseplate and anchors, using original loads, with the outcome being a factor of safety equal to 3.96. This is less than the acceptance criteria of 4.0 in the Palisades IEB 79-14/79-02 acceptance criteria. The calculation also evaluated a 9" x 9" x 5/8" baseplate with 1" diameter anchor bolts. This size plate and anchor bolts resulted in the acceptance criteria being satisfied. The pipe stress calculation was revised during the original OC1189-0016-NL02

• 3 IEB 79-14 period which resulted in substantially lower support loads. Since the 8" x 8" x 1/2" baseplate was so close to the acceptance criteria (ie, 3.96 vs 4.0} an engineering judgement, based on the revised loads, was made as to the as-built acceptability. A subsequent support calculation has been performed to show the acceptability of the support.

Deficiency No 6*

The original IEB 79-14 support calculation for hanger HB35-H933 assumes a 3 11 x 3" x 3/8" x 14" long angle to be oriented in the east-west direction with weld along the entire length (ie, 14") to an embedded plate. Subsequent walk-downs during the inspection revealed the angle is actually oriented in the north-south direction with a weld length of approximately 5".

Additionally, calculation anomalies were found to exist for hangers HB35-H933 and H934: Subsequent *engineering analysis performed for HB35-H933 shows the as-built configuration to be acceptable (a separate calculation was not com-pleted for H934 as H933 was the bounding condition). Maintenance activities have been completed to improve the weld quality for both supports. Record drawings have been revised to show proper angle orientation.

Deficiency No 7 During a field verification conducted during the inspection, it was noted that a steam trap on a 1" drain line from the steam supply to the auxiliary.feedwater pump was not well supported. The 1" drain line was modeled using the ADLPIPE computer code and was considered to be in excess of FSAR allowables. A modification (SC-89-338) has been compl~ted to add supports so the piping is within FSAR allowables. This 1" pipe was not directly evaluated in the IEB 79-14 program due to the nominal pipe size.

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• 4 Deficiency No 8 During the inspectors' review of calculations performed to support resolution of Deficiency No 1 it was noted that a mathematical error was made for total weld length. Revision to the calculation has been performed with the results indicating no adverse affect on calculation results.

Deficiency No 9 During inspector review of an ongoing modification for the auxiliary f eedwater turbine driven pump steam supply relief valve piping configuration (SC-88-138),

discrepant conditions such as stress intensification factor usage, hanger location, and response spectra were noted within Stress Package 03356. As part of resolution to Deficiency No 4, the inspector's concerns were incorporated as was deemed appropriate. '

Deficiency No 10 As a result of the snubber reduction program, calculated loads at support locations were increased or decreased as snubbers were replaced by struts or removed. When loads increased, the support acceptability was qualitatively evaluated and documented in text rather than in a revised calculation. At the inspectors'.request, a few support calculations were reviewed in detailed cal-culation form relative to the new loads. The original IEB 79-14 Bechtel calculation for GC1-H712 and the revised calculation were reviewed by .the inspector. Mathematical errors in determining the moment of inertia, section modulus, and cross sectional area in the original Bechtel calculation were noted, however, the revised calculation indicates the supports' acceptability.

Deficiency No 11 The original IEB 79-14 calculations for supports GC1-Hl26 and Hl40 did not OC.l 189-0016-NL02

• 5 consider friction loads on the supports. An engineering analysis has been performed for both supports to include friction. The results indicate the as-built condition to be acceptable.

Deficiency No 12 As a result of field verificatio.ns performed during the inspection, it was noted that the as-built condition of EB1-H25 did not match the design drawing.

It was noted the attaching angle clip and associated weld were larger than specified on the drawing. Subsequent analysis indicated the as-built condi-tion acceptable. Record drawings have been revised to reflect the as-built configuration.

Deficiency No 13 Local stresses on piping due to welded attachments were not evaluated for supports GC3-Hl25, GC1-H136, and GC1-H137. Evaluation of welded attachments to piping was not considered to a great extent during the Palisades IEB 79-14 program. The evaluation criteria for the Palisades 79-14 program provided the following criteria:

"The following defines the scope of responsibility of the pipe support group with regards to the evaluation of existing as-built welded attachments to the pipe:

A. Check the weld in shear (make sure there is enough weld metal to take the applied load)

• B. Check the attachment (eg, lugs) in bearing (make sure there is enough material surface area to take the applied load).

C. The effect of the attachment with regards to the pipe itself is excluded from pipe support analysis."

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• 6 An engineering calculation which was considered to be bounding for identified deficiencies was performed for GC3-Hl25. The conclusion was that as-built conditions are acceptable.

