ML20136D650

Forwards Rev 1 to Confirmatory Reactor Bldg Basemat Analysis for Hope Creek Generating Station Pse&G in Response to NRC Request for Addl Info Re Idvp.Addl Info Will Be Provided Prior to Fuel Load to Verify Closeout of Idvp

ML20136D650
Person / Time
Site:	Hope Creek
Issue date:	12/26/1985
From:	Martin T Public Service Enterprise Group
To:	Ankrum G Office of Nuclear Reactor Regulation
Shared Package
ML20136D655	List: ML20136D650 ML20136D658
References
NUDOCS 8601060233
Download: ML20136D650 (9)

Text

r Public Service Electric and Gas Company Thomas J. Martin 80 Park Plata, Newark, NJ 07101 201-430 8316 Maihng Address. P.O. Box 570, Newark, NJ 07101 Vice President

. Engineering and Construction December 26, 1985 Director of Nuclear Reactor Regulation United States Nuclear Regulatory Commission 7920 Norfolk Avenue Bethesda, Maryland 20814 Attention: Mr. G. Ankrum Quality Assurance Branch Division of Quality Assurance, Vendor, and Technical Training Center Programs Gentlemen:

INDEPENDENT DESIGN VERIFICATION PROGRAM ADDITIONAL INFORMATION HOPE CREEK GENERATING STATION DOCKET NO. 50-354 As you requested, the followin.g is the current status of~

the closcout of the Hope Creek Independent Design Verification Program (IDVP). Included as Attachment 1 is the specific status of eight, on-going programs identified in the IDVP Final Report, namely Hazards, As-Built Reconciliation, Seismic II/I, Load Verification, Setpoint Calculation, Environmental Qualification, Reassessment of Raceway and HVAC Hangers and Ducts, and Equipment Anchorage. Particular detail is included on the Hazards, and Setpoint Calculations, as you requested. Our evaluation of the Confirmatory Reactor Building Basemat Analysis and the Confirmatory Analysis, Rev. O, dated October 1985, is included as Attachment 2.

Our evaluations and conclusions are covered in a Memorandum to the Chief Project Engineer - Hope Creek, from the Site Engineering Manager - Hope Creek, dated December 20, 1985.

Attachment 3 is our required follow-up of Observation Report No. 145, regarding the issue of seismic qualification of safety related panels. This report provides confirmation of the response to OR 145.

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Director of Nuclear Reactor Regulation 12/26/85 is our review and evaluation of flange bolt design for non-NSSS action pumps. This was requested in connection with the close-out of OR 144.

General Discussion As a result of the IDVP, Sargent & Lundy issued 221 Observation Reports, of which 159 were determined to be valid. Public Service Electric and Gas Company committed to independently verify that all follow-up action items are being completed and correctly implemented. This verification process requires a review by PSE&G personnel of all commitments made in the responses to the OR's, and includes review and inspection of document changes such 63 FSAR, specifications or procedures, a review of calculations, or a review of the results of plant walkdowns. In some cases, independent walk-downs are performed by PSE&G personnel to verify other conclusions.

Currently, essentially all follow-up actions have been completed and verified. The exceptions to this are OR's involved in the eight programs mentioned above. The status of the close-out of each of these programs is provided on . As you will note, all of the programs are essentially complete, and we expect they will be complete by fuel load. It is PSE&G's conclusion that each of these programs is being implemented in accordance with the commitments made in the IDVP, and that each demonstrates that the plant is designed adequately, and in accordance with the FSAR.

FSAR Changes Several ors required changes to the FSAR. In all but one case, the FSAR change notices have been approved by PSE&G. Many of the changes have been included in FSAR amendments. The remaining, including the change notice not yet approved by PSE&G, will be included in Amendment 14, which is projected to be issued prior to fuel load. The FSAR change which has not been approved involves the ABR program, and is being prepared as that program is being completed.

On-Going Programs As noted, Attachment 1 provides the overall status of the eight programs which were identified in the IDVP. As you requested, the following additional information is provided.

Director of Nuclear Reactor Regulation 12/26/85 Hazards Program - Final Phase General This program addresses the effects of postulated High and Moderate Energy Line Breaks (HELB and MELB). The HELB review evaluates the consequences of potential pipe whip, jet impingement, flooding and compartment pressurization. The MELB review evaluates the consequences of flooding and water spray.

The Hope Creek design has addressed HELB and MELB considera-tions through three mechanisms. Design criteria documents were issued during the design and construction phase specify-ing separation requirements, defining hazards, and providing for hazard reviews. Examples of these criteria documents are D7.3, " Procedure for Documentation and Criteria of Plant Separation", issued for comment on March 4, 1982 and use on March 20, 1983, and D7.9, " Field Routed Procedure",

issued for comment on July 28, 1981 and use on March 6, 1983.

