IR 05000498/1999005

Insp Repts 50-498/99-05 & 50-499/99-05 on 990301-05.No Violations Noted.Major Areas Inspected:Engineering

ML20205E598
Person / Time
Site:	South Texas
Issue date:	03/29/1999
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20205E587	List: IR 05000498/1999005 ML20205E583
References
50-498-99-05, 50-498-99-5, 50-499-99-05, 50-499-99-5, NUDOCS 9904050234
Download: ML20205E598 (9)

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ENCLOSURE U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

Docket Nos.: 50-498;50-499 License Nos.: NPF-76; NPF-80 -

Report No.: .50-498/99-05;50-499/99-05 Licensee: STP Nuclear Operating Company Facility: South Texas Project Electric Generating Station, Units i and 2 Location: FM 521 - 8 miles west of Wadsworth Wadsworth, Texas Dates: March 1 to 5,1999 Inspectors: R. L. Nease, Senior Reactor inspector, Engineering and Maintenance Branch P. A. Goldberg, Reactor inspector, Engineering and Maintenance Branch Approved By: Dr. Dale A. Powers, Chief Engineering and Maintenance Branch ATTACHMENT: Supplemental Information l

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I 9904050234 990329 PDR ADOCK 05000498 G PDR

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f -2-l EXECUTIVE SUMMARY - .

,, South Texas Project Electric Generating Station, Units 1 and 2 NRC Inspection Report No. 50-498/99-05; 50-499/99-05 l Enaineerina l .

Safety evaluations and an engineering evaluation report supporting replacement of the-Unit 1 steam generators prepared by licensee contractor personnel were performed in accordance with licensee procedures, met regulatory requirements, and were of high quality (Section E2.1).

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-3- l Report Details 111. Enoineerina j E2 Engineering Support of Facilities and Equipment E Evaluation of 10 CFR 50.59 Safety Evaluations for Steam Generator Replacement

) Insoection Scope (50001)

The inspectors reviewed selected cafety evaluations associated with the replacement of i the Unit 1 steam generators scheduled for the fall of 2000 Unit 1 Refueling Outage l This review was performed in accordance with NRC Inspection Procedure 50001, I

" Steam Generator Replacement inspection," and the purpose was to evaluate design changes to verify that they were performed in accordance with the requirements of 10 CFR 50.59. Specifically, the inspectors reviewed the following safety evaluations:

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Change 96-2842-4 Reactor Coolant System Work l Change 96-2843-1 Large Bore Secondary System - 1RE08 Baseplates Change 96 2843-2. Large Bore Secondary System Pipe Work (FW, AF, MS)

Change 96-2844-1 ~ Steam Generator Water Level Tubing / Piping  !

Removal / Reinstallation  ;

Change 96-2844-2 Steam Generator Blowdown System Piping l Removal / Reinstallation Change 96-2845-1 Make Biological Shield Wall D Removable

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! Change 96-2845-11 STP Unit 1 Steam Generator Replacement Polar Crane l Evaluation Report The licensee used Procedure OPGP05-ZA-0002,"10 CFR 50.59 Evaluations,"

Revision 9, for conducting 10 CFR 50.59 safety evaluations. A recent NRC review of the licensee's 10 CFR 50.59 process, including a review of this procedure, is discussed I in NRC Inspection Report 50-498; -499/98-1 Observations and Findinc::

The licensee evaluated and, where necessary, reperformed the original accident analyses to address the changes associated with the new steam generators. The results were presented to the Office of Nuclear Reactor Regulation in December 199 To date, the small-break-loss-of-coolant-accident was the only accident that was revised and submitted for approval. Changes to the remaining accident analyses are being evaluated under 10 CFR 50.59. The resulting changes were expected to be completed