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ATTACHMENT 2 Consumers Power Company Palisades Plant Docket 50-255 RESPONSE TO NRC CONCERN ON COMPLETION OF ANCHOR BOLT REPLACEMENT PER IE BULLETIN 79-02 November 21, 1989 3 Pages OC1189-0016-NL02

• 1 Concern to Attachment C of a February 14, 1980 docketed Consumers Power Company report provided a summary report of Palisades Plant compliance to the interim operating criteria for IE Bulletin 79-02. Page 5 of .this attachment stated; "The ultimate capacity of existing pipe support anchor bolt sizes was compared with the calculated support loads. In 91.5% of the cases (a total of 514 anchors analyzed), the factors of safety were found to be equal to two or more. Further detailed analyses on those supports having safety factors of less than two are in progress. The results to date indicate that the calculated safety factors will be in excess of two. Based on this, it is expected that the remaining population of anchor bolts will have a safety factor of two or more."

Subsequent docketed reports from Palisades did not clearly specify the disposition of the 44 (8.5% x 514) anchor bolts with less than a safety factor of 2. Also, there were no positive statements on how the safety factor of 4 or 5 would be reestablished.

Response

The evolution of IE Bulletin 79-02 and IE Bulletin 79-14 requirements were specified in 10 publications over an eight month period. Consumers Power Company met the Bulletin 79-02 requirements by implementing a testing and inspection program to verify the anchor bolts met or exceeded the interim operating criteria. When the Bulletin 79-02 project had adequate data to provide a 95% confidence that in 95% of the cases all anchor bolts had been installed properly and the safety factor of 2 was met, the Bulletin 79-02 project was terminated. The restoration of the safety factor of 4 or 5 was included in the scope of work for the Bulletin 79-14 project.

0Cll89-0016-NL02

• 2 The safety factor of 4 of 5 could not be verified by inspection or test, it had to be verified by comparison of the required or analyzed load to the manufacturer's rating. The poor quality and lack of retrievability of the Palisades stress calculations dictated *the project include a complete reanalysis of all seismic rated piping. This meant all piping calculations and hanger loads must be rerun prior to verifying the safety factor of 4 or 5.

The NRC's reference to the above statement in question referred to the status of the recalculation efforts as of February, 1980. In order to make defendable conclusive comments on th~ closeout of the Bulletin 79-02 project and the Bulletin 79-14 project, quality records were retrieved and interrogated. The objective evidence collected is:

1. Consumers Power Company Surveillance Report dated June 4, 1980, the ninth in a series of surveillances, which indicated "179 anchors failed removal, 99 anchors failed the 200% design load pull test, and 1236 failed engineer-

. ing criteria, for a total of 1514 failures, all of which have been repaired or replaced." This indicates the major reason for replacing anchor bolts was to achieve the calculated safety factor of 4 of 5.

2. Bechtel Quality Control Inspection Record dated June 24, 1980 indicated an interim turnover by reviewing the inspection record file log to verify "that the required receiving, installation and testing inspections for the equipment, components and structures included in the system, subsystems or other designated items to be turned over have been satisfactorily completed."

3. Consumers Power Company Quality Control Surveillance Report dated December 23, 1981 was written to document the final report covering modifications carried over from the last refueling outage. The report concluded "All of the 271 identified fixes have now been completed. No QA concerns were identified for this project during the 1981 refueling outage."

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4. Bechtel Quality Control Inspection Record dated January 28, 1982 indicated a final project review which concluded "All specified work, including repairs has been satisfactorily.completed, and no unspecified work has been added without the proper engineeri~g approvals."

5. In September and December of 1979 Region III issued two mid-project

,inspection reports documenting five separate inspections. On November 29, 1983, the final Bulletin 79-02 and Bulletin 79-14 close out inspection report was issued concluding "As a result of this follow-up review and previous reviews, the inspector considers that all pertinent issues and findings relative to IE Bulletin 79-14 have been resolved. As a result of this follow-up review and previous reviews, the inspector considers that all pertinent issues and findings relative to IE Bulletin 79-02 have been resolved."

In addition to these quality verification documents a few Field Change Requests were reviewed to verify anchor bolts and base plates were being repaired or replaced during modifications resulting from IE Bulletin 79-14 reanalysis efforts. Two of the five document packages inspected included such evidence. This data leads Consumers Power Company to the conclusion all existing anchor bolts have a safety factor of 4 or 5 when the existing hanger calculated loads are compared to the manufacture's_-specification.

It is understood the type of deficiencies identified by the NRC and Consumers Power Company may identify some anchor bolts to have a safety factor of less than 4. Therefore, anchor bolts and base plates have been included within the scope of the Piping Reverification Program.

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