Separation reviews were held during 1982 and 1983 utilizing the project model and drawings to locate problem areas early in the construction phase. The results of these separation reviews are documented on separation review data sheets, as required by D7.3 Appendix E. Final verification of the design has always rested on a final hazard walkdown, to be performed when construction on systems and areas in the plant are completed. Project Specification G-19.1, " Procedure for Jobsite Review of Hazards Effects", issued for use on February 28, 1985, provides guidance for performing these walkdowns and requires they be performed approximately two weeks prior to room turnover from BPC to PSE&G. This requirement assures that the review is properly based on as-built conditions.

The scope of the existing hazards review program was not appreciably changed as a result of the IDVP. Increased emphasis has been placed on written guidance documents and clear documentation of review decisions and results.

Status This program is currently 93 percent complete. The final walkdowns occur approximately two weeks prior to room / area turnover to PSE&G. It is expected that this program will be complete by fuel load.

Director of Nuclear Reactor Regulation 12/26/85 Verification Process The verification of this program by PSE&G personnel is accomplished by several methods. PSE&G participated in great detail in the preparation of the specifications which provide the requirements of the program. Our personnel are participating in the preparation for walkdown, and in some cases are involved in the walkdowns. In addition, indepen-dont walkdowns are performed where we deem necessary.

In order to assure that all required areas of the plant have been reviewed, PSE&G inspects the Separation Review Data Sheets to verify that all areas are covered. A ' uailed review of the entire hazards analysis is then completed by PSE&G on a sampling basis to assure and verify the technical requirements of the specifications are being met.

It is noted that to date, only minor changes have been required as a result of the final walkdowns.

Conclusion As a result of the review and verification performed to date, it is PSE&G's conclusion that the hazards program, including the final portion is being done in accordance with the speci-fications and commitments made in the IDVP. It is further concluded that this program and its results demonstrate that the plant is adequately designed, and meets the commitments of the FSAR.'

Setpoint Calculation Program General The setpoint calculation program was initiated in September of 1984 by Bechtcl. At that time, " committed" calculations for all Balance of Plant lE instruments were started.

Calculations were issued as " committed" since the final seismic and environmental qualification reports were not available for the devices for which the calculations were performed. When committed calculations are made " final",

the complete " committed" calculation is essentially redone utilizing final data and including omitted data, references, equations and justifications which caused the original calculation to be " committed". Adherence to Bechtel procedure EDP 4.37 assures that all committed calculations have been

Director of Nuclear Reactor Regulation 12/26/85 finalized and include a statement of which factors were considered and which were not and the basis for each exclusion.

f PSE&G is currently generating separate, final setpoint calcula- l tions following the GE setpoint methodology which is under reviei by the NRC. The PSE&G setpoint effort as covered in '

PSE&O Site Engineering Instructionc (SEI) 3.4. uses data from the Bechtel work but supercedes the setpoint calculations done by Bechtel.

The setpoint program includes calculations for those setpoints established by Bechtel, General Electric and other vendors.

These calculations are being used to verify the setpoints established in the issued setpoint documents.

The scope of this work includes " mini" and " full" calculations for Q setpoints. The " mini" calculations verify instrument accuracy, calibration tolerance, setpoint, recalibration tolerance, reset point and scaling calculations. The " full" calculation includes the effects of environmental and seismic considerations.

Status As shown on Attachment 1, the overall setpoint calculation program is considered 90 percent complete, although essentially all committed calculations are complete. It is important to note that the work done by PSE&G to date under SEI 3.4, has resulted in no setpoint changes from the committed calcula-tions. It is expected that the full calculations for all Q setpoints will be complete by fuel load.

Verification Process In order to meet the IDVP commitment to provide independent verification of the setpoint program, the following is being done. Independent, experienced PSE&G engineers are reviewing the program to assure all required setpoints are being calcu-lated and entered onto the setpoint register. Further, on a sampling basis, calculations are being reviewed to assure that all requirements of the appropriate procedures are being met.

This verifica-ion process is on-going, and is expected to be complete by fuel load.

Director of Nuclear Reactor Regulation 12/26/85 Conclusions As a result of the work done to date, it is concluded that the setpoint program is being implemented in accordance with the appropriate procedures and the commitments made in the IDVP. It is further concluded that the results of this program demonstrate that the plant is adequately designed and meets the commitments in the FSAR.

Documentation Disposition As discussed with Mr. J. Milhoan, the documentation gathered and prepared by Sargent and Lundy is currently on file in Chicago. In addition, other material is available in Bechtel offices in San Francisco. These files will be retained until completion of tb; 100 hour0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> operational test, at which time we will dispoc a of the IDVP material.

Final el oe-out Prior to fuel load, we will provide you with additional correspondence to verify to you the final closecut of the IDVP. If we can be of further assistance, please do not hesitate to contact us.

Very truly yours, t a b r w z.

T. J. Martin Vice President -

Engineering and Construction WG:mlb Attachments C Ms. Elinor Adensam, Director USNRC - Project Directorate 3 D. H. Wagner USNRC Licensing Project Manager R. W. Borchardt USNRC Senior Resident Inspector

ATTACfD1ENT 1 IDVP - PROGR7M CLOSE OUT PROGRAM PROGRAM BEING IMPLEMENTATION PROGRAM METHOD OF TECH ADBQUACY FOLLON-UP RESPONSIBILITY DEFINED IMPLEMENTED COMPLETE / STATUS MEASURING RESTRAINTS OF RESULTS REQ'D IIAEARDS W. PAVINCICH YES E 93g ROOMS / AREAS ABR ACCEPTABLE MINOR ABR - PIPE & R. EIRK YES M PINAL SUPPORTS 97g calfs AREA ACCEPTABLE MINOR TURNOVERS S2ISMIC II/I G. LUM YES YES NOME 99g ROOMS / AREAS NONE ACCEPTABLE TO DATE LOA 3 VERIPICATION M. REESER YES YES 93g CALCS NONE ACCEPTABLE MINOR LETPOINT CAirS. R. ROSKO YES YES NONE 90% CALCS NONE ACCEPTABLP TO DATE ENVIRONMENTAL QUAL. R. D'ORAEIO YES YES 99g REPORTS MONE NONE ACCEPTABLE TO DATE dEASSESS RACEWAY & C. LUM YES YES NONE HVAC RANGERS & DUCTS 97g CAIES gogg ACCEPTABLE TO DATE EQUIP. ANCHORAGE M. REESER YES YES WONE 1001 CALCS NONE ACCEPTABLE

ATTACHMENT 2 O PSEG Public Service Electric and Gas Company P. O. Box A Hancocks Bridge, New Jersey 08038 December 20, 1985 Hope Creek Generating Station File 403.2 (SE.85.12.16-65)

To the Chief Project Engineer - Hope Creek COMPLETION REPORT ON REACTOR BUILDING BASEMAT CONFIRMATORY ANALYSIS INDEPENDENT DESIGN VERIFICATION PROGRAM

REFERENCES:

1. LETTER FROM PUBLIC SERVICE ELECTRIC & GAS COMPANY (MR. R. L. MITTL) TO NRC LICENSING BRANCH 2 (MR. W. BUTLER), DATED OCTOBER 30, 1985

2. OBSERVATION / RESOLUTION / COMPLETION REPORT NOs. 7 AND 40

3. FINAL REPORT, HOPE CREEK GENERATING STATION, INDEPENDENT DESIGN VERIFICATION PROGRAM ~

4. FINAL REPORT - CONFIRMATORY REACTOR BUILDING BASEMAT ANALYSIS, REVISION 1 HOPE CREEK GENERATING STATION During the Independent Design Verification Program, IDVP, for Hope Creek Generating Station, PSE&G and BPC agreed to perform a confirmatory analysis for the reactor building basemat using smaller size elements in the plan and through the thickness of the basemat. This analysis addresses all basemat comments raised by Sargent & Lundy as referred to in References 2 and 3.

The purpose of this report is to verify and document the com-pletion of the analysis mentioned above.

PSE&G has reviewed the " Final Report - Confirmatory Reactor Building Basemat Analysis, Revision 1," (Reference 4) and the associated calculation furnished by BPC. Following are our comments for the review:

• A verified BPC Structural Analysis Computer program, BSAP (CE800 ), was used to perform the analysis.

• The analytical model uses element sizes which are smaller than that used in the original basemat analysis. It has smaller plan size elements and five elements'through the thickness of the mat. A parametric study was carried out for the comparison (Figures C-7 through C-ll of Reference 4) between 3, 4, and 5 layers. Results indicate that a model with five layers of elements provides element stresses which are sufficiently accurate to calculate the appropriate bending moments. The five-tier basemat model was discussed with and accepted by Sargent & Lundy prior to the analysis.

Chief Project Engineer - 2 Hope Creek

• Methodology, approaches, and results used i performing the analysis were reviewed and found to be consistent with the report (Reference 4), and acceptable.

• Loads and load combinations used in the analysis are in accordance with Design Criteria D2.1.

• Twisting moment was addressed properly in the report using CECAP Computer program. The provided reinforcement is

' adequate for all major loadings including the effects of twisting moment.

• Capacity reduction factor was addressed. A minimum value of 0.84 as allowed by ACI 318-71 was used in the evalua-tion of the combined moment and axial forces. .

• Both horizontal and vertical construction joints were re-viewed and found to be adequate to transfer the design loads.

• Thermal load, torus uplift load, and seismic inertia loads due to containment flooding were properly accounted for by the existing design. This is again demonstrated in the report.

• Shear and flexural reinforcements meet the applicable code requirements; calculated maximum reinforcement and concrete stresses are less than the allowable values.

After reviewing the subject report ( Re f erence 4 ) , it is concluded that the confirmatory analysis for reactor building basemat addresses all Sargent & Lundy IDVP review comments, and the analysis results further' demonstrate that HCGS reactor building basemat is adequately designed and is in agreement with applicable design codes, design criteria and FSAR commitm nts.

W C. W. Churchman Site Engineering Manager -

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Attachment:

1. BLP 18,053, " FINAL REPORT-CONFIRMATORY REACTOR BUILDING BASEMAT ANALYSIS,

REVISION 1" C A. S. Kao l M. C. Reeser SE File i STAIRS