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-4-by September 1999. All of the safety evaluations reviewed by the inspectors were performed by either Bechtel Engineering or Westinghouse, and reviewed and accepted by the license Desian Chanae Packaae and Unreviewed Safety Ouestion Evaluation 96-2842-4 The team reviewed Design Change Package and Unreviewed Safety Question 96-2842-4," Reactor Coolant System Work," Supplement 0, which was prepared for the replacement of the Unit 1 steam generators during the Unit 1 Refueling Outage 9. The design change package is for the removal of the existing Model E steam generators, which will be replaced with Model 94 steam generators. The modification analysis consisted of severing the reactor coolant piping from the existing steam generators, providing temporary supports for the reactc r coolant piping, removing the old steam generators, and installing the replacement steam generators and permanent piping supports. The licensee's representative stated that this design change did not stand alone but was dependent on Design Change Package and Unreviewed Safety Question Evaluation 96-2842-3. The licensee's representative stated that this design change package and safety evaluation was not yet complete, but would be available in September 1999. This package was to contain a safety evaluation based on the accident analysis performed by Westinghouse for the replacement steam generator j Design Change Package 96-2842-4 was dependent on the accident analysis, which was completed by Westinghouse in 1977. The licensee's representat.ve stated that, if there are any changes to the accident analysis, this design change package would be revise The inspectors observed that the licensee did not intend to hydrostatically test the primary and secondary system welds to the steam generators since ASME Code  !

Case N-416-1, " Alternate Pressure Test Requirement for Replacement items By 1 Welding Class 1,2, and 3,Section XI, Division 1," had been adopted. The NRC '

program office authorized the use of this code case for the licensee in a letter dated July 14,1997. The inspectors learned that the licensee intended to perform radiography on all welds that attach the replacement steam generators to the primary and secondary systems in accordance with Section lil of the ASME Code. In addition, the licensee's representative stated that the steam generators would be hydrostatically tested by the manufacturer. The details of the return-to-service testing was not yet formalize The inspectors noted the licensee determined in the safety evaluation that thmo was no unreviewed safety question. The inspectors observed that the design charG > package and associated 10 CFR 50.59 safety evaluation were comprehensive and of good qualit Desian Chanae Packaae and Unreviewed Safety Question Evaluation 96-2843-1 The inspectors reviewed Design Change Package and Unreviewed Safety Question Evaluation 96-2843-1, "Large Bore Secondary System - 1RFO8 Baseplates,"

Supplement 0, which was prepared for installation of 16 baseplates on the steam generator side of the biological shield walls during the spring of 1999 Unit 1 Refueling Outage 8. These baseplates will support new pipe supports for rerouted feedwater lines. The pipe supports and rerouted feedwater lines will be installed during the fall of

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2000 Unit 1 Refueling Outage 9. The inspectors noted that the design analysis only j provided for sustaining the dead weight of the plates themselves, including Seismic

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Category ll/l loading for the period of time between Unit 1 Refueling Outages 8 and The licensee's representative stated that these baseplates will be re-analyzed for loads ,

from the pipe supports resulting from the feedwater pipe stress analysis to support I installation of the pipe supports and the rerouted feedwater lines. The permanent piping, pipe support, and base plate changes were to be evaluated in a future 10 CFR 50.59 safety evaluatio The licensee's safety evaluation did not identify an unreviewed safety question. The inspectors found that the licensee's safety evaluation for installation of Design Change Package and Unreviewed Safety Question Evaluation 96-2843-1 was adequate for j meeting the requirements of 10 CFR 50.5 J l

Desian Chanae Packaae and Unreviewed Safety Question Evaluation 96-2843-2 i

The inspectors reviewed Design Change Package and Unreviewed Safety Question {

Evaluati3n 96-2843-2,"Large Bore Secondary System Pipe Work (FW, AF, MS)," l Supplement 0, which supported the removal and rerouting of feedwater, auxiliary feedwater, and main steam lines, and their associated temporary and permanent ;

support systems, inspectors noted that the licensee performed piping stress )

calculations for feedwater, auxiliary feedwater, and main steam lines when temporarily j supported during the interim period between their removal and reinstallation. Temporary j pipe supports were evaluated for Seismic Category 11/1 design. The licensee's piping l stress analyses for the permanent piping installations and the supporting sfstems ;

resulted in stresses within allowable limits. The licensee's evaluation included high j energy line breaks, flooding, potential missile targets, and changes in material mass in I containmen j The licensee's 10 CFR 50.59 safety evaluation did not identify an unreviewed safety '

question. The inspectors found that the licensee's evaluation supporting P sgn Change Package and Unreviewed Safety Question Evaluation 96-2843-2 was conservative and comprehensiv Desian Chanae Packaae and Unreviewed Safety Question Evaluatio,196-2844-1 The inspectors reviewed Design Change Package and Unreviewed Safety Question Evaluation 96-2844-1, " Steam Generator Water Level Instrumentation Tubing / Piping Removal / Reinstallation," Supplement 0, which was prepared to disconnect the steam ,

generator water levelinstrumentation piping and tubing in the vicinity of the steam l generators to allow for the removal of the existing generators. in addition, the design l change package contained instructions for reconr acting the piping and tubing at the )

new locations on the replacement steam generators. The inspectors noted that

setpoints were rot included as part of this design change package. The design change l

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package was to be implemented during the Unit 1 Refueling Outage 9. The inspectors I noted that the steam generator water level instrumentation consisted of wide and narrow range instrumentation systems with a total of nine tap locations per generator, which would be affected by this effort.

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-6-The licensee's 10 CFR 50.59 safety evaluation reviewed the steam generator replacement activities associated with the modification of the steam generator water level instrumentation, The inspectors found that the design change package and the safety evaluation were comprehensive and of high qua'lit Desian Chanae Packaae and Unreviewed Safety Question Evaluation 96-2844-2 The inspectors reviewed Design Change Package and Unreviewed Safety. Question Evaluation 96-2844-2," Steam Generator Blowdown System Piping Removal /

Reinstallation," Supplement 0, which was prepared to disconnect the steam generator blowdown piping from the existing steam generators and reroute and connect the piping to the replacement steam generators. The rerouting of the piping wac necessary to accommodate changes in the blowdown and shell drain nozzles locations on the replacement steam generators. This design change was to be implemented during Unit 1 Refueling Outage The inspectors noted that the licensee's 10 CFR 50.59 safety evaluation reviewed the design, removal, inctallation, and return-to-service activities associated with this design change packag The inspectors dete.mineo uiat the design change package and tN associated safety resaluation were detailed and of high qualit Desian G wae Packaae and Unreviewed Safety Question Evaluatiorevam The inspectors reviewed Design Change .osckage and Unreviewed S afety Question Evaluation 96-2845-1, "Make Biological Shield Wall D Removable," Supplement 0, which was prepared to make a 20 h *t long by 14 feet high section of the biological shield wall in front of the reactor containment building equipment hatch and near Steam Generator D removable. In addition, the design change package provided instructions to reattach the wall section to the adjoining wall by steel splice plates and bolts. This work was to be performed during Unit 1 Refueling Outage 8. Removal of the wall to provide clearance for steam generator replacement, removal of the old steam generators and installation of the replacement steam generators through the equipment hatch, and reinstallation of the wall was not part of this design change package. The safety evaluation was to be accomplished during Unit 1 Refueling Outage i The inspectors noted that the licensee's 10 CFR 50.59 unreviewed safety question evaluation was prepared because the description of the biological shield wall in the L Updated Final Safety Analysis needed to be revised. The licensee determined that an l unreviewed safety question did not exist. The inspectors observed that the design

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change package and associated safety evaluation were comprehensive and of high quality.

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, g Condition Reoort Enainserina Evaluation 96-2845-11 '

The inspectors reviewed Condition Report Engineering Evaluation 96-2845-11,"STP Unit 1 Steam' Generater Replacement Polar Crane Evaluation Report," dated March 3,-

l 1998, which evaluated the existing polar crane capacity for the new loads resulting from (1) the addition of a temporary handling device used to remove existing Unit 1 steam generators and (2) installation of the replacement steam generators. The combined

weight of the temporary lifting device and the replacement steam generators exceeds

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the rated capacity of the polar crane; therefore, the licensee determined that the use of the polar crane as described will be performed as a ." Planned engineered Lift" in accordance with ASME B30.2-1996," Overhead and Gantry Cranes." Based on review of the polar crane service records, an inspection of the crane and its support system,

and their evaluation of the crane for the engineered lifts, the licensee's crane contractor, l 1 Whiting Corporation, specified operating restrictions during the engineered lifts. In addition, the design of the polar crane and supporting system (including the attachment to the containment shell) was reviewed by Bechtel Engineering. 'Bechtel Engineering 4 . determined, as documented in an attachment to this ' condition report engineerin0 i

' eraiuation, that there was sufficient margin in the design of the supporting system to handle the required loads. The inspectors noted that this condition report engineering evaluation did not address the structural qualification of the temporary lifting device, the actual rigging, or the 10 CFR 50.59 safety evaluation. These were to be considered in a separate condition report engineering evaluatio The inspectors found that the licensee's evaluation supporting Condition Report Engineering Evaluation 96-2845-11 was comprehensive and adequately addressed the . )

issues within its scop Conclusions 1 j

Safety evaluations and an engineering evaluation report supporting replacement of the i

Unit 1 steam generators prepared by licensee contractor personnel were performed in accordance with licensee procedures, met regulatory requirements, and were of high -

qualit i E7 Quality Assurance in Engineering Activities j

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' E Oversiaht of the Steam Generator Fabrication Process I

- The inspectors reviewed licensee quality assurance documents regarding oversight of j the steam generator fabrication process. The licensee currently had two full-time {

inspectors covering both fabrication shifts at the Westinghouse Pensacola Plant. The l

[' licensee continued to provide oversight of the Unit 1 steam generator fabrication ;

process, which,~as discussed in NRC Inspection Report 50-498; -499/98-14, dated l August 31,1998, was considered to be aggressive and proactiv j

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-84 The Unit 2 sicam generators were to be fabricated at Equipos Nucleares, S.A. In Maliano, Spain. Fabrication has not begun; however, as of December 1998, the licencee had assigned two South Texas Project plant employees to the Equipos Nucleares, S.A. fabrication facility for full-time oversight of the development of shop and fabrication drawings. Although too early to draw a conclusion on the effectiveness of the licensee's oversight of the fabrication of the Unit 2 steam generators, the inspectors .

were encouraged by the licensee's early oversight activitie V. Manaaement Meetina X1 Exit Meeting Summary On March 5,1999, the inspectors presented the inspection results to members of licensee management. The licensee representatives acknowledged the findings presente The inspectors asked the licensee representatives whether any materials examined during the inspection should be considered proprietary. The licensee representatives stated that some material was considered proprietary and requested it be returned. No proprietary information was included in this report, but such information will be returned to the license '

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ATTACHMENT SUPPLEMENTAL INFORMATION .

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[ PARTIAL LIST OF PERSONS CONTACTED Licensee l

R. Attar, Supervisor, Design Engineering D. Chamberlain, Consulting Engineer, Design Engineering W. Cottle, President and Chief Executive Officer D. Echard, Repair and Replacement Engineer R. Engen, Civil / Structural Engineer -

- R. Fincher, Manager, Quality Maintenance M. Gibbons, Local Leak Rate Test Progrr.n Engineer W. Harrison, Licensing M. Kanvos, Manager, Mechanical-Civ C. McIntyre, Director, Engineering P ojects M. Oswald, Supervisor, Design Enr.neering S. Patel, Senior Consulting Engint er, Design Engineering Department U. Patil, Senior Consulting Engin.ser, Steam Generator Replacement Project L. Peter, Shift Supervisor / Stear Generator Replacement Project i D. Prater, Consulting Enginee,, Manager Return-to-service Testing K. Richards, Steam Generator Replacement Project S. Thomas, Manager. Design Engineering l S. Timmaruju, Stearn Generator Contract Technical Coordinator l M. Van Noy, Licensing Engineer-l T. Walker, Manager, Procurement Quality -

D.,Wirfs, Quality / Steam Generator Replacement Project D. Wohleber, Manager, Steam Generator Replacement Project NRC i l

N. O'Keefe, Senior Resident inspector, South Texas Project W. Sifre, Resident inspector, South Texas Project j

G. Guerra, Resident inspector, South Texas Project I

i I 1,NSPECTION PROCEDURES USED IP 50001 Steam Generator Replacement inspection ,

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~ LIST OF DOCUMENTS REVIEWED i

DOCUMENT TITLE DATE Change 96-2842-4 Reactor Coolant System Work May 5,1998

l Change 96-2843-1 : Large Bore Secondary System - 1RE08 March 3, '1998

.- Baseplates Change 96-2843-2 ' Large Bore Secondary System Pipe Work (FW, October 1,1998 AF, MS)

Change 96-2844- Stearn Generator Water Level Tubing / Piping August 13,1998 l Removal / Reinstallation -

. Change 96-2844-2 Steam Generator Blowdown System Piping July 27,1998 '

Removal / Reinstallation l Change 96-2845-1' Make Biological Shield Walt D Removable August 4,1998 j t 4

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Change 96-2845-11 STP Unit 1 Steam Generator Replacement Polar March 3,1998 Crane Evaluation Report I